10. PRACTICAL RADIATION MONITORING AND DOSE EVALUATION
10.1 Monitoring
10.1.1 Workplace monitoring
10.1.2 Individual monitoring
10.2 Dose evaluation
10.2.2 External exposure
10.2.2.1Calculation of external dose
10.2.2.2Dose reconstruction
10.2.2.3Whole body dose and extremity dose
10.2.3 Internal exposure
10.2.3.1Calculation of internal dose
10.2.3.2Dose calculation code
10.2.3.3Whole body dose and organ dose
10.2.4 Total dose
10. PRACTICAL RADIATION MONITORING AND DOSE EVALUATION
10.1 Monitoring
10.1.1 Workplace monitoring
10.1.1.1 General principle
放射線・原子力緊急事態では、空間線量率、空気中放射性物質濃度、表面汚染密度
の測定を実施すべきである。また、放射線の過渡的な変化に注目し、放射線のレベル
の変化にも注意を払うことが重要である。
In radiation and/or nuclear emergency situation, ambient dose rate, airborne
contamination and surface contamination in the workplace of emergency workers should be
monitored. Furthermore it is important to pay attention to the transient change in the radiation
revel.
作業場所の測定には、その目的に応じて、線源関連のモニタリングと作業関連のモ
ニタリングの２つのタイプがある。線源関連のモニタリングは、線源や施設、設備の
状態を監視する目的のため、定点において連続的あるいは定期的に行うことが望まし
い。作業関連のモニタリングは、緊急作業者の個人線量を適切に管理するため、作業
が行われる場所において作業中随時行うことが望ましい。
Workplace monitoring can be divided into two types, source related monitoring and task
related monitoring, in accordance with the purpose of the monitoring. In source related
monitoring, it is desirable that a monitoring is performed continuously and/or periodically at
the fixed monitoring point in order to supervise the situation of the source and the facilities. In
task related monitoring, it is desirable that monitoring is performed beside emergency workers
during the task implemented in order to control individual dose of emergency workers
adequately.
定置型の放射線モニタリング設備が設置されている施設においては、コントロール
ルームなどに設置された放射線監視パネルで放射線状況を包括的に確認することがで
きる。その場合には、警報、作業環境の放射線状況のみならず、放射線モニタリング
に必要な電源が確保されているかどうか、モニタリング装置そのもの健全性に関心を
払うことも必要である。
In some facilities where area monitoring equipment is installed, the radiological situation in
the facilities can be confirmed comprehensively by indicator panel or terminal computer in the
control room of the facilities. It is important for radiological safety to pay attention to not only
the radiological indication and alert but also soundness of power supply and equipment of
radiation monitor itself.
定置型の放射線モニタリング設備がない施設においては、サーベイメータなどの可
搬型の放射線モニタを用いて放射線の測定を行う。線源の位置と放射線作業を行う場
所を考慮して適切なモニタリングポイントを決める。そのポイントにおいて定期的に
放射線のモニタリングを行う。測定の頻度は放射線レベルの変化を考慮して決定すべ
きである。測定の頻度はできる限り多いことが望ましいが、緊急時の放射線モニタリ
ングに従事する作業者の過剰な被ばくを避けるためには、頻度の最適化が必要である。
モニタリング設備の定期的な点検、校正や、緊急事態を想定した放射線測定装置の必
要数の確保などの事前の準備も重要である。
In the facilities without area monitor, radiation measurement should be performed by using a
portable radiation monitor or sampler such as survey meter or portable dust sampler. One
should decide an appropriate monitoring point in consideration of the workplace condition
such as distribution of radiation source and working time and position of emergency workers in
the workplace. Periodical monitoring should be performed at the fixed monitoring point.
Monitoring frequency should be decided considering change of radiation revel. In severe
accident of nuclear reactor it is preferable to monitor as often as possible, but optimization of
monitoring frequency will be needed to avoid over exposure of emergency workers who
engaged in emergency radiation monitoring.
The measurement result should be recorded clearly in a form of time table or time chart.
Implementation of periodical maintenance and calibration of monitors is important for accurate
measurement. To secure sufficient number of radiation monitors is also important for nuclear
emergency preparedness.
