The Series of Environmental Radioactivity Measuring Methods
DR-CNT(No.17)

Method for Measurement ofEnvironmental Gamma-rays with aContinuous Monitor

Amendment of December 2017
Radiation Monitoring Division
Radiation Protection Department
The Secretariat of
Nuclear Regulation Authority

Table of Contents

Chapter 1

Introduction....................................................................................................................... 1

Chapter 2

Glossary ............................................................................................................................ 2

Chapter 3

Configuration of a continuous monitor............................................................................. 6

3.1 Outline ................................................................................................................................... 6
3.2 Continuous monitor detector.................................................................................................. 6
3.3 Recording of measurements................................................................................................... 8
3.4 Telemeter system ................................................................................................................... 8

Chapter 4

Measuring system using a continuous monitor............................................................... 10

4.1 Outline ................................................................................................................................. 10
4.2 Installed continuous monitor.............................................................................................. 10
4.3 Transportable continuous monitor ...................................................................................... 13
4.4 Mobile survey system .......................................................................................................... 15

Chapter 5

Installation of a measuring system.................................................................................. 19

5.1 General considerations of the installation of a continuous monitor..................................... 19
5.2 Installed continuous monitor................................................................................................ 20
5.3 Transportable continuous monitor ...................................................................................... 25
5.4 Mobile survey system .......................................................................................................... 26

Chapter 6

Measurement and calibration .......................................................................................... 28

6.1 Precautions for use ............................................................................................................... 28
6.2 Installed continuous monitor................................................................................................ 31
6.3 Transportable continuous monitor ...................................................................................... 33
6.4 Mobile survey system .......................................................................................................... 34

Chapter 7

Analysis and evaluation of measurement results ............................................................ 36

7.1 Outline ................................................................................................................................. 36
7.2 Analysis and evaluation during normal operation ............................................................... 36
7.3 Analysis and evaluation in case of an emergency................................................................ 37
7.4 Dose ................................................................................................................................... 40
7.5 Analysis and evaluation of each measuring system ............................................................. 41

Chapter 8

Measurement of environmental gamma-ray spectrum by a NaI monitor ....................... 44

8.1 Outline ................................................................................................................................. 44
8.2 System configuration required for environmental gamma-ray spectrum measurement ...... 44
8.3 Preparation for measurement ............................................................................................... 45
8.4 Primary processing of measured data .................................................................................. 46
8.5 Data acquisition for energy calibration ................................................................................ 47

Chapter 9

Response function used by NaI monitor for measured data ........................................... 48

9.1 Description of a response function ...................................................................................... 48
9.2 Creation of a response function ........................................................................................... 48
9.3 Application of a response function ...................................................................................... 49

Chapter 10

Processing of data on environmental gamma-ray spectrum measured by NaI monitor . 50

10.1 Energy calibration of the measured pulse-height distribution ............................................. 50
10.2 Conversion to an incident spectrum..................................................................................... 51
10.3 Representation of the results ................................................................................................ 52
10.4 Calculation of dose contribution by radionuclide ................................................................ 52
Explanation
Explanation A General matters on environmental gamma-rays measured by a continuous monitor ...57
Explanation B Principle, configuration, and characteristics of each monitor.......................................58
Explanation C Reference gamma-ray source .......................................................................................76
Explanation D Countermeasures against scattered gamma-rays during calibration .............................78
Explanation E Considerations of continuous measurement using a transportable monitoring post.....82
Explanation F Considerations of the mobile survey system ................................................................87
Explanation G Analysis of measurement results ..................................................................................93
Explanation H Integrity check of a continuous monitor by comparative measurement .....................103
Explanation I

Contribution ratio of the dose rate during an accident................................................106

Explanation J Creation of a response function ..................................................................................109
Appendix
Appendix 1

Calculation example of a response function ...............................................................123

Appendix 2

Example of data processing of NaI monitor by stripping method ..............................135

Appendix 3

Example of unfolding of NaI (Tl) scintillation spectrum by successive
approximation .............................................................................................................146

Appendix 4

References ..................................................................................................................150

Chapter 1

Introduction

The first edition of this measuring method was established in 1982 to measure environmental gammaray dose rates by an installed continuous monitor, mainly as part of environmental radiation
monitoring around nuclear facilities. It was revised in 1996, and it is still being developed.
After the accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Station(hereinafter referred to as the “Fukushima Dai-ichi NPS accident”) caused by the Great East Japan
Earthquake on March 11, 2011, the existing monitoring framework was reviewed. In light of thesecircumstances, the present situation is reflected in the current revision with the basic concept andobjectives inherited, considering the recent development of measuring instruments, as well as theexperience acquired, and problems found after the Fukushima Dai-ichi NPS accident.
In the past, the role of continuous monitoring was to provide information that would contribute to themonitoring of any abnormalities caused by nuclear facilities and the evaluation of external exposure
doses. However, after the Fukushima Dai-ichi NPS accident, it may be used as information necessary
for determining whether further protective measures should be taken for the residents, etc., in case of a
nuclear disaster.
From the experience of the Fukushima Dai-ichi NPS accident, it turned out that, once a nuclear
disaster occurs, the effects of released radionuclides would have extensive and long-term impacts.
Because conventional “Installed-type” continuous monitoring is insufficient, various methods havebeen developed and utilized for efficient and rapid monitoring. Based on the current situation, thismeasuring method also includes “transportable” continuous monitoring and “Mobile surveys.” In
addition, the gamma-ray energy spectrum, together with the environmental gamma-ray dose rate,
provides very useful information for estimating the causes of abnormal values. Therefore, part of thecontent of the radioactivity measurement series No. 20 “Environmental gamma-ray spectrometry” isprovided for integration.
“Measurement of the environmental gamma-ray dose rate” described in this measuring method must
be applicable to various situations, because uncertainties are involved every time a measurement istaken. Sorting the uncertainties involved in the measurements in advance and understanding theircharacteristics will enable responding to a nuclear disaster such as the Fukushima Dai-ichi NPS
accident. This measuring method should be used as a basis for determination when users perform
appropriate measurement in their own situation.
The main users of this measuring method include the local government personnel responsible foroverseeing nuclear facilities.

1Chapter 2

Glossary

This chapter explains the terms used in this measurement series, and describes the important terms
used in the environmental gamma-ray dose rate measurement.
Natural radioactive nuclides (natural radionuclides) (Reference 1)
Naturally occurring radionuclides.
Artificial radionuclides (Reference 1)
Radionuclides artificially produced by nuclear fission, activation, etc.
Energy characteristics
The characteristics of the response or calibration constant of a radiation detector that depend on theenergy of radiation.
Directional characteristics
The characteristics of the response or calibration constant of a radiation detector that depend on theincident direction of radiation to the detector.
High-energy components and low-energy components
Environmental radiation can be mainly classified into cosmic rays, radiation from radioactive
materials existing in the environment, and radiation from radioactive materials released from
nuclear facilities, and the energy of radiation is mostly 3 MeV or less, except for cosmic rays. In thismeasuring method, the energy above 3 MeV is classified as a high-energy component, and theenergy below 3 MeV is classified as a low-energy component.
Response
The ratio of an indicated value of the monitor to the amount to be measured. As sensitivity usually
includes a sensitivity limit and is confusing, we call it response here. It is the inverse of a calibration
constant.
Pulse-height distribution
A pulse-height distribution of the spectrum of the pulse as measured.
Corrected pulse-height distribution
A distribution determined by correcting the pulse-height distribution such that the ratio of the peakvalue to energy is constant.
Incident gamma-ray spectrum
A spectral distribution, based on the corrected pulse height distribution, determined by applying a
response function to transform the spectrum to the gamma-ray energy one.
Self-contamination
Self-contamination becomes evident when a small number of radionuclides are contained in thematerials of radiation detector, and the measuring system is affected; this would create a problem in
low-level radiation measurements.
Nuclear Emergency Preparedness Guideline
A guideline presented by the Nuclear Regulation Commission after the Fukushima Dai-ichi NPS
accident, providing a concept of monitoring during a nuclear disaster.

2

UPZ
The areas where emergency protective actions are to be prepared as specified in the Nuclear
Emergency Preparedness Guideline. This stands for the Urgent Protective Action Planning Zone.
This is applied to the areas within approximately 30 km of nuclear facilities.
Tissue weighting factor (Reference 1)
A factor that multiplies the equivalent dose of a tissue or an organ, to account for the relative
probabilistic damage resulting from each exposure. The value of this factor is assigned to eachtissue or organ.
Radiation weighting factor (Reference 1)
A relative factor that is multiplied by the absorbed dose, to account for differences in the effects ofdifferent types of radiation impacting on health.
Nondestructive inspection (Reference 1)
A generic term for a method of inspecting an object without destroying it. Some inspections use
radiation, ultrasound, electromagnetic induction, fluorescent dyes, etc. It is also called anondestructive test.
Radiopharmaceuticals (Reference 1)
Medicine labeled with a radionuclide and used in diagnosis and treatment.
Folding (Reference 1)
A mathematical method for determining the output pulse-height distribution of a radiation detector
from the energy response function of the detector and the energy spectrum of incident radiation.
Unfolding (Reference 1)
A mathematical method for determining the energy spectrum of the original incident radiation fromthe output pulse-height distribution of a radiation detector and the energy response function of thedetector.
Response function
A function used to determine the spectrum of a gamma ray incident on the detector from thedetected pulse-height distribution, including the efficiency of the detector in terms of the detector's
response at the time of monoenergetic photon incidence.
Successive approximation method
An analytical method to minimize inaccuracies that occur in (deconvolution or unfolding) when
determining a gamma-ray incident spectrum from multiple response functions (a response matrix),
by successively increasing the order of approximation.
Geometry (Reference 1)
A term used to express the spatial arrangement of various elements. The term “geometricarrangement” is also used.
Tissue-equivalent material (Reference 1)
A material with an effective atomic number equal to that of a biotissue.
Air-equivalent material (Reference 1)
A material with an effective atomic number equal to that of air.
Phantom (Reference 1)
A human phantom used to simulate and measure the attenuation, scattering, or distribution ofradioactive materials in a living body.
ICRU Sphere (Reference 1)
A spherical homogeneous phantom with a diameter of 30 cm, consisting of tissue-equivalentmaterials with a density of 1 g/cm3. Its mass composition is 76.2% oxygen, 11.1% carbon, 10.1%
hydrogen, and 2.6% nitrogen.
3