10.1.1.2 Measurement of ambient dose rate
空間線量率の測定には、２つのタイプがある。一つは作業環境の定点において連続
的あるいは定期的に行う測定であり、他は放射線作業に着目して行う測定である。
There are two types of an ambient dose measurement. One is the measurement performed
continuously or periodically in the fixed point of workplace, and other is the measurement
performed paying attention to radiological work.
前者は、定点の空間線量率の変化に着目しているので、作業環境の変化に対して敏
感で、異常の早期発見にも有用である。一般に、この測定は、γ線エリアモニタ、中
性子線エリアモニタなど定置式のモニタによって行われる。また、可搬型の放射線モ
ニターを用いて、定点で定期的に測定を行う方法がある。
The former is the measurement which paid its attention to time change of the air dose rate in a
fixed point. Therefore, it is sensitive to change of work environment, and useful to the early
detection of abnormalities. Generally this measurement is performed by installed monitors,
such as a gamma ray area monitor and a neutron area monitor. Moreover, there is the method
of measuring periodically in a fixed point using a portable radiation monitor.
後者は、作業の開始前、作業中、作業終了後などに可搬型のサーベイメータを用い
て実施され、測定された空間線量率が作業計画で想定している管理基準の範囲内にあ
ることを確認するなど、作業による個人の線量が計画された値から逸脱しないように
監視することを目的とする。
The latter aims at supervising so that the individual dose may not deviate from the planned
control revel. Therefore, measurement is carried out using a suitable survey meter before the
start of work, during work, and after the end of work.
A suitable and efficient survey instruments and area monitors which are matched to the
specific task in radiological emergency situation should be capable of providing direct readings
of the dose equivalent rate in microsieverts per hour (mSv/h or mSv•h-1). A smaller number of
instruments indicate the absorbed dose rate in micrograys per hour (mGy•h-1). Ionization
chamber survey instrument and/or scintillation survey instrument is usually used for
monitoring of X, gamma and/or beta radiations. Specialized instruments are necessary to
measure neutron dose equivalent rates. Dose rate meters may not be able to provide an accurate
response to rapidly changing or pulsed radiation fields. Integrating dose rate meters and
dosimeters are more appropriate in such circumstances.
Survey instruments used for dose assessment should be type tested and calibrated in
terms of the operational quantities H*(d) and H′(d), and should operate within prescribed
criteria for overall accuracy, taking into account the dependence on radiation energy, direction
of incidence, temperature, radiofrequency interference and other influence quantities. The
energy and direction dependencies of the response are particularly important.
Photon Survey instrument
Ionization chambers are used to manufacture accurate dose rate meters for photon
radiations and beta radiations. A typical construction is illustrated:
A — the cylindrical detector wall serves as the cathode (negative electrode) and is
normally made of air-equivalent, carbon coated plastic or aluminum.
B — the axial anode (positive electrode).
C — beta window made of thin foil (3-7 mg·cm-2).
D — protective buildup cap (200-300 mg·cm-3) made of toughened plastic or
aluminum.
The buildup cap is used to improve the detection efficiency when measuring high energy
photon radiations. It is removed when measuring dose rates due to low energy photons (10 to
100 keV) and beta radiations.
Detector volumes of a few hundred cubic centimeters are needed to measure exposures
in nanocoulombs per hour (nC·h-1). The processor converts these units to the appropriate
radiation dose rates from about 10 μSv·h-1. Ambient variations of temperature, humidity and
air pressure will affect the detector and should be corrected as appropriate. Detectors may be
screened against extraneous radiofrequency interference.
Scintillation survey instruments are also used for photon radiations. The essential
components of a scintillation counter are:
S — scintillator. A substance (a phosphor) contained within an opaque material. Ionizing
radiations interact with the scintillator, which almost immediately converts some of
the absorbed energy into a flash of light.
L — light guide transfers the scintillation to the photocathode (C) of the photomultiplier
(M).
C — photocathode is a translucent, light sensitive coating of material (e.g., antimonycaesium) on the photomultiplier window. When it absorbs light it emits a
proportionate number of electrons.
M — photomultiplier tube contains electrodes called dynodes. A successively in-creased
potential difference (about 2000 V overall) draws electrons to each dynode in turn.