Pulse amplifier (Reference 1)
An amplifier with wide-band characteristics for amplifying a pulse signal.
Preamplifier (Reference 1)
An amplifier connected immediately after a radiation detector and used to improve the impedance
matching and signal-to-noise ratio.
Proportional amplifier (Reference 1)
A pulse amplifier or a DC amplifier that outputs a signal of a peak value proportional to the peakvalue of the input signal. This is generally connected after a preamplifier and used as the main
amplifier.
Analog-to-digital converter (Reference 1)
A device for converting an analog signal into an output in digital quantities. This is used in a pulseheight analyzer, etc. Abbreviated as ADC.
Pulse-height analyzer (Reference 1)
A device for measuring the frequency distribution of the peak value of a pulse signal.
Single-channel pulse-height analyzer (Reference 1)
A type of pulse-height analyzer that selects input pulses within a given pulse-height range andoutputs logical signals to enable counting. Abbreviated as SCA.
Multichannel pulse-height analyzer (Reference 1)
A type of pulse-height analyzer with a large number of storage elements that accumulates input
signals for each pulse height to enable counting. This is also called a multichannel pulse-height
analyzer. Abbreviated as MCA.
Photomultiplier tube (Reference 1)
A vacuum tube for converting light into an electrical signal, which is essentially composed of aphotocathode and an electron multiplier. Abbreviated as PMT.
Detector window (Reference 1)
The part of a detector designed to facilitate the permeation of the radiation to be detected.
Environmental radiation monitoring equipment (Reference 1)
A generic term for equipment used to measure radiation dose rates, etc. near nuclear or radiation
facilities. This includes equipment to measure the concentrations of radioactive materials containedin environmental samples such as soil and sea water, and meteorological observation facilities.
Monitoring station (Reference 1)
A centralized facility for radiation measurement equipment that measures exposure dose rates andradioactivity of airborne dusts in the field around nuclear facilities or radiation facilities.
Monitoring post (Reference 1)
An installed facility in the field that continuously measures exposure dose rates, neutron fluence
rates, etc. in the environment.
Monitoring car (Reference 1)
An automobile for mobile measurement, equipped with a device to measure radiation orradioactivity in the environment.
Scintillation (Reference 1)
Light emission based on the excitation of atoms by ionizing radiation, with a duration of several
microseconds.
Scintillation decay time (Reference 1)
The time that the photon emission rate of scintillation takes to decay to 1/e of its initial value (e isthe base of a natural logarithm) after a single excitation.
Effective center (of a detector) (Reference 1)
4

A point in the effective volume that can be regarded as a center used for determining the output
characteristics of the detector relative to the exposure dose or exposure dose rate. Depending on theconditions, a non-negligible difference may be produced from the geometric center.
Counting loss (Reference 1)
When counting pulse signals, the measured count value might be lower than the actual value owing
to the dead time of the detector or counter, resolving time, pile-up of pulses, etc.
Dead time (Reference 1)
The time during which detection and counting capabilities are lost after the counter tube, countercircuit, etc. operate once. Counting cannot be performed for the incident radiation generated during
this time.
Resolving time (Reference 1)
The shortest time difference necessary for distinguishing two subsequent pulses or ionizing eventsas two independent pulses or events.
Pile-up (Reference 1)
A phenomenon in which the pulse-height indication becomes inaccurate as a result of a subsequent
pulse overlapping the tail of the preceding pulse.
Traceability (Reference 2)
The property of a measured result or a standard value that can be linked to a given standard througha chain of contiguous comparisons in which all uncertainties are expressed. National or international
standards are usually adopted as the standard.
Reference gamma-ray source
A gamma-ray source for which traceability has been established with national standards for the dose
rate certification. The exposure dose rate at a point of 1 m from the center of the source is usuallycertified.
Certified equipment with a display
Equipment certified by the system of design certification and specific design certification. The
equipment is sold by manufacturers and importers and can be used by filing a brief report to thecompetent authority (Nuclear Regulation Authority).
Weathering effectAttenuation of radiation dose rate and radioactivity concentrations owing to natural factors such as
wind and rain.

5

Chapter 3

Configuration of a continuous monitor

3.1 Outline
Various measurement principles and signal processing methods are adopted in the equipment used forcontinuous monitoring, and the output measurements are the resulting numerical values containing
various components. The components in the measured environmental gamma-ray dose rate areclassified into natural radiation and artificial radiation, each of which is affected by temporalfluctuation characteristics, energy spectral distribution, and meteorological factors.
The user is requested to fully understand the above description. For the measurement principles andsignal processing methods of the equipment used for continuous monitoring, refer to this chapter; forthe various components contained in the measured values, their fluctuation factors, and spectral
analysis, refer to Chapter 7 “Analysis and evaluation of measurement results” and Chapter 10“Processing of environmental gamma-ray spectrum measured data by a NaI monitor.”
Note that the detectors and various systems described in this measuring method are representative
examples of the equipment necessary for conducting environmental gamma-ray dose ratemeasurement, and handling of other equipment is not rejected. Especially in the event of a nuclear
disaster, optimal monitoring should be performed with limited personnel, time, and equipment.
Therefore, it is important to meet the specifications necessary to achieve the given objectives under therespective circumstances.
General matters on environmental gamma rays measured by a continuous monitor are described in
Explanation A.
3.2 Continuous monitor detector
This section describes the features of each monitor detector (Reference 3).
3.2.1
Ionization chamber-type detector
An ionization chamber-type detector is designed to detect radiation using the ionization of radiation,and it is a very primitive measuring instrument that has been used since the discovery of radiation.
With the incidence of radiation (gamma rays), secondary electrons are generated by the interaction
between the radiation (gamma rays) and the ionization chamber wall. When the generated electrons
pass through the gas in the ionization chamber, the filler gas is ionized to positive ions and electrons,
which are collected on the cathode and anode, respectively, by applying high voltage; then, a current
flows. The dose rate is expressed in an analog output of a current–voltage conversion by passing acurrent through a high resistance, or in a digital output of a charge by current integration at fixed timeinterval.
There are two types of ionization chamber dosimeters, atmospheric air type (1 atmosphere) andpressurized type. The latter is widely used as a continuous monitor. The atmospheric air type (1atmosphere) was not common because of its larger detector compared with the pressurized type;however, after the Fukushima Dai-ichi NPS accident, it has been used for measurement in areas withhigh dose rate. In terms of the detector energy characteristics and dose rate linearity, the atmospheric
air type (1 atmosphere) has good energy characteristics. The response of the pressurized type depends
on the material of the detector wall and the filler gas. The range of dose rate measurement can be large
for both the atmospheric air type (1 atmosphere) and the pressurized type. Details are provided in
Explanation B.
(1) Atmospheric air ionization chamber (1 atmosphere)
This type of detector has a sealed structure using a wall material (resin, etc.) of approximately 5 mmin thickness, which is considered to be equivalent to air, with an effective volume of approximately
6

20 L.
(2) Pressurized ionization chamberThis is an ionization chamber in which the gas filled in the radiation detection section is pressurized
to atmospheric pressure or higher. For example, an ionization chamber of this type has aluminum or
stainless steel wall, in which argon or nitrogen gas with a volume equivalent to 20 L or more at
atmospheric pressure is filled at a pressure of 8–20 kgf/cm2 (approximately 8–19 atmospheres), andhigh sensitivity is achieved in a small-sized chamber.
The energy characteristics and the dose rate linearity are as described in Explanation B.1; whencomparing the types using aluminum and stainless steel as the wall materials, a large difference is
identified in the range of 100 keV or less. This means that, when the proportion of 133Xe (81 keV),which is subject to measurement in case of an emergency, in the radioactive plume to the total dose
rate is large, caution should be taken in handling the measurements.
3.2.2

NaI (Tl) scintillation detector

Because a NaI (Tl) scintillation detector is highly sensitive to gamma rays, it is useful for measuringlow-level dose rates. However, it has a characteristic of showing excessive response in the low-energy
region (around 100 keV) owing to the influence of energy characteristics, and the G(E) function
method is employed as a typical tool for correcting the excessive response. A description of the G(E)function method is given in Explanation B.2.
With the incidence of radiation, the excitation emission (fluorescence) generated in the NaI (Tl) crystal
is converted into photoelectrons by a photomultiplier tube and amplified to output a pulse proportional
to the intensity of the light. Because the output pulse signal is proportional to the energy of theincident radiation, qualitative or quantitative determination of radionuclides can be achieved byanalyzing the gamma-ray spectrum.
The characteristic of the detector is that the absorption reaction of radiation is not directly measured in
the form of ionization or energy, but is indirectly measured by light, and the energy resolution is low
because of a small number of photoelectrons generated owing to the low efficiency of the optoelectron conversion. However, because the electron multiplication factor of a photomultiplier tube isvery high, the signal and noise can be distinguished and separated by a pulse-height discriminator inspite of the very low signal; thus, the radiation detection can be performed with high stability.
3.2.3

Silicon semiconductor detector

A silicon semiconductor detector has an equivalent measurement principle to that of an ionizationchamber detector and is sometimes called a “solid ionization chamber.”
Secondary electrons are generated as a result of the interaction between incident radiation and silicon
crystals. By the ionization of secondary electrons, an electron–hole pair is generated, each of which iscollected on the anode and the cathode, respectively, by applying a high voltage; then, a current flows.
A signal is output as a pulse, and the dose rate is calculated by multiplying with a conversion factor.
The characteristics of a silicon semiconductor detector, including the energy characteristics, aredescribed in Explanation B.3.
Further, the sensitivity of a silicon semiconductor detector is inferior to that of a NaI (Tl) scintillation
detector. This feature of low sensitivity is exploited in a detector used for high-dose-rate regions.

73.2.4

CsI (Tl) scintillation detector

A CsI (Tl) scintillation detector has the same measurement principles as the NaI (Tl) scintillation
detector. This detector has a feature of high sensitivity to gamma rays, as with the NaI (Tl) scintillation
detector. Generally, a scintillation detector is composed of a scintillator, which is phosphor, and aphotodetector, etc. A typical photodetector is a photomultiplier tube. However, the fluorescencepulselength differs depending on the type of scintillator, and the maximum-sensitivity pulselength also
differs depending on the type of a photodetector it is combined with. At present, with the developmentin electronic technology, an optimum combination is achieved for commercialization, considering
these features.
3.3

Recording of measurements

The recording of the measured values in each monitor can be performed in a digital format, in which
an electronic file is stored in the measuring device itself, or in a format in which an analog output iscommitted to a chart paper or similar. In many cases, data are sent to a “telemeter system” at a certaintransmission interval and are also recorded in the system.
3.4

Telemeter system

A telemeter system collects data from each continuous monitor and performs storage, analysis, anddistribution in a centralized manner. An example of the functions of a telemeter system is shown
below:(1) Data collection[1] Continuous monitor (an installed measurement station) * Weather data can also be collected, if
necessary[2] Transportable monitoring post[3] Monitoring car[4] Simplified electronic dosimeter[5] Continuous monitors provided by affiliates (different monitoring agencies)[6] Neighboring municipalities(2) Transmission method, etc.
An appropriate transmission method should be selected from wired, satellite, and analog wirelesslines, according to the installation environment. In addition, server multiplexing or multisite
integration should be implemented in the system.
(3) Data distribution
If there is an information sharing system among organizations related to the monitoring institution,
data are distributed to the relevant systems. Data are also distributed to neighboring municipalities. In
addition, because information sharing and distribution to related organizations may be required fromtime to time, the specifications should be applicable to various situations.
(4) Data display
Capabilities should be provided for uploading the collected data to the home page for public use inreal time and outputting the data to display devices installed in relevant places. As with “(3) Data
distribution,” the specifications should be applicable to various situations.
(5) Filing and storage of data
Capabilities should be provided for storing data in a database for a long time and transferring the past8

data. Because storage of information is essential despite its huge volume, memory expansion should
also be achievable, considering an increase in storage capacity.
(6) Monitoring and analysis of data
Capabilities should be provided for enabling real-time monitoring by a GIS*1 map and outputting ofvarious forms, graphs, and reports. In addition, a capability of performing a real-time spectral
analysis should be provided. Note that average, maximum, and minimum values should be arbitrarily
set and output, because these values contribute to extracting any abnormal values that may becontained in various forms.
(7) Notification function
When an abnormal value is detected, a notification function by telephone, e-mail, or alarm lamp will
enable an early response.
(8) Collection interval
The minimum interval of collecting dose rate data and meteorological data shall be 2 min or less.
However, the interval shall be possible to be set to 10 min, 1 h, etc., at any time during normal
operation. The interval between collections of spectrum data should be set to approximately 10 min
or more, depending on the type and size of the detector.
In addition, it is necessary to arbitrarily set an optimum collection interval while considering theamount of data and power consumption, depending on whether the situation is normal operation or incase of an emergency.