The number of electrons increases at each dynode. The number of electrons is
amplified by about one million.
Phosphors include solid organic materials such as anthracene and stilbene, liquid
solutions of organic materials (liquid scintillants), solid solutions of organic materials (plastic
scintillants) and activated inorganic crystals such as sodium iodide and caesium iodide which
are activated by trace quantities of thalium (NaI(Tl) and CsI(Tl)).
The scintillation rate is proportional to the radiation fluence and not the dose rate, but
scintillation dose rate meters, that operate over limited energy ranges, can indicate low and
high dose rates even at low photon energies. However, the photomultiplier’s demand for a
stable voltage supply and high component costs make these instruments comparatively
expensive. Their bulk, weight and vulnerability to shock also limit their usefulness. However,
portable instruments are now available which provide spectrometry as well as dosimetry.
Neutron survey instrument
Air filled ionization chambers with thin internal coatings of boron are effective but
proportional counters which contain either boron trifluoride (BF3) or helium-3 (3He) gas form
more sensitive detectors. A typical instrument construction is shown:
D — display.
P — proportional counter incorporating BF3 or 3He.
M — cylindrical or spherical polyethylene moderator to thermalize incident fast neutrons
for detection.
S — perforated cadmium or boron plastic sheath which modifies the energy response
according to neutron quality factors to enable the direct measurement of dose
equivalent rates.
Gas filled proton recoil detectors are lined with polythene and may also incorporate
ethylene gas.
10.1.1.3 Measurement of airborne contamination and radioactive gas concentration
Airborne contamination meters and gas monitors indicate the potential internal exposure
when a radioactive material is distributed within the atmosphere.
Airborne contamination meters are used to detect radioactive aerosols which may be
present within the atmosphere. These may be dispersion aerosols (dusts), condensation aerosols
(smoke) or liquid aerosols (mists). The instruments used normally draw potentially
contaminated air at a constant rate through a filter. The instrument may then be capable of
detecting the accumulated radioactive material on the filter or the filter may need to be assessed
elsewhere.
Radioactive iodine such as 131I, 132I in workplace might include particulate iodine,
gaseous elemental iodine, and volatile organic iodine compounds. The sampling methods used
for radioactive iodine depend on the expected forms of the iodine and the presence of possible
interfering radionuclides. A high efficiency particulate filter, such as a glass fiber filter, should
be used in airborne sampling for particulate iodine. Activated charcoal cartridges in which the
charcoal has been impregnated with tri-ethylene diamine (TEDA) or silver-zeolite (zeolite that
has been impregnated with silver) should be used in airborne sampling for the gaseous
elemental iodine and/or volatile organic iodine. Collection efficiency of activated charcoal cartridge
to iodine depends on the physicochemical form of iodine and the environmental condition such as the
thickness of activated charcoal layer, the velocity of sampling airstream, the atmospheric temperature,
the relative humidity and the presence of possible interfering radionuclides.
Table xx example of collection efficiency of activated charcoal cartridge/filter to iodine
Source of
iodine
Experiment[1]
Experiment[2]
Laboratory[2]
Chemical
form
elemental
iodine
20%
organic
iodine
80%
organic
iodine
100%
Material
of
collection
cartridges
charcoal
cartridge
with
18mm
thickness
(CHC50)
glass
fiber
filter
with
activated
charcoal
Chemical
agent
impregnated
in charcoal
cartridge
TEDA
10wt%
Snl2 1wt%
Relative
humidity
(%)
collection time (h)
0.5
2.0
8.0
99
98
97
96
96
85
65
65
99
60
40
35
60
95
90
87
87
90
70
62
53
53
100
none
TEDA 5wt%
Collection efficiency (%)
34-75
(*1) collection time: 24h-48h
[1] M. Naritomi et. al; J. Nucl. Sci. Technol. 10(5),292(1873)
15
17
Velocity
of air flow
or air
sampling
volume
20 cm/sec
20 cm/sec
8593(*1)
50 L/min
[2] S. Kato et. al; J. Japanese health physics society, 21(1986) p9-15
Gas monitors contain a radiation detector and continuously sample the air directly to
measure the presence of radioactive gases. The contaminant must be identified in order to
determine the activity concentration in becquerels per cubic meter (Bq·m-3).