*1

This stands for the Geographic Information System.
9

Chapter 4

4.1

Measuring system using a continuous monitor

Outline

The detectors described in Chapter 3 are used in a variety of systems as continuous monitors. Eachdetector is unique, with different advantages and disadvantages. The detectors utilized in each system
are selected so as to exploit their advantages.
This chapter presents examples of how the detectors described in Chapter 3 are exploited in eachsystem.
Note that the transportable monitoring posts described in Section 4.3 have been installed and operated
in the same manner as the installed continuous monitors since the Fukushima Dai-ichi NPS accident.
4.2
4.2.1

Installed continuous monitor
Ionization chamber monitor

The pressurized ionization chamber described in Item 3.2.1 is used as an installed continuous monitor,
mainly for the measurement in high-dose-rate regions. The specifications of ionization chambermonitors vary slightly from manufacturer to manufacturer, but they are nearly identical. An ionizationchamber monitor consists of a detector, thermal insulation cover, temperature controller, andmeasuring section. The factors that affect the measured values, such as energy characteristics, aremainly the wall materials and their thickness, as well as the species of filler gases and their pressure.
The measurement principle is that charged particles incident on the ionization chamber and charged
particles generated secondarily ionize gas molecules as they pass through the gas in the ionizationchamber, and a current generated by a potential difference between the generated electrons and ion
pairs is detected. The current value is converted into an irradiation dose (C/kg or R), and thenconverted into an air kerma (kinetic energy released per unit mass), air absorbed dose, ambient dose
equivalent, etc., to calculate the dose rate.
Therefore, in terms of the structure, it is necessary to consider that the wall thickness of the ionizationchamber is sufficient to establish the charged particle equilibrium, the strength is sufficient, and nolarge attenuation of incident gamma rays is identified.
In conducting the measurement of environmental gamma-ray dose rates, the contribution of selfcontamination attributed to the raw materials used for manufacturing the detector, may have asignificant impact on the measured values, depending on the quality of the raw materials. In terms of
the presence of self-contamination, the user should understand the contribution of self-contaminationin advance by asking the manufacturer or a third-party organization when the equipment is delivered.
The shapes of ionization chamber dosimeters used as continuous monitors are mainly spherical. This isintended to mitigate the influence of directional dependence on the measured values.
Table 4.1 illustrates examples of the specifications of an ionization chamber monitor.

10

Table 4.1

Specification examples of an ionization chamber monitor
Specification examples of a pressurized ionization chamber

Wall materials,
thickness,
capacity, andshape ofionizationchamber

2 mm of stainlesssteel, 1.2 mm of iron,
14 L, spherical

3 mm of aluminum,
14 L, spherical

2 mm ofaluminum, 14 L,
spherical

3 mm ofaluminum, 14 L,
spherical

Species of filler
gases and pressure

Argon or (and)nitrogen at 4
atmospheres

Argon or nitrogen at 4
atmospheres

Argon andnitrogen at 8
atmospheres

Argon at 6
atmospheres

Energy region of
detected gamma
rays

60 keV–∞

Measurable dose
rate range

10 nGy/h – 100 mGy/h

Temperature
control system

Heating or blasting type

Materials andthickness of adetector insulationcover

3 mm of resin

Energycharacteristics

approximately
80 keV – 3 MeV ±30%

Dose rate linearity

4.2.2

Heating type

1.2 mm of aluminumor 3 mm of resin

1 mm ofaluminum

1.5 mm of resin

approximately 80 keV – 3 MeV ± 10%

approximately 0.4 μGy/h – 1 mGy/h ± 5%

NaI monitor

This is mainly used for the measurement of the low-dose rate region; in the past, the DBM method*1
was widely used, in most cases, for automating the dose conversion by the G(E) function method,using electronic circuits. The DBM method*1 has a feature that the sensitivity is lowered owing tocounting loss that may occur at a high counting rate, because a long time is required to perform
operations in the circuit. Consequently, the effect of counting loss is currently suppressed byautomatically digitally processing the G(E) function method.
The widely used NaI (Tl) scintillation detectors installed in installed continuous monitors mostcommonly have cylindrical shapes of 2 in. diam × 2 in. and 3 in. diam × 3 in., with a feature of
producing an maximum excessive response of approximately 10% when the incident angle of a photon
is approximately 30°–120°.
Table 4.2 illustrates examples of the specifications of a NaI monitor.

*1

This stands for the Discrimination Bias Modulation method.

11

Table 4.2

Specification examples of a NaI Monitor
Specifications of a NaI Monitor

Dimensions andand shape of detector

2 in. diam × 2 in.,cylindrical

Energy compensation• Digital processing (the G(E) function method)• DBM method

Temperaturecompensation circuit
Temperature
controller

Materials of theinsulation cover

3 in. diam × 3 in.,cylindrical

3 in. diam, sphericalYes• Heating type
• Blasting type
• 3.0 mm of resin• 2.5 mm of resin• 1.5 mm of resin• 2.0 mm of aluminum and 2.0 mm of resin• 1.2 mm of aluminum• 1.0 mm of aluminum

LLD

50 keV

ULD

3000 keV

Discri.
Measurable dose rate
range
Materials for the NaI
crystal optics window
Materials for thephotomultiplier tube
window

10 nGy/h –- 10 μGy/h• Borosilicate glass• Low-K glass• Quartz• Borosilicate glass• Low-K glass• Quartz* Discri.: Abbreviation for discriminator (a pulse height discriminator)

4.2.3

Simplified electronic dosimeter

A simplified electronic dosimeter originally worked as an personal exposure dosimeter (an integrating
dosimeter) and has been extensively used. Most simplified electronic dosimeters are equipped with asilicon semiconductor detector. As specified in the Nuclear Emergency Preparedness Guideline issued
by the Nuclear Regulation Authority after the Fukushima Dai-ichi NPS accident, environmental
radiation monitoring is required in a wide range of up to approximately 30 km from nuclear facilities,
as a UPZ, and this type of dosimeter has been used as a continuous monitor in case of an emergency.
In general, the dose rate calculation method converts the collected pulses into a dose rate using theconversion factor obtained at 662 keV of 137Cs. Note that this approach does not allow for nuclide
identification and compensation of energy characteristics, owing to the loss of energy information.
Table 4.3 illustrates examples of the specifications of a simplified electronic dosimeter.

12

Table 4.3

Specification examples of a simplified electronic dosimeter (Reference 4)Specifications of a simplified electronic dosimeter

Detector
Measured dose rate
range
Measured energy
range
Measuring timeWorking temperatureenvironment
Operating humidity
environment
Energy characteristics(137Cs as the standard)
Dose ratecharacteristic(137Cs as the standard)
Measurement
accuracy(137Cs as the standard)
Directional
characteristics(137Cs as the standard)
Temperature
characteristic
Power source

4.3

Silicon semiconductor detector
0.2 μSv/h – 10 mSv/h (the ambient dose equivalent rate)
60 keV – 1.5 MeV, gamma (X) rays
2 min or less (in case of an emergency) to 10 min. (during
normal operation)±20%, −10 °C to +40 °C±20%, 40%–90% RH, 35 °C as the standard−50%–30%, 60 keV to below 100 keV±30%, 100 keV – 1.5 MeV±20%, 0.2 μSv/h – 10 mSv/h±10%, 1μSv – 999.99 mSv* Only when measurement is carried out by the main body ofthe electronic dosimeter, using the integration method.
±30%, 0° ± 60°±20%, * Within the working temperature range, +20 °C as thestandard• Primary power source: 100 VAC• Secondary power source: Battery (must be operational for 1
week or more)* Photovoltaic power generation facilities can be installed.

Transportable continuous monitor

A transportable continuous monitor is a device to be used as a continuous monitor by temporarily
fixing transportable monitoring posts, etc. that can be moved and transported without being installed.
Transportable monitoring posts are intended to be used in case of a nuclear disaster. The measurable
dose rate range is supposed to be from a background level to hundreds of mGy/h (hundreds of mSv/h);these monitoring posts are intended to be used in areas with no installed continuous monitors installed,
or used as a substitute for a measuring instrument with which it is impossible to take measurements forsome reason. After the Fukushima Dai-ichi NPS accident, they have been utilized as continuous
monitors, even when monitoring during normal operation, by equipping them with an external powersource and a transmission system.
4.3.1

Transportable monitoring post

(1) NaI (Tl) scintillation detector
The G(E) function method or the DBM method is adopted for dose calculation.
Some of the typical devices use a NaI (Tl) scintillation detector to take measurements in the lowdose-rate regions, and use other types of detectors, such as a silicon semiconductor detector, to takemeasurements in the high-dose-rate regions. This is a specification to compensate for the13

disadvantage of the NaI (Tl) scintillation detector that dose rates cannot be accurately evaluatedowing to the effects of pile-up or counting loss when a high dose rate is detected. However, othertypes of devices can take measurements over the entire range of measurable dose rates using only a
NaI (Tl) scintillation detector. In this case, a dose calculation method such as the G(E) function
method is adopted in the low-dose-rate region. In the high-dose-rate region, the dose calculation is
performed based on the current value instead of the pulse count processing, such that themeasurements up to the high-dose-rate regions can be taken without being affected by counting loss.
However, note that energy information is lost in this case; thus, energy characteristics cannot becompensated by means of nuclide identification, a G(E) function, or a DBM circuit.
(2) Silicon semiconductor detector
As explained in (1), there are some devices that are equipped with a NaI (Tl) scintillation detector forthe low-dose-rate region and a silicon semiconductor detector for the high-dose-rate region. There arealso other devices that are equipped with only a silicon semiconductor detector. Because such devices
are lighter and more power-efficient than the NaI (Tl) scintillation detector, they are designed toenable also taking measurements from the low-dose-rate region by increasing the number of elements
and improving the detection efficiency.
The dose calculation method is performed in the same manner as for a simplified electronic
dosimeter.
Table 4.4 illustrates examples of specifications of a transportable monitoring post.

14

Table 4.4 Specification examples of a transportable monitoring post
Type A

Type BLow doseHigh doseType CLow doseHigh dose

Low doseHigh dose

Detector

NaI (Tl)scintillation
detector

Siliconsemiconductor
detector

NaI (Tl) scintillation
detector

Silicon semiconductor
detector

Dimensionsof detector

2 in. diam × 2
in. cylindrical-

2 in. diam × 2 in. cylindrical-

Measured
dose rate
range

BG–11 μGy/h

9 μGy/h – 100
mGy/h

0.01–500μGy/h

300 μGy/h –100 mGy/h

BG–100 mGy/h

Measuredenergyrange

50 keV – 3
MeV

50 keV–∞

50 keV – 3
MeV

50 keV–∞

50 keV–∞Dose
calculation
methodG(E) function
method bydigitalprocessing

Pulse
conversion
systemDBM

Current
measuringsystemPulse conversion system

Collection
interval

1 min or 10 min

1 min or 10 min

1 min or 10 min

Timeconstant

10 s, 100 s

10 s, 60 s, 100 s

Auto toggle

RecordingsystemElectronic memory system

Electronic memory system

Electronic memory system

100 VAC and built-in battery* Solar power generation withexternal batteries is optionallyavailable

100 VAC and built-in
battery* Solar power generation
with external batteries is
optionally available

100 VAC and built-in
battery* Solar power generation
with external batteries is
optionally available

approximately 345 × 348 ×
205 mm3

approximately 670 × 440 ×
450 mm3

approximately 600 × 480 ×
250 mm3

approximately 10 kg

approximately 15 kg

approximately 30 kg

PowersourceDimensions
Weight4.4

Mobile survey system

A mobile survey system can efficiently monitor a wide range of areas by exploiting its mobility.
Mobile survey systems are classified into two types: an installed type in which a detector is installed toa vehicle called a monitoring car, and a portable type in which the system including a detector can beeasily removed and installed, even in a different vehicle. The following are the detectors used in the
installed type and the portable type.
Note that, in a special usage of a mobile survey, the specifications (the measurement interval, time15

constant, directional characteristics, etc.) suitable for the mobile survey shall be identified and utilized.
For more information, see Explanation F.
4.4.1

Installed type(1) NaI (Tl) scintillation detector
The type of devices commonly used are those with the same specifications as the installed continuous
monitors described above. The typical shape of the detector is also cylindrical, with dimensions 2 in.
diam × 2in.
(2) Silicon semiconductor detector
This is used for measuring the high-dose-rate areas, as with transportable monitoring posts.
The specifications are the same as those of transportable monitoring posts.
4.4.2

Portable type

Currently, CsI (Tl) scintillation detectors are used as one of the representative types of transportablemobile survey systems. The complication of installing this type of device is eliminated by exploiting
its features of light weight and small size, and housing the unit, including a transmission system, in asmall, robust, and dedicated case.
Tables 4.5 and 4.6 illustrate examples of specifications of the installed type and the portable type.

16

Table 4.5 Specification examples of a mobile survey system (Installed type)Specifications of a mobile survey system (Installed type)

Detector

Low doseHigh dose

2 in. diam x 2 in. cylindrical
NaI (Tl) scintillation detector

Silicon semiconductor detector

10 nGy/h – 10 μGy/h

10 μGy/h – 100 mGy/h

50 keV – 3 MeV

50 keV–∞

G(E) function method

Pulse conversion

Measured dose rate
range
Measured energy
range
Calculation method
for energy
compensated dose
rates
Housing for thedetector

Detector cover, including a mounting table for the source
calibration jig

Measuring time

1–99999 s, arbitrary setting is allowed

Temperature control

Circulating the vehicle interior air

Data collection
system

(1)
(2)
(3)
(4)
(5)

Measures against
vibration

Installation of a vibration isolator

Power source

Equipped with a generator for measuring instruments* Output is approximately 4500 W.

Distribution panel

A breaker is installed by separating the power sources for themain breaker and for each measuring system.

PC for data collection
Data collection and processing software
Printer
Communications system for data transmission
GPS

17

Table 4.6 Specification examples of a mobile survey system (portable type)Specifications of a mobile survey system (portable type)
DetectorCsI (Tl) scintillation detector
13 mm × 13 mm × 20 mm

Photo detection
element

Multi-pixel semiconductor

Measured dose rate
range

1 nSv/h – 300 μSv/h

Measured energy
range

30 keV – 2 MeVCounting efficiency

40 cpm or more at 0.01 μSv/h of 137Cs

Energy resolution

8 % 137Cs 662 keV

Calculation method
for energy
compensated dose
rates

G(E) function method

Housing for thedetector

Aluminum storage box
175 mm × 345 mm × 240 mm
With an exhaust fan for heat removal

Measuring time

1–60 s, arbitrary setting is allowed

Operating temperature−10 °C to 50 °C
range

Data collection
system

(1)
(2)
(3)
(4)

Embedded computer
Data collection and processing software
Communications system for data transmission
GPS

Power source

12 V / 24 VDC power source* Power supply shall be allowed from an automobile accessory
socket and an accessory power source.

18

Chapter 5

Installation of a measuring system

5.1 General considerations of the installation of a continuous monitor
The following points should be considered when installing a continuous monitor.
(1) Select a location for installing a monitor considering the site of the nuclear facility, thedistribution of wind direction, the distribution of towns and villages, topography, etc.
(2) Select a location where the surroundings are open, which is easily accessible, and where thedrainage is good. In addition, from the viewpoint of drainage, a location with adjacent paddy
fields should be avoided as much as possible. Because paddy fields are filled with water or dried
depending on the seasons, the environmental gamma-ray dose rate might fluctuate. Even in fallow
areas, owing to the characteristics of paddy fields, the water retention rate is high, and the soil
may remain moist for a long time because of rainfall.
(3) Avoid locations in adverse environments with high temperature, high humidity, etc.
(4) Minimize the shielding effect caused by the topography, buildings, and structures around thedetector.
(5) When multiple continuous monitors are used concurrently at a same point, they must be installed
as close as possible, to the extent that they do not interfere with each other, and the conditions
such as the ground height must be uniform.
(6) The location where a continuous monitor is installed should be surrounded with fences so that thedose rate does not fluctuate owing to human effects.
(7) When data transmission is performed using radio waves, sufficient attention must be paid to thesite of the radio station, directivity of the antenna, and preventive measures against the sneak ofradio waves from the power supply circuits and cables, considering the effect of the emitted radio
waves.
(8) In case of a nuclear disaster involving a large-scale natural disaster, the infrastructure such as thepower supply may be destroyed. Assuming such a case, two or more power supply systems
should be provided by installing a private power generator or a solar power generation system
together. However, generators need to secure fuel (light oil, LPG, etc.) that can be continuously
supplied, and solar power generation needs to secure a backup power source during periods of
insufficient sunlight (at night, in bad weather, etc.). In addition, satellite communications, etc.,
which are less affected by disasters, should also be provided, assuming that the means of datatransmission might be cut off.
(9) Unexpected lightning may cause the malfunction of electrical systems of the measuring
equipment through AC power sources. To avoid such a situation, it is desirable to install anarrester or similar.
(10) During the Great East Japan Earthquake, continuous monitors were damaged by the tsunami after
the large-scale earthquake and became unusable; thus, when continuous monitors are installed ata site near the sea, consideration should be given to installing them on a hill not affected by atsunami. However, if a village is located along the coast, it would be unavoidable to install acontinuous monitor at a site near the sea. As a result, if the monitor should become unusable, atransportable continuous monitor is required as a substitute.
(11) In terms of the ground height for installing a detector, it should be considered as a matter common
to all types of detectors and all installation conditions, in addition to considering the purpose ofmeasurement, that a detector should be installed at a height allowing easy maintenance, because
the adjustment of equipment is required, including periodical gain adjustment.
19

5.2 Installed continuous monitor
5.2.1
Ionization chamber monitor
Ionization chamber monitors, which are used as installed continuous monitors, have certain points tobe considered specific to them, other than those described in Section 5.1. The following are theconsiderations specific to an ionization chamber monitor.
(1) An atmospheric pressure-type ionization chamber should be installed in an instrument shelter orsimilar, to avoid direct sunlight.
(2) If a monitor that was installed fails to measure, a transportable monitoring post must be used forcompensation.
(3) The ground height for installing an ionization chamber monitor should be decided by selecting an
installation environment suitable for each purpose, with clear intention.
[1] In an alley, at a height of 1 m above the groundA height of 1 m above the ground is the basic ground height for measurement that contributes to theevaluation of exposure dose to the human body. This mainly intends to measure gamma rays fromradioactive materials deposited on the ground surface.
[2] In an alley, at a height of approximately 1.5–1.8 m above the ground
This is a ground height with the same purpose as the height of 1 m above the ground, as mentionedin [1]. However, in heavy snow areas, it is assumed that detectors will be buried in the snow. As aresult, this ground height is set for ensuring effective measurement, even in winter.
[3] On a monitoring building
A monitor is installed at a height of 0.6–1.0 m above the roof of a monitoring building (at a heightof approximately 3 m above the ground). The height above the monitoring building is selected to
avoid gamma rays from radioactive materials floating in the sky being shielded, which occurs whena monitor is installed in an alley beside the building. Installing a monitor on the building has theadvantages that the area occupied can be reduced and the monitor can sensitively respond toradioactive plumes floating in the sky. However, note that the measurements include thecontribution of radioactive materials contained in the roof materials of the building and a wide
range of areas in the alley.
[4] At a height of dozens of meters above ground
Among the continuous monitors installed so far, some have been installed on the roofs of buildingsof height, for example, three to five stories above the ground, in order to take average measurements
over a wide range of areas, especially for measuring fallouts. This has the advantage of being
sensitive to radioactive plumes, as with the case of installing on a monitoring building, as
mentioned in [3].
Figures 5.1 through 5.4 show examples of the installation of an installed continuous monitor undervarious conditions.
5.2.2
NaI monitor
The NaI monitor, which is used as an installed continuous monitor, has the same points to beconsidered as those mentioned in Items (2) to (3) of Section 5.2.1, in addition to those of Section 5.1.