Airborne contamination meters and gas monitors may be used to assess airborne
contamination in the workplace. Personal air samplers (PAS) are used to monitor the often
more significant hazard within the breathing zone of an individual worker. These instruments
are often passive devices which do not provide immediate results. They only provide
retrospective assessments of the working conditions but may also provide estimates of intakes.
Instruments that are capable of detecting the radionuclide may be used as active devices
to provide an alarm signal when the airborne radioactivity concentration reaches a pre-set value.
10.1.1.4 Surface contamination monitors
Surface contamination monitors usually are designed to measure a specific type of
radiation and often have optimum detection efficiency over a limited range of radiation
energies. For example the detector may respond only to alpha particles or gamma radiation or
beta and gamma radiation. It may perform better in detecting high energy beta particles rather
than those of low energy; or it may be designed to detect low energy gamma radiation but not
high energy. It is important to select an instrument that has a detection efficiency optimized for
the radiation (or isotope) of interest.
Most surface contamination monitors indicate in counts/s (or s-1) or counts/min and the
instrument needs to be calibrated for the particular radiation being detected to enable the
indicated reading to be converted into meaningful units such as Bq/cm2. Some instruments are
designed to allow either the calibration response factor to be programmed into the instrument
or the isotope being used, perhaps as a radiotracer, to be selected from a list on the instrument
so that response is automatically corrected and the reading is displayed directly in Bq/cm2.
Alpha contamination monitors are intrinsically sensitive to radioactive material of alpha
emitter because they do not respond to gamma radiation and consequently have no background
count rate. However, they are vulnerable to mechanical damage and cannot be used reliably to
measure surface contamination where the surface is irregular (for example, uneven or curved)
or covered in a thick layer of radioactive material of alpha emitter (which self-absorbs the
radiation) or wet (with degrees of moisture producing variable self-absorption).
A beta contamination monitor will indicate whether radioactivity of beta/gamma emitter
is present within a facility. It is unlikely that a beta contamination monitor will provide
accurate quantitative measurements of the surface contamination (in Bq/cm2) because
assumptions made about the radioactive constituents of the contaminant may not be entirely
correct and significant self-absorption of the beta radiation occurs in all but thin layers of
contamination. At best, beta contamination measurements provide a reliable indication of the
need for radiation protection measures and further investigation by sampling and radionuclide
analysis.
10.1.2 Individual monitoring
10.1.2.1 External exposure
外部被ばく線量の測定は、個人線量計を用いて行う。個人線量計には電源を必要としない
パッシブ線量計と電源を必要とするアクティブ線量計がある。
パッシブ線量計は、積算線量を測定する線量計で、線量の読み取りは専用の読み取り装置
を用いる。個別の電源が不要であり災害時にも確実に線量の測定を行うことができる。線量
計の価格は比較的安価である。積算線量が記録されるため、使用前の線量の確認や積算線量
をリセットするための処理を必要とする。また、単位時間毎の個人線量など過渡的な線量の
変化を記録することはできない。代表的なパッシブ線量計には、TLD、OSL、ガラス線量計、
フィルムバッジなどがある。
A individual dosimeter is used for measurement of an external dose. There are a passive dosimeter
which does not need a power supply, and an active dosimeter which needs a power supply in a personal
dosimeter. A passive dosimeter is a dosimeter which measures an accumulated dose. Reading of a
passive dosimeter is performed using exclusive reading equipment. Since the battery is unnecessary, the
passive dosimeter can ensure measurement of a dose, also when charge is impossible. Moreover, the
price of passive dosimeters is reasonable. However, time change of doses, such as an individual dose
for every unit time, can not be measured. There are TLD, OSL, a glass dosimeter, a film badge, etc. in a
typical passive dosimeter. Although TLD, OSL, and the glass dosimeter can carry out repeated use, the
check of the dose before use, or process for resetting an accumulated dose is needed.