20

Figure 5.1

Example of installation on an alley, at a height of 1 m above the ground(Aichi Prefecture)

Figure 5.2

Example of installation on an alley, at a height of 1.5–1.8 m above the ground(Aomori Prefecture)

21

Figure 5.3

Example of installation on a monitoring building (Ibaraki Prefecture)

Figure 5.4

Example of installation at a height of dozens of meters above the ground(Aichi Prefecture)22

5.2.3
Simplified electronic dosimeter
A simplified electronic dosimeter, which is used as an installed continuous monitor, has specific points
that must be considered, in addition to those mentioned in Section 5.1. The following are the points
specific to a simplified electronic dosimeter to be considered.
Simplified electronic dosimeters have been introduced to cover areas where monitoring is notconducted, because the monitoring scope has been expanded based on the experience acquired after
the Fukushima Dai-ichi NPS accident.
(1) Rack
Use a robust and durable material such as stainless steel. However, because it acts as a scatterer andshield, consideration must be given to making it as thin as possible. The rack is equipped with adetector storage box, transmission equipment, data logger, battery storage box, and photovoltaic
power generator. A storage box containing a detector is installed at the front of the rack so that theeffective center of the detector is positioned at a height of 1 m above ground. Therefore, the battery
storage box should not be installed on the front side of the rack, so as not to act as a shield.
Photovoltaic power generation equipment must be installed so as not to cover the top of the detector.
(2) Data transmission
Generally, the time series measured data taken during a short period of time must be stored in themeasuring system, and it must be separately output as a data file. However, to be used as acontinuous monitor, data should be collected in real time, as with the installed continuous monitor as
used in an ionization chamber monitor and a NaI monitor, such that the measurements taken can be
confirmed. As a result, if a data transmission system is not provided, additional equipment isrequired. Examples of a data transmission system are shown below:• Mobile phone lines• Satellite mobile phone lines
In using a mobile phone line, if the installation status of the radio wave base station and the coveragefor communications were understood in advance, data transmission would be ensured in a relatively
stable radio wave condition. Because each provider has its own base station for mobile phone lines,
the radio wave condition depends on the installation environment. In the current situation of
infrastructure, where mobile phones are widely used, a relatively stable radio wave condition can beachieved in densely populated areas such as urban areas, regardless of the provider selected.
However, in sparsely populated areas such as mountainous areas, you may not be able to acquire
stable radio wave conditions, depending on the providers. Additionally, you may not be able to use
your mobile phone line, even when you select any provider.
In such a situation, a satellite mobile phone line that can be used even in case of an emergency,
regardless of the existence of a radio wave base station, should be equipped or installed together as
an alternative.
Because satellite mobile phone lines are expensive, compared with mobile phone lines, the datatransmission method should be selected based on the radio wave condition of each line.
(3) Driving power source
A simplified electronic dosimeter can usually be driven by providing an external AC power source. A
model equipped with built-in batteries can be driven within a time period corresponding to theircapacity. However, if you want to use it as a continuous monitor, you need equipment that cancontinuously supply power for a certain period of time. Because a power supply is also required in an
attached system such as a data transmission system, it is difficult to supply power only with the builtin battery. Consequently, the following power supply methods should be used in combination.

23

[1] AC power source
An AC power source is advantageous in that it can stably supply power. However, it may not beeasy to secure AC power source in some cases. In addition, even in a situation where AC powersource can be used, when power is supplied from an outdoor power outlet used in a generalhousehold, it is desirable to install an arrester, such as those described in Section 5.1.
[2] External battery
An external battery should be provided as a standby power source so that it can be used if an AC
power or other supply power should be unavailable. Batteries currently in use, such as shipbatteries, are of a large capacity and some of them have a maximum driving capacity of
approximately 7 days by connecting two batteries in series. Even if an external battery becomes
exhausted, a function to charge it when the other power supply is restored should be provided. Since
an external battery is physically large, consideration should be given to its positioning, such as
placing it directly under the measuring instrument body, so that it does not become a shield for
gamma rays to be measured.
[3] Photovoltaic power generation
Photovoltaic power generation is a method that ensures a relatively stable power supply over a long
period of time even, in areas where no AC power supply is available, such as in mountainous areas
with no power cables, or in situations where the infrastructure is unusable owing to power outages,
etc. When this method is employed, sunlight (ideally direct sunlight) must be secured. In terms of
measurement, an open environment is desirable for installation. However, even if there were no
objects to block the sunlight at the time of installation, the situation may change depending on thegrowth of trees and construction of new buildings. When installing a generator, attention should be
paid to the period of installation and prediction of changes that may occur in the surrounding
conditions.
Figure 5.5 shows an example of the installation of a simplified electronic dosimeter.

Figure 5.5

Example of the installation status of a simplified electronic dosimeter(Shimane Prefecture)24

5.3 Transportable continuous monitor
Based on the fact that, after the Fukushima Dai-ichi NPS accident, transportable monitoring posts
among transportable continuous monitors have been installed and operated in the same manner as
installed continuous monitors, the points to be considered in installing them are summarized asfollows:
5.3.1
Transportable monitoring post
A transportable monitoring post, which is used as a transportable continuous monitor, has points to beconsidered specific to them, apart from those described in Section 5.1. The following are the points tobe considered specific to transportable monitoring posts:(1) Rack
The transportable monitoring posts currently on the market are structured to be installed at a height of
1 m above the ground by installing a rack or similar at the lower part of the equipment. Because theattached rack is simple, a more robust rack is required when the monitor is used as a continuous
monitor for a long period. However, if an excessively robust rack is installed in the lower part of themeasuring instrument or in its vicinity, it would act as a scatterer and shield of gamma rays fromradioactive materials deposited on the ground surface. Therefore, it must be made as thin as possible.
When selecting a rack, the influence of the shielding effect should be considered; it is also importantto understand the degree of the shielding effect of the rack by comparing with the measured values
taken before installation and after the whole system is installed, including the rack.
(2) Data transmission
This is performed in the same manner as a simplified electronic dosimeter.
(3) Driving power sourceThis is performed in the same manner as a simplified electronic dosimeter.
When a photovoltaic power generation facility is employed, the measuring instrument itself would
receive direct sunlight by giving priority to the sunshine condition. Although a transportable
monitoring post is equipped with a temperature compensation circuit, it is rarely equipped with atemperature control device, unlike an installed continuous monitor. As a result, the response may
significantly vary owing to the failure of the temperature compensation circuit to respond to a sudden
temperature change. To secure a stable power supply for a long term, solar power generation
facilities should be effectively exploited, and sunlight is essential. However, this is not a desirable
condition for the measuring instrument.
Consequently, it is important to understand that there may be a certain effect of temperature changes
as a variation factor of the measured values. Details are given in Explanation E.3.
Items apart from those above may be considered to be used in the same way as with the installed
continuous monitor. Figure 5.6 shows an example of the installation status of a transportable
monitoring post.

25

Figure 5.6

Example of the installation status of a transportable monitoring post(the positioning of the detector: 1 m above the ground, with solar power generation and external
battery facilities)

5.4

Mobile survey system

The system used for mobile surveys has some specific points to be considered. The following points
are specific to the mobile survey system:
In this section, the points to be considered are described by giving a typical monitoring car as an
example of the installed type and giving a measuring device as an example for the portable type,which was developed after the Fukushima Dai-ichi NPS accident and has been utilized since then.
(1) Installed type
An installed mobile survey system refers to a so-called “monitoring car.” Therefore, the detector isclassified into two types: an exterior type to be installed above the vehicle, and an interior type to beinstalled at a predetermined position in the vehicle, so as to be installed at a height of 1 m above theground.
Figure 5.7 shows a typical example of the external installation.
The detectors outside the vehicle are often installed at a height of approximately 2 m above theground.
The power supply should be secured when installing measuring equipment. A typical method is tosupply power from a vehicle battery, but in this case, the battery capacity needs to be larger than thatused only for the vehicle. In addition, because the power supply voltage is unstable when the battery
is being charged from the power generation function (dynamo) of the vehicle, the battery is notsuitable as is, as the power source for a precision instrument. In such a case, an uninterruptible power
supply (UPS) or similar should be connected to ensure a stable power supply voltage.
The interior-type detector is installed at a height of approximately 1 m the above ground in thevehicle.
Because the detector is installed in the vehicle, the shielding effect of peripheral equipment and thevehicle body itself should be evaluated. The geometrical conditions of the detector and its peripheral
equipment must always be the same to ensure that the influence of the shielding effect is constant,and objects that can act as shields should not be loaded while taking measurements. The mattersconcerning the power supply are the same as those of the exterior-type detector mentioned above.
26

Figure 5.7

Example of an installed type (exterior-type) (Reference 5)(2) Portable type
The portable travelling survey system has widely been in use after the Fukushima Dai-ichi NPS
accident, and it is expected that more monitoring organizations will consider introducing it in the
future.
A typical example of the portable type is shown in Figure 5.8.
Detectors used in the portable type include a system in which a commercial survey meter is
temporarily installed in the vehicle, and a system in which it is stored in a special container for the
purpose of mobile surveys and which can be loaded on a variety of vehicles. In both systems, thematters concerning the installation are based on the following concept:
The system, including the detector, should be installed and installed at any position in the vehicle. In
order to use the measurements taken in the vehicle as the dose rate outside the vehicle, it is necessary
to use the correction coefficient of both the outside and inside of the vehicle in accordance with the
installation status, and the detector should be installed while taking measurements, to avoid changing
of its position. When a commercial survey meter is used, the power supply source is should be usedin accordance with the specifications of the survey meter. Generally, it is driven by dry batteries. In
addition, for systems designed for mobile surveys, some mobile survey systems are currently
powered from the vehicle battery by connecting to the vehicle’s accessory socket.

Figure 5.8

Example of a portable type (left: when installed; right: inside of the storage case)(Reference 6)27Chapter 6
6.1

Measurement and calibration

Precautions for use

A continuous monitor has a complex mechanism, is left unattended for long periods in harsh outdoor
environments, and is furthermore required to minimize missing measurements. To satisfy these strict
conditions, a well-tested product should be used, and daily thorough maintenance and management isalso required.
Of the many items to be considered when using a continuous monitor, the following are the major
points:(1) Effectively understand the characteristics of the monitor to be used (energy characteristics,
directional characteristics, temperature characteristics, etc.), and acquire sufficient information on
the structure of the detection unit and the functions of the electronic circuit from the
manufacturer, if possible. In the event that the performance of the monitor is seems inadequate forthe purpose of monitoring, appropriate actions shall be taken.
(2) Handle the monitor correctly according to the instructions.
(3) Conduct inspections and calibrations at regular intervals, as necessary.
(4) Because the detection unit is generally very vulnerable to impacts, extreme care should be taken
when handling.
(5) As the detection unit is supplied with a high voltage, the high-voltage terminals and similar
should be periodically cleaned, because they may cause electrical leakage owing to moisture anddust adhesion, thereby causing noise. If a desiccant is used, replace it periodically.
(6) For a model containing an arrester (a lightning arrester) to prevent accidents caused by lightning
strikes, regular inspections should be conducted, especially in areas where lightning strikes occur
frequently.
(7) The measurement results of a continuous monitor include various elements such as theinstallation status and measurement condition of the monitor; therefore, clarifying these elements
together with the measurement results will give us the information contained in the measurement.
Examples of recording formats for installed and transportable continuous monitors are illustrated
in Table 6.1.
In addition to the items common to the monitors mentioned above, the following items are to be notedwith respect to the monitors for each method.
[1] Ionization chamber monitor
The ionization chamber monitor periodically checks for zero points (some are automated).
Additionally, pay attention to pressure changes due to the breakage of the sealing. Changes inpressure can be detected by a significant decrease in the readings in a pressurized ionizationchamber, and by a dependence on atmospheric pressure in an atmospheric pressure chamber.
[2] NaI monitor
Because a NaI monitor can generally exhibit certain temperature characteristics, attention should be
paid to diurnal and seasonal changes in the indications, and appropriate actions should be taken if
necessary.
[3] Simplified electronic dosimeter
Attention must be paid to temperature characteristics. Many of the simplified electronic dosimeters
also have a silicon semiconductor detector mounted. In this case, it should also be noted that theeffect of the energy characteristics is larger than that of other detectors.
28