アクティブ線量計は、放射線の計数を積算して単位時間及び設定した時間の積算線量を測
定する個人線量計である。線量計毎に電源を必要とするため、線量計の単価はパッシブ線量
計に比べて高価である。また、使用前に電源レベルを確認する必要があり、数日間にわたる
連続的な使用には適さないため、災害時には注意を要する。線量率の変化や単位時間の積算
線量の時間変化などを記録することができるタイプのものもあり、放射線の管理に有用であ
る。代表的なアクティブ線量計には、Si 半導体検出器を用いた電子式線量計がある。
An active dosimeter is an individual dosimeter which integrates the counts of radiation and
measures the accumulate dose of unit time and the set-up time. Since a battery is needed for each
dosimeter, the unit price of a dosimeter is expensive compared with a passive dosimeter. Since the
active dosimeter needs to check the power of a battery before use and generally is not suitable for
continuous use over several days, it requires attention for use at a power blackout in the disaster. The
dosimeter which can record change of a dose rate or an accumulated dose in unit duration is useful to
radiation control. There is an electronic dosimeter which used Si semiconductor detector in a typical
active dosimeter.
10.1.2.2 Internal exposure
Typical methods of individual monitoring for intakes are whole body counting, organ counting
(such as thyroid or lung monitoring) and analysis of samples of excreta.
Sampling of the breathing zone with personal air samplers is also used. In many circumstances
involving exposure due to radionuclides, workplace monitoring will be needed. Monitoring procedures
may be introduced to demonstrate satisfactory working conditions or in cases where individual
monitoring is unable to provide adequate protection of the worker. Such workplace monitoring may
also be appropriate when levels of contamination are low, for example in a research laboratory using
small quantities of radioactive tracers.
Monitoring for the estimation of doses from intakes of radionuclides may include one or more of
the following techniques:
(a) Sequential measurements of radionuclides in the whole body or in specific organs;
(b) Measurements of radionuclides in biological samples such as excreta or breath;
(c) Measurement of radionuclides in physical samples such as filters from personal or fixed air
samplers, or surface smears.
Measurements can be used to calculate the intake of a radionuclide, which, when multiplied by
the appropriate dose coefficient, leads to an estimate of committed effective dose. Dose coefficients for
a wide range of radionuclides are given in the BSS [ ], and those for selected radionuclides are
reproduced in Table A-1 of the Annex. In some circumstances, the results of direct measurement may
be used to calculate dose rates to the whole body or to individual organs.
プルトニウム同位体のようなα線放出核種の場合や 90Sr などのようにβ線だけを放出す
る核種の場合には、便や尿などの排泄物中の放射能分析を行って摂取量を算定する。放射性
物質の摂取の時期が明確である場合には、感度の良い線量評価のために放射性物質の摂取直
後から数日間の 24 時間の便と尿を採取することが望ましい。
In the case of internal contamination by the alpha particle emitter such as the plutonium isotopes
or the pure beta particle emitter such as 90Sr, radiochemical analysis of the excrement, such as faeces
and urine, is conducted. If the time of intake is known, it is desirable to collect daily (24h total) urine
and/or faeces for the first 3 - 5 days after the intake in order to perform highly sensitive dose evaluation.
限度に近い線量が予想される場合には、摂取した放射性物質の体内残留量や排泄率を考慮
して適当な頻度で複数回の測定を行うことが望ましい。それらの測定値から最も合理的な摂
取量を求めることができる。
When it is predicted that there is exposure near a dose limit, it is preferable to make two or more
measurement in suitable frequency in consideration of the retention or excretion of the radioactive
substance concerned. Then the most rational intake can be calculated from those measured value.
10.2
Dose evaluation
After an accident has occurred, the radiological consequences may be complicated by trauma or
other health effects incurred by the workers. Medical treatment of injuries, especially those that are
potentially life threatening, generally takes priority over radiological operations, including exposure
assessment. In such cases, post-accident dose assessment should be conducted when the situation has
been brought under control.
10.2.1 External exposure
10.2.1.1 Calculation of external dose
An external dose is calculated based on the measurement result of a individual dosimeter. For radiation
protection purposes the measured operational quantities Hp(10) and Hp(0.07) are interpreted in terms of
the protection quantities effective dose E and equivalent dose to the skin and extremities HT. To do this,
realistic assumptions have to be made with respect to the type and uniformity of the radiation field and
the orientation of the worker within the field [15]. Under these conditions, the dosimeter reading gives a
good estimate of the worker’s exposure without underestimating or severely overestimating the relevant
protection quantity.