[4] Transportable monitoring post
This has temperature characteristics depending on the type of the measuring instrument installed, aswith a NaI monitor and simplified electronic dosimeter. The effect of the characteristics on themeasured values is described in Explanation E.3. Owing to the nature of the transportable
monitoring post, this is a simplified device that is not equipped with a temperature control device.
Therefore, a variation factor due to temperature changes is included as a representative factor formeasured values.
[5] Mobile survey system
Unlike a transportable monitoring post and others, because the detector and the ambient temperatureof the detector, which affect the measurements, are controlled to some extent by circulating air inthe vehicle, they do not greatly affect the measured values. Because of the nature of the survey
purpose using a mobile survey system, the coverage for measurement or the directional
characteristics accounts for an important variation factor. It is important to carry out themeasurement with the understanding of the directional characteristics, including the vehicle that is
the means of travel.
In addition, it can be easily imagined that a mobile survey system is utilized in the middle of an
emergency. In such a case, if the vehicle is used for monitoring, the tire house of the vehicle and the
air filter may become contaminated, and the background may increase. To reduce this effect, take ameasurement at a same point where the dose rate does not vary before starting measurement andafter finishing it, obtain the contamination level of the vehicle, and conduct an evaluation by
subtracting it from the measured value, as required.

29

Table 6.1 Example of the recording format for installed and transportable continuous monitors
Report on the survey results of the environmental radiation dose rate by continuous
monitoring
Survey year

Address
Measurement
location

Latitude::Research laboratory in charge

Longitude::

Chief investigator

Name of the point
Identification of the monitor

Contact
NaI monitor, ionization chamber monitor, others()

Height above ground (m)cosmic rays (with / without)With / without

Environmental dose rate
Date of measurement

Weather
Average valueStandard deviation

Unit at the time of measurementnGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)nGy/h (air absorbed dose rate)
nSv/h (ambient dose
equivalent rate)30

Measuringtime

Counts ofdata

Remarks

6.2
6.2.1

Installed continuous monitor
Calibration and adjustment“Adjustment” refers to the act of changing various settings appropriately to keep the equipment
healthy, and as a result, JIS specifies that a measured value is acceptable if it is within the allowable
range of the “relative reference error ± 20%.”On the contrary, “calibration” is the act of providing a calibration constant necessary to correct thedifference between a measured value and a certified value from the ratio between the measured value
and the certified value by securing traceability with national standards.
As mentioned above, adjustment and calibration are different acts, and adjustment does not substitutefor calibration. However, because it is difficult to conduct calibration on a continuous monitor that isalready installed, adjustment is performed instead, in practice.
As a requirement for the measured values of a continuous monitor, if it is treated as a “monitor,” theintegrity for satisfying this requirement is ensured by performing adjustment. However, if the data
contributing to external exposure dose assessment are supposed to be obtained, calibration should be
conducted, by which traceability with national standards can be secured. Reference gamma-ray
sources used for calibration work are described in Explanation C.
(1) Calibration
At the place where the monitor is installed, it is difficult to perform a calibration matching the one
conducted at the irradiation field, or conduct a test that corresponds to the acceptance test by the
manufacturer. However, at the place where the monitor is installed, it is possible to conduct acalibration by using a standard with clear traceability by national standards, or conduct a practicalcalibration by employing the source method using the reference gamma-ray source stipulated in JIS Z
4511: 2005, or the inverse square law.
For the reference gamma-ray source, the value of the irradiation dose rate at a point 1 m from anormal source is certified. If the line connecting the source and the detector center is parallel to theground surface or to the floor surface, the lower the height of the source and the detector from theground surface, and the larger the distance of the detector from the source, the larger will be thecontribution rate of scattered gamma rays, within a certain range. On the contrary the higher theheight and the smaller the distance, the smaller will be the contribution rate (see Explanation D).
However, when the distance becomes very short, the change in the effective center cannot be ignored.
Major calibration conditions and methods are described below:[1] The nuclides for the reference gamma-ray source used for monitor calibration shall be 60Co or
137
Cs. However, 226Ra may also be permitted.
[2] The calibration direction shall be horizontal or vertical to the central axis of the detector.
Further, a direction with less effect from scatterers in the vicinity should be selected.
[3] The source should be installed at a distance equal to the calibration distance of the reference
gamma-ray source above the calibration direction from the effective center of the continuous
monitor detector. In addition, the contribution of scattered rays should be considered inaccordance with the height and distance, and corrections should be made if necessary (seeExplanation D).
[4] The distance between the reference gamma-ray source and the detector is preferably 1 m or
more. Typically, the distance between the source and the detector is 1 m.
[5] Weather conditions that usually affect calibration results should be avoided.
Specifically, weather conditions such as rainfall, in which the background dose rate varies
during a calibration operation, must be avoided.
31

[6] Ensure that the performance of the monitor is sufficiently stable.
[7] The calibration constant K is calculated from the following equation:
K=
where,

̇̇̇̇ : Dose rate of a standard gamma-ray source during calibration

̇ : Indicated value during irradiation with a standard gamma-ray sourcė : Indicated value by dose rate of environmental gamma rays

[8] Calibrations are desirably conducted four times a year or every quarter, considering the changes
in the weather and environment of the four seasons. However, if four times a year is notpracticable, the integrity of the equipment can be ensured by calibrating at as short intervals as
possible using a confirmation source.
[9] The date and time of calibration, calibration conditions, and results shall be recorded.
When calibration is performed at many points, the effect of scattered ray varies owing to thedifference in the surrounding environment at each point. Therefore, monitoring organizations
should preferably conduct a comparative calibration with the monitor to be calibrated at the
installation site, by using a well calibrated dose (rate) meter as the reference measuring
instrument in an environment with little scattering, if possible.
(2) Adjustment
In adjusting a monitor, individual equipment and settings should be checked. Items of inspection andadjustment are shown below:[1] Visual inspection and cleaning of components[2] Insulation resistance[3] Temperature indication accuracy and alarm operation of the heating system[4] Accuracy of power supply voltage[5] Checking of the accuracy of dose rate indication and the scaler operation[6] Confirmation of zero points[7] Gain adjustment[8] Source irradiation test
6.2.2
Ionization chamber monitor(1) Determination of the measuring time
Ionization chamber monitors are often set up for measuring high dose rates. Approximately 2 min issufficient for the integrated count time for one lot of data of the measured values taken at a high doserate, and approximately 2 min is also desirable for identifying the variation in measured values incase of an emergency. During normal operation, a maximum of approximately 10 min is appropriate
as the integrated count time for one lot of data, considering the dose rate level of natural radiation in
Japan.
(2) Data processing[1] Recording
Data recording is performed as described in Section 3.3. The recorded data contain themeasurements for dose rate.

32

[2] Processing
Data processing is performed as described in Section 3.4. Mean, maximum, and minimum values
are calculated and analyzed, and outliers are extracted.
6.2.3

NaI monitor(1) Determination of the measuring time
When a 3 in. diam × 3 in. NaI (Tl) scintillation detector is used, the measuring time of a few minutes
is sufficient to obtain adequate accuracy for the integrated count time for one lot of data of themeasured values, judging from the variation in statistical errors in the count values. However, inpractice the measuring time should be determined by considering the variability.
To distinguish the variations between natural radiation and facility-induced radiation based on the
difference in the cycle of variations, it is appropriate to determine the integrated count time for one
lot of data, in the range of a few minutes to 10 minutes, for the reasons described in Explanation B.
To reproduce the variation pattern of dose rates more faithfully, a method can be employed in which
the integrated time is set to approximately 2 min and a smoothing or averaging operation is
performed in the subsequent data processing.
(2) Data processing[1] Recording
Data recording is performed as described in Section 3.3. The recorded data contain measurements
for dose rates, count values, etc. In the case of a system for obtaining spectral information, spectral
data are also recorded.
[2] Processing
Data processing is performed as described in Section 3.4. Mean, maximum, and minimum values
are calculated and analyzed, and outliers are extracted.
6.2.4

Simplified electronic dosimeter(1) Determination of the measuring time
The concept of the measuring time in a simplified electronic dosimeter is the same as explained in
Subsection 6.2.3.
(2) Data processing[1] Recording
Data recording is performed as described in Section 3.3. The recorded data contain the dose ratemeasurements. If power is supplied by an external battery, monitoring of battery voltage is alsoessential, and this information should also be transmitted and recorded.
[2] Processing
Data processing is performed as described in Section 3.4. Mean, maximum, and minimum values
are calculated and analyzed, and outliers are extracted.
6.3

Transportable continuous monitor

6.3.1

Calibration and adjustment

The concept of calibration and adjustment in a transportable continuous monitor is the same as
explained in Subsection 6.2.1. In the case of a system in which energy information is not available,
adjustment, etc. are performed in the comparison of measured values with certified values for thesource (the irradiation value).
33

6.3.2

Transportable monitoring post

(1) Determination of the measuring time
The concept of the measuring time for a transportable monitoring post is the same as described in
Subsection 6.2.3.
(2) Data processing[1]

Recording

Data recording is performed as described in Section 3.3. The recorded data contain measurements ofdose rates, count values, etc. In the case of a system in which spectral information is obtained,
spectral data are also recorded. If power is supplied by an external battery, monitoring of battery
voltage is also essential, and this information should also be transmitted and recorded.
[2]

Processing

Data processing is performed as described in Section 3.4. Mean, maximum, and minimum values
are calculated and analyzed, and outliers are extracted. However, the transmission interval of data
during normal operation may be set by considering the power consumption and the allowableamount of power provided.
6.4

Mobile survey system

Owing to the nature of a mobile survey, measurements are often taken while moving at all times,
rather than at rest. One way of using a monitoring car is to take measurements at installed points while
stopped; however, because the concept of the installed-point measurement is similar to that of an
installed continuous monitor and a transportable continuous monitor, only the method of taking
measurements while moving is mentioned here.
6.4.1