In cases where the worker moves about the workplace, four types of multidirectional field should
generally be considered:
(a) with radiation incident predominantly from the front half space (anterior–posterior, or AP
geometry) or
(b) from the rear half space (posterior–anterior, or PA), or
(c) with radiation incident symmetrically from all directions perpendicular to the body (rotational,
or ROT), or
(d) with radiation incident isotropically from all directions including above and below (ISO)
If the radiation is expected to come from the rear (e.g. for the driver of a vehicle transporting
radioactive materials), the dosimeter should be worn on the back. For strongly penetrating radiation it
may be assumed that Hp(10) measured by a personal dosimeter worn on the chest approximates the
effective dose sufficiently accurately, at least for radiation which is incident from the front or is
cylindrically symmetrical (ROT). Thus, one dosimeter worn on the front (or rear) of the trunk generally
provides a satisfactory assessment of the effective dose. However, if the dose approaches the relevant
limit, an appropriate correction factor should be applied for AP, PA, ROT or ISO geometry, based on a
knowledge of the radiation and the conditions of exposure.
Survey instruments and area monitors are calibrated in radiation fields that irradiate the detector
volume uniformly, with the centre of the volume used as a reference point. However, many operational
fields irradiate the detector in a non-uniform manner (e.g. close to point sources or narrow beams).
These situations need special attention and it may be necessary to establish a correction factor that can
be applied to the readings to give a corrected dose rate.
10.2.1.2Dose reconstruction
個人線量計を着用していない場合や、個人線量計を着用していても、不均一な被ばく状況
などによって個人線量計による測定結果をそのまま実効線量や等価線量の代用として用いる
ことができない場合がある。このような場合には、実験や計算によって実効線量や等価線量
の評価を行うことが必要となる。必要に応じて実験的に被ばくの状況を模擬して個人線量計
の測定結果の検証を行う。
There is a case where the personal dosimeter is not worn. Or there is a case where the
measurement result of a individual dosimeter may be unable to use as substitution of an effective dose
or equivalent dose because of non-uniform exposure. In these cases, it is necessary to perform
evaluation of an effective dose or an equivalent dose by an experiment or calculation. In order to verify
the measurement result of a personal dosimeter, the reappearance experiment of the exposure situation
is conducted if needed.
10.2.1.3Whole body dose and extremity dose
全身に均等に被ばくした場合は、個人線量計の値を実効線量と見なす。不均等な被ばくの
場合には、個人線量計の値や、線量再構築などの結果から、組織ごとの等価線量を評価し、
組織荷重係数を乗じて実効線量を評価する。手の皮膚などの等価線量は、指リング線量計な
どによってＨp(0.07)の値を評価する。
When worker is exposed to the radiation to the whole body uniformly, the value of a personal
dosimeter can be seen to be equivalent to the effective dose. In case of non-uniform exposure,
equivalent dose of each tissues or organs should be evaluated based on the measured value by
individual dosimeter and the result of dose reconstruction. An effective dose can be evaluated by
multiplying those equivalent doses by the corresponding tissue weighting factors. Equivalent doses of a
skin in a hand can be evaluated by using the value of Hp (0.07) measured by a finger ring dosimeter.
10.2.2 Internal exposure
Accidents may result in releases of radioactive materials into the working environment with the
potential for high doses to the workers. Details of implementation of dose assessment of exposure due
to intakes of radionuclides are described in the IAEA SAFETY GUIDE No. RS-G-1.2.
10.2.2.1Calculation of internal dose
Once assessment of internal exposure has commenced, as much information should be gathered as
is practicable. For example, information will be needed on the time and nature of the incident and the
radionuclides involved, and on the timing of bioassay samples and measurements of body activity. This
information may be necessary not only for exposure assessment, but also to assist in medical assessment,
to guide medical treatment of the victim (which may include chelation therapy or wound excision), and
to assist later in reconstruction of the accident or incident itself and in long term medical follow-up of
the victim.
Because intakes associated with accidents or incidents can result in committed effective doses
which approach or exceed dose limits, individual and material specific data are normally needed for
exposure assessment. These data include information on the chemical and physical forms of the
radionuclide(s), the particle size, airborne concentrations, surface contamination levels, the retention
characteristics in the individual affected, nose blows, face wipes and other skin contamination levels
and external dosimetry results. The various items of data will often seem to be inconsistent or
contradictory, particularly if the intake period is uncertain. An adequate assessment of dose can be
made only after considering all of the data, resolving the sources of inconsistency as far as is possible,
and determining the most likely and worst possible scenarios for the exposure and the magnitude of any
intake.
事故の状況によっては、作業者が摂取したすべての核種を測定できない場合がある。例えば、
この事故においては、１３１Ｉ、１３２Ｉ、１３２Ｔｅなどの短半減期の核種を検出できな
かった。その原因は、事故直後にモニタリングを開始できなかったためである。その理由は、
緊急作業に従事した作業員が、長期間現場にとどまったためである。このような場合には、
作業環境における放射線モニタリングの結果などから、作業者が摂取したと考えられるすべ
ての核種を決定し、摂取経路を考慮して摂取量を算定する。吸入摂取の場合には、核種毎の
空気中放射性物質濃度に作業者の呼吸率及び作業時間を乗じて摂取量を算定する。個人モニ
タリングにおいて、核種の一部が実測されている場合は、ｗｏｒｋｐｌａｃｅ ｍｏｎｉｔ
ｏｒｉｎｇの結果に基づいて、摂取したと考えられる核種の組成を決定する。そうすれば、
実測された核種の摂取量と、決定した核種組成に基づいて、すべての核種の摂取量を求める
ことが可能な場合もある。
All the nuclides which the worker ingested or inhaled may be unable to be measured in the severe
situation of accident. For example, in TEPCO dai-ichi nuclear power plant accident, the nuclide of short
half-life, such as 131I, 132I, and 132Te, was not able to be detected in a part of emergency workers. The
cause was because monitoring was not able to be started immediately after accident. The reason is
because the authorized personnel engaged in the emergency activity remained in the field for a long
period of time in order to complete their mission. In such a case, from the result of the workplace
monitoring, etc., all the nuclides considered that the worker ingested should be determined, and the total
intake should be calculated from the result in consideration of an intake condition. In the case of
inhalation, when radioactive substance concentration in the air for every nuclide is measured in
workplace monitoring, intake should be calculated by the way in which radioactive substance
concentration is multiplied by a worker's ventilation parameters and work hours. When some nuclides
are measured in personal monitoring, the nuclide composition considered to be ingested or inhaled is
determined based on the result of workplace monitoring. Then, based on the nuclide presentation
determined as the intake of the surveyed nuclide, it may be possible to calculate the intake of all the
nuclides.
10.2.2.2Dose calculation code
内部被ばくに関係するモニタリング量と摂取量、あるいは摂取量と実効線量の関係は、
核種、物理化学的性状、摂取経路、急性摂取や慢性摂取などの摂取パターン、摂取からモニ
タリングを実施するまでの経過時間、モニタリングの頻度や種類などに依存する。これらの
パラメータを考慮して、内部被ばく線量の計算を行うための計算コードが準備されている。
In general, the predicted value of retention, the predicted value of excretion and the dose
coefficient depend on the many parameters such as nuclide, physical and chemical form, path of intake,
intake pattern(acute or chronic), time after intake, etc. . For the purpose of considering these parameters,
the calculation cord for calculating a dose of internal exposure is prepared [NRPB, NIRS, JAERI, etc..]
10.2.2.3Whole body dose and organ dose
全身の線量は実効線量で表す。実効線量は、摂取量に実効線量係数を乗じて求められる。
組織、臓器別の線量は等価線量で表す。等価線量は摂取量に等価線量係数を乗じて求められ
る。実効線量係数、等価線量係数は文献[ICRP Pub. 72]に与えられている。
The whole body dose is expressed with an effective dose. An effective dose is calculated by
multiplying an intake by dose coefficient. The tissue dose and organ dose are expressed with an
equivalent dose. An equivalent dose is calculated by multiplying an intake by an equivalent dose
coefficient. The effective dose coefficient and the equivalent dose coefficient are given in the literature
[ICRP Pub.72].
10.2.3 Total dose
実効線量
外部被ばくによる実効線量と内部被ばくによる預託実効線量を合計する。
Effective dose
The effective dose by external exposure and the committed effective dose by internal exposure
should be totalled.