Calibration and adjustment

If a survey meter or similar is used, periodic calibrations must be conducted as with the case of normal
usage. In addition, the adjustment required to secure the integrity of the measuring instrument isperformed in the same manner as with the equipment used in an installed station, etc.
In the case of a system in which a measuring instrument is installed in a vehicle, the measured value
obtained must be converted into a dose rate of 1 m above the ground outside the vehicle. Examples of
the calculation method for correction factors of both the inside and outside a vehicle are provided in
Chapter 7, “Analysis and evaluation of measurement results.”
6.4.2

Determination of the measuring time

In the case of a mobile survey, the shorter the interval between measuring times, the closer theintervals at which the measured values on the travel route can be obtained. This is because the shorterthe interval between measuring times, the more sensitively the variation in dose rates in the vicinity ofthe travel route can be identified. However, as the interval between measuring times becomes longer,
the average dose rate distribution in the vicinity of the travel route can be obtained more accurately.
In practice, the mobile speed during a mobile survey is set to the speed limit stipulated by law. The
34

relationship between the mobile time and the migration length is described in Explanation F.2.
Because the detection efficiency differs depending on the type of the detector, variation in themeasured values obtained is important information to set the measuring time. Explanation F.2 shows
the variation in the measured values, which is defined in the general measurement specifications used
in a mobile survey system.
6.4.3

Processing of data

With regard to the systems developed as monitoring cars, the measured values are generally checked
inside the vehicle in real time, using a rate meter*1 or similar. These data are accumulated as a digital
file and stored in a data collection system and a PC for control. In addition, with the recent
improvements in the accuracy of a GPS*2 accompanied by the advances in equipment, real-time
mapping has been widely used. The mapping data are collected and managed by a cloud server orsimilar, such that the data can be checked anytime and anywhere, as long as an accessible tool isavailable at hand.
If a commercial survey meter is used, a transportable mobile survey system is also available, in whichdata can be checked anytime and anywhere, with an accessible tool available at hand, by collecting
and managing the mapping data with a cloud server, etc., as with the monitoring car mentioned above,in addition to checking the data with the display of the survey meter.

*1*2

A device that displays the dose rate and counting rate information in analog quantities on the meter.
GPS: Abbreviation for "the Global Positioning System"
35

Chapter 7
7.1

Analysis and evaluation of measurement results

OutlineAs the concept common to all systems, it has been general until now that continuous measurement
records of environmental gamma rays were analyzed by extracting significant variations against themean values. This chapter describes typical analysis and evaluation methods and the concept during
normal operation, as well as the handling of the monitoring data. In addition, during and after the
Fukushima Dai-ichi NPS accident, certain situations have been identified in which this concept does
not hold true; thus, analysis and evaluation methods for emergency and post-accident monitoring arealso described.
7.2

Analysis and evaluation during normal operation

The “Environmental Radiation Monitoring Guideline” (Reference 7), which was provided by theformer Nuclear Safety Commission, states that the purpose of monitoring during normal operation is
to “protect the health and safety of residents living in the vicinity of nuclear facilities, to confirm thatthe environmental exposure dose of residents living in the vicinity of nuclear facilities due toradioactive materials or radiation is sufficiently below the annual dose limit, and to provide the results
to the residents living in the vicinity.” In addition to this purpose, “Grasping the status of accumulationof radioactive materials in the environment,” “Early detection of unexpected releases of radioactive
materials or radiation from nuclear facilities and assessment of their impact on the surrounding
environment,” and “Establishment of a system for environmental radiation monitoring in case of anincident or an emergency” are described. To achieve these objectives, examples of analysis andevaluation methods are presented below:
7.2.1

Calculation of mean values and standard deviation

Continuous measurement records of environmental gamma rays extract and analyze significant
changes against the mean values. Therefore, with regard to the measurement results over a long periodof time, the average and dispersion of the measured values per unit time are obtained. The unit timeshall be set to 1 h and shall be shortened as necessary.
Take an arithmetic average of the measurements in the unit time. When each of the n measurements is
calculated by, the averageis determined by
= ∑
/ .
2
2
The variance σ is calculated by σ = ∑ ( − ) /(n − 1), and the square root σ is set as thestandard deviation.
The average and standard deviation depend on which population is focused upon. The selection of a
population is described in Explanation G.1.
7.2.2

Extraction of outliers

Outliers are identified by determining whether the measured value in the unit time is excessive or
insufficient with respect the average value according to a given standard, considering the distributionof these values. There are many ways to adopt a standard, but in this measurement method, a measured
value exceeding± 3σ is identified as an outlier. As described in Explanation G.2, the probability ofexceeding this value varies depending on the form of the population distribution; for example, theprobability is 0.3% for a normal distribution and 2% for an exponential distribution.
In addition, because many NaI monitors, which are now widely used, can acquire energy information,
it is also effective to extract outliers by using this information.
36

As a specific example, focusing on the fact that most of the artificial radionuclides released in the caseof a nuclear disaster are radionuclides with photon energy in the low-energy region (approximately 80–1400 keV), the ratio between the counting rate and the total dose rate (the total dose rate/the counting
rate in the low-energy region) between these energy regions is always calculated, and if the value islower than the ratio calculated in the normal period, the effect of the artificial radionuclides issometimes observed; however, if the value is higher than the ratio calculated in the normal period, theeffect of natural radionuclides due to rainfall, etc. is observed.
In addition, in a system in which the dose rate calculation is carried out by the G(E) function method
using a NaI monitor, a measure of the increase in the dose rate owing to the contribution of artificial
radionuclides can be understood, by weighting the count values of the energy pulse-height distribution
obtained by the NaI detector according to the G(E) function, in which the conversion coefficientdiffers for each energy in order to compensate for the energy characteristics, and using the “pass rate”
calculated by comparing the count values obtained before and after the weighting operation.
It is possible to confirm that the existing monitor is operating satisfactorily and obtaining the correct
measured values through a comparative measurement using a portable measuring instrument (NaI (Tl)scintillation spectrometer, etc.) that performs calibration and adjustment or secures sufficient integrity.
A description of the comparative measurement is given in Explanation H.
7.2.3

Analysis of the cause of outliers

Measured values that are identified as outliers because of exceeding 3σ include normal values at therate as shown above. In addition, in continuous observation, environmental radiation fluctuates for a
specific period owing to natural phenomena and artificial factors, and outliers are observed at a rate
exceeding the statistical probability. This should be considered when analyzing the causes of outliers.
In addition to the statistical probabilities that depend on the size and distribution of a population, themain factors that produce outliers include: (i) precipitation, (ii) snow cover, (iii) other natural
phenomena such as variations in radiation levels due to inversion layers, etc., (iv) nuclear explosion
experiments in the atmosphere, (v) artificial radiation sources such as nuclear facilities, and (vi)failures in measuring instruments.
To analyze these factors, the following should be done.
(1) Obtain information on each factor.
(2) Select a measuring time interval.
(3) Select a population.
Explanation G gives examples of selecting a population, the causes of major outliers, and defining atime interval required for these analyses. Figure 7.1 shows an example of a flowchart for outlier
analysis.
The recently developed continuous monitors have, in some cases, a capability of obtaining spectral
information concurrently with the dose rate measurement. In such a case, the dose rate contribution foreach nuclide can also be evaluated by spectral analysis. Analysis and evaluation methods are describedin Chapter 10.
7.3

Analysis and evaluation in case of an emergency

This section describes the data that may be measured in situations where radioactive materials may
have been released from nuclear facilities.
7.3.1

Nuclides to be measured in case of an emergency and after an accident37

In case of an emergency, the nuclides to be measured are different between when a radioactive plume
passes through and after radioactive materials such as radioactive cesium are deposited on the groundby fallout, etc. Although the final goal of environmental gamma-ray dose rate measurement is not toevaluate the concentrations and dose rate for each nuclide, it should be understood that the measured
values may be overestimated owing to the difference between the nuclide and its energy, because thedetector has energy characteristics.
Recently, with the improvement in electronic equipment, monitors can increasingly obtain energy
spectrum information. In case of an emergency, this energy spectrum information can be effectively
used to identify the artificial radionuclides that contribute to the dose rate.
Explanation I presents an example of a situation in which the dose rate has increased in case of an
emergency and after an accident.
After an accident, the dose rate decreases with time owing to the physical attenuation of artificial
radionuclides, weathering effect, and penetration of the soil in the direction of vertical distribution.
Therefore, outliers can be quickly extracted by performing the evaluation shown below:• Update the threshold at an interval of approximately 1 month, considering the decrease in dose
rate due to physical attenuation, etc.
• Acquire the integrity of the measuring system through comparative measurement38One-hour value not included in the
average value ± (3 x standard
deviation)

Failures inthemeasuringsystemNo

No

Rainfall andsnowfall

Impact of
nucleartesting

No

Effects from
nuclear facilities

No
Yes
Effects ofprecipitations
Yes

Yes

Yes

Yes

Equipment
adjustment

Figure 7.1

Cause identified

Cause identified

Example of a flowchart for outlier analysis

39

Cause identified

No

Statistical variations
and others

7.4

Dose

There are various definitions for environmental gamma-ray dose rates, including “air absorbed dose(Gy)” and “ambient dose equivalent (Sv).”
Doses should be selected according to the purpose of monitoring, and the measurer is requested toclearly state the dose together with the observation results when reporting.

Table 7.1 Definitions relating to doses
Dose

Indication

Description

Irradiation dose

Coulomb/kilogram(C/kg)

Absorbed dose

Gray (Gy)

Kerma

Gray (Gy)Dose that is the basis of the electric charge dose
generated by radiation at 1 kg of air
Former unit is Roentgen (R)Indicates the energy of radiation absorbed by a
substance.
1 Gy = 1 J/kg
In environmental radiation monitoring, air absorbed
dose is measured.
Defined as the sum of the initial kinetic energy ofcharged particles generated by radiation per unit mass ofa substance. If the acting substance is air, it is called the“air kerma (Gy)” and the air kerma and air absorbed
dose may be considered equal.

Equivalent dose

Sievert (Sv)

Effective dose

Sievert (Sv)

Ambient dose
equivalent (1 cmdose equivalent)Sievert (Sv)

In assessing the effects on the human body, the type andenergy of radiation are considered.
It is calculated for each tissue or organ by multiplying
the absorbed dose of the relevant tissue or organ by aradiation weighting factor.
In assessing the effects on the human body, the sites
exposed to radiation are considered.
It is calculated by multiplying the equivalent dose of thetissue or organ by the tissue weighting factor andsumming the total dose for the whole body.
During normal operation, environmental radiation
monitoring is calculated as:
1 Gy (air absorbed dose) × 0.8 = 0.8 Sv
in case of an emergency, 1 Gy = 1 Sv
Defined in the Environmental Radiation Monitoring
Guideline as the protection amount for which an
exposure dose evaluation is supposed to be performed.
This is the dose equivalent at a depth of 1 cm from thesurface of a 30-cm sphere (the ICRU sphere) withdensity and elemental composition simulating thehuman body.