等価線量
水晶体
外部被ばくについては、頭部や襟などに着用した個人線量計によって測定したγ線、β
線、中性子線による Hp(3)の合計値を水晶体の等価線量とする。
Equivalent dose
lens of eyes
The equivalent dose of lens of eyes should be evaluated based on Hp (3) measured by the
personal dosimeter worn on a head or a collar, etc. When photon, beta-rays, and neutron are
intermingled in workplace, those Hp(3) should be totalled.
皮膚
強い透過性の放射線による被ばくが主となる作業環境や弱い透過性の放射線による被ばく
であっても全身にほぼ均一な被ばくとなるような作業環境おける外部被ばくに関しては、個
人線量計によって測定された Hp(0.07)を皮膚の等価線量と見なして良い。局部的に皮膚の線
量が高くなると想定される場合には、その部位にも局部の線量計を装着し、その部位で測定
された Hp(0.07)を皮膚の等価線量と見なす。指リング線量計はその例である。皮膚汚染や着
用していた作業服などに汚染があった場合には、その表面汚染密度の測定結果と核種毎の換
算係数及び汚染からの被ばくを受けていた時間から皮膚の等価線量を計算する。核種毎の換
算係数は文献に与えられている。
Skin
In a workplace where a worker receives almost uniform exposure in the whole body, it can be
considered that Hp (0.07) measured by the personal dosimeter is an equivalent dose of the skin.
When it is assumed that the dose of the skin becomes high locally, extremity dosimeter should be set on
the part of body and it can be considered that Hp (0.07) measured in the part is an equivalent dose of the
skin. When skin contamination or the work clothing contamination was discovered, the skin dose
should be evaluated based on the measurement result of surface contamination on the skin or clothes.
The surface contamination and the skin dose conversion factor for every nuclide are given to literature.
甲状腺
外部被ばくに関しては、個人線量計で測定したＨp(10)に基づき必要ならば補正を行っ
て評価する。均一な被ばくに関しては、Ｈp(10)を甲状腺の等価線量と見なしてよい。不均
一な被ばくに関しては、甲状腺と個人線量計の着用の位置の違いによってもたらされる
Hp(10)の違いについて補正を考慮する。
内部被ばくに関しては、個人内部被ばくモニタリングによって求めた摂取量に核種別、
甲状腺の等価線量係数を乗じて甲状腺の等価線量を算出する。外部被ばくによる等価線量
と内部被ばくによる等価線量を合計して甲状腺の等価線量とする。
Thyroid
The external dose of the thyroid can be estimated from Hp (10) measured with the personal
dosimeter. In uniform exposure, Hp (10) can be considered to be an equivalent dose of the thyroid. In
non-uniform exposure, the discrepancy in Hp (10) brought about by the difference between the location
of the thyroid and the wear position of individual dosimeter should be corrected if necessary.
The internal dose of the thyroid is calculated by multiplying an intake by a dose coefficient. The
equivalent dose by external exposure and the equivalent dose by internal exposure are totalled, and it is
considered as the equivalent dose of the thyroid.
その他の組織、臓器
外部被ばくに関しては、全身に一様に被ばくを受ける場合には、個人線量計の Hp(10)
を着目する臓器の等価線量と見なして良い。不均等な被ばくの場合には、必要に応じて線量
再構築によって着目する臓器の等価線量を推定する。
内部被ばくに関しては、単位摂取量あたりの組織、臓器別の線量係数が文献に与えられて
いるため、それを使って組織、臓器の等価線量を求める。
外部被ばくによる等価線量と内部被ばくによる等価線量を合計して組織、臓器毎の等価線
量とする。
Other tissues and organs
In uniform external exposure, Hp (10) can be considered to be an equivalent dose of the thyroid.
In non-uniform exposure, the equivalent dose of the organ aimed is estimated by suitable method, such
as dose reconstruction. Since the equivalent dose coefficient according to each tissue and organs is
given to literature [ICRP xx] the equivalent dose of tissue and organs can be calculated using it. The
equivalent dose of each tissue and organs is defined as the sum of the equivalent dose by external
exposure and the equivalent dose by internal exposure.