40

7.5

Analysis and evaluation of each measuring system

Matters common to all measuring systems are described in and before Section 7.4, and this section
describes the matters that differ in the method and the concept of analysis and evaluation for eachsystem.
7.5.1

Installed continuous monitor (an ionization chamber monitor, a NaI monitor, and a simplified
electronic dosimeter)(1) Ground height for measurement
As described in Chapter 5, an installed continuous monitor is measured at different heights forvarious purposes, and the measurement height used should be clearly stated together with theobservation results, when reporting.
(2) Evaluation of results according to the detector
As described in Explanation I.2, when a radioactive plume passes through, the contribution ratio of
low-energy (81 keV) photons due to 133Xe is large, which is assumed to be the main factor resulting
in a high dose rate. In this situation, it is necessary to select an appropriate detector, such as an
ionization chamber monitor without the effect of counting loss.
However, after radioactive materials have been deposited on the ground due to fallout, etc., measured
values can be accurately evaluated by selecting a measuring instrument, considering the measurable
dose rate range.
Refer to Explanation B for the characteristics of detectors required to select a measuring instrument
depending on the situation.
A simplified electronic dosimeter has directional characteristics of high sensitivity to the front of thedetector. In other words, because the sensitivity to the rear of the detector is lower than that to thefront, when the coverage for measurement is set, the direction of installing the detector must be
considered.
7.5.2

Transportable continuous monitor (a transportable monitoring post)(1) Extraction of outliers
The location of a transportable monitoring post is described as “sites where an installed observation
station is not operating owing to the effects of natural disasters or where an installed observation
station is not installed” in the “Emergency Monitoring (Supplementary reference materials for the
Nuclear Emergency Preparedness Guideline)” (Reference 8). The statement of “the sites whereenvironmental radiation dose rates need to be additionally identified, depending on the source
conditions” is also included.
If a planned installation site can be selected in advance during normal operation, the reference data
for acquiring the background level must be accumulated during normal operation.
Note that, if the equipment is installed at “a site with no installed observation station installed,”
outliers shall be able to be extracted based on the same concept as that of the installed station.
(2) Understanding the effect of temperature characteristics
As discussed in Chapter 5, a transportable monitoring post is equipped with a temperaturecompensation circuit but not a temperature control system. In addition, a relatively open situation in
the upward direction is desirable as the installation condition, and solar power generation is often
adopted as a power supply system. In such a case, the detector itself may be exposed to direct
sunlight, causing a response change that cannot be compensated by the temperature compensation
41

circuit. Therefore, it is important to understand the change in response due to temperature, whenmeasurements are taken at the same point for a certain period of time.
7.5.3
Mobile survey system(1) Calculation of a correction factor considering vehicle shielding, etc.
When an installed type is installed outside the vehicle, the primary radiation from radioactive
materials around the vehicle can be directly measured. However, the contribution to the measured
values is reduced within a few meters from the immediate vicinity of the vehicle, owing to theshielding effect by the vehicle body. Therefore, in the case of the exterior type in which the detector
is installed on top of the vehicle, the measurement coverage is extended from a position slightly awayfrom the vehicle to an outward range. Vehicles equipped with a detector normally installed to the roof
(thrusting over the vehicle) are suitable for the measurement when a radioactive plume is passing,and they have been calibrated. When this is used to measure the dose rate from radioactive materials
deposited on the ground, the evaluation method should be considered, taking into account thegeometry such as the shielding and height of the vehicle.
If a measuring instrument is installed in a vehicle, even though it is an installed type or a portable
type, measurements should be evaluated by obtaining the correction coefficients of both the insideand outside of the vehicle in advance. The calculation procedures of the correction coefficients areshown below:[1] Park the vehicle in a flat open area and at a point where the dose rate is uniform around thevehicle.
[2] Perform the measurement using a portable measuring instrument installed at a predetermined
position in the vehicle.
[3] Move the vehicle, and perform measurement with the detector at a height of 1 m above
ground, at the point used in [2] where the vehicle is no longer positioned.
Perform these measurements at several points with different dose rates.
[4] Calculate the ratio of the results obtained in [2] and [3] above, and use it as the correction
coefficient for the inside and outside of the vehicle.
The correction coefficient calculated by this method includes a correction term for converting theshielding effect of the height and the vehicle body.
If possible, it is desirable to obtain the measurement for calculating the correction coefficient at
multiple points with different dose rates and to use it as a regression equation from the dose rates ofboth the inside and outside of the vehicle.
In addition, the above concept is based on the assumption that the measured values taken outside thevehicle, which are supposed to be the basis, are correct. Consequently, national standards andtraceability must be secured for the measuring instrument to be used.
By applying the concept of the correction coefficient for both the inside and outside the vehicle to themeasured values obtained by an installed-type mobile survey system in which the detector is installedoutside, the evaluation considering the geometry such as the shielding and height of the vehicle canalso be solved.
When employing the correction coefficients for the inside and outside of the vehicle described above,a validity check must be conducted by calculating the correction coefficients for inside and outside
the vehicle at the same point in advance, using the measurement point that was previously used for
calculating the correction coefficients as the reference point.
The concept of the correction coefficients for inside and outside the vehicle explained in this sectionis that the correction coefficients for inside and outside the vehicle must be obtained for the situationsof normal operation and in the case of a nuclear disaster separately, because the energy distribution
and the incident angle of incident photons are different between normal operation and in the case of a
nuclear disaster. If it is difficult to calculate the correction coefficient for inside and outside thevehicle in the case of a sudden nuclear disaster, the correction coefficient for inside and outside thevehicle of “1.3,” which was tentatively adopted after the Fukushima Dai-ichi NPS accident, should
be used, and a correction coefficient for inside and outside the vehicle should be calculated later.
42

(2)

Extraction of outliers

To extract outliers in a mobile survey system, the dose rate in the background on a regular routeshould be measured during normal operation, in order to acquire the level. Numerical values for thenatural environmental gamma-ray dose rate of (the background) vary greatly depending on thesurrounding environment. Therefore, it is important to grasp the background level from the normal
time, in order to help to extract outliers. Explanation G.3 shows the variation in environmental
gamma-ray dose rates owing to the surroundings (nature).
Further, when measuring dose rates using several different vehicles or methods, it is important for theevaluation of the dose rate distribution to conduct comparative measurements at the same points inadvance, and to understand the relationship between different measurement conditions.
In case of an emergency, monitoring cannot always be conducted on regular routes where
investigations are carried out during normal operation. In such a case, absolute values at each pointshould be evaluated at the initial stage of the accident. In addition, when the dose rate has decreased
owing to physical attenuation and human activities after several years have elapsed since theaccident, evaluations should be performed by discriminating the contribution of natural radionuclides
from the contribution of artificial radionuclides deposited owing to the impact of the accident.

43Chapter 8
8.1

Measurement of environmental gamma-ray spectrum by a NaI monitor

OutlineBy analyzing the gamma-ray spectrum obtained using a NaI monitor, the contribution dose rate, fromuranium series nuclides, thorium series nuclides, and 40K in the case of natural radionuclides, can beevaluated. For artificial radionuclides, the contribution dose rate due to specific radionuclides can also
be calculated by the resolution of a NaI (Tl) scintillation detector, if the energy pulse-height
distribution is not so complicated. The calculation of the contribution dose rate is performed roughly
as follows:(1) Environmental gamma-ray spectra are measured with an adjusted NaI monitor.
(2) Energy calibration of the pulse-height distribution obtained is performed.
(3) Unfolding is performed by the stripping method or the response matrix method using the response
function prepared in advance, and dose analysis is performed after conversion into an incident
gamma-ray spectrum is carried out.
After the environmental gamma-ray spectrum acquired using a NaI monitor is transmitted to thetelemeter system, spectral analysis is carried out at any time by the software program incorporated in
the system, and the effects of artificial radionuclides, if any, can be confirmed by evaluating thecontribution dose rate for each nuclide.
The following sections describe the basic principles and methods of environmental gamma-ray
spectrum analysis.
8.2

System configuration required for environmental gamma-ray spectrum measurement

Many NaI monitors currently in place are used to measure gamma-ray spectra.
The NaI monitor developed as a continuous monitor consists of a NaI (Tl) scintillation detector (ascintillator, photomultiplier tube, and preamplifier), main amplifier, ADC circuit, dose rate arithmetic
circuit, data display unit, data processing unit, recorder, and MCA. A typical system diagram of theconfiguration of a NaI monitor is shown in Figure 8.1.
In conducting a environmental gamma-ray spectrum measurement, analyses are performed on thebasis of the pulse-height distribution accumulated by the MCA. After being collected by the dataprocessing unit or the telemeter system, the pulse-height distribution accumulated by the MCA isconverted into an energy spectrum and analyzed for the dose calculation and so on.

44Detection unit

NaI
(Tl)

Measuring unit

Photomultiplier
Preamplifier
tube

Main amplifier

ADC circuit
High-voltage powersourceRecorderDose rate arithmetic circuit

MCA

Data display

Data processing unit

Data processing

Figure 8.1

Example of the configuration of a NaI monitor

8.3 Preparation for measurement
To carry out environmental gamma-ray spectrum measurements as part of the measurement by acontinuous monitor, data are continuously and automatically collected, usually without taking special
measurements separately.
Here, as a basic concept for environmental gamma-ray spectrum measurement, the basic proceduresfor separately measuring with a portable NaI spectrometer are shown.
The measuring system consists of: (1) a radiation detector, (2) preamplifier and main amplifier for
amplifying and shaping radiation signal pulses, (3) pulse-height analyzer, and (4) recorder for themeasurement results. To obtain reliable measurement results, it is necessary to fully understand thefunction and characteristics of each component of the system and combine the components for use
such that they can operate well. The key parts of adjustment and selection for determining the system
requirements are:[1] Applied voltage of the detector,[2] gain of the main amplifier, waveform shaping conditions, and DC level of the DC reflex
circuit,[3] and the conversion gain of the ADC for the pulse-height analyzer and the number of channels
used.
The waveform shaping conditions are not only related to the pulse resolution but also change thetemperature dependence of the pulse-height values in relation to the decay time of the scintillation
pulse, which must be considered when a condition is selected. The waveform shaping time constant isusually chosen in the range of 0.5–2 μs.
The number of channels used for the MCA shall be 500–1000 channels, considering that the energy
width per channel of the spectrum obtained by the NaI (Tl) scintillation detector is several to 20 keV,and usually 5–10 keV/ch is appropriate; further, this number can be set such that cosmic ray
information can be obtained at 3 MeV or more.
Once a policy has been established for each of the above points, check and adjust the following items
while confirming that each unit is operating satisfactorily.
45[1] DC level adjustment of the DC reflex circuit[2] Noise level testing[3] Testing and confirmation of the measured energy range[4] Check for any malfunctions in MCA[5] Testing of the detector energy resolution (depending on the half-value width)[6] Linearity testing of pulse-height versus a channel number
For the adjustment in [1] (not available in some devices), a reference source (e.g., 662-keV gamma
rays of 137Cs) shall be used by observing the pulse waveform with an oscilloscope.
For [2], this test shall be conducted in parallel with the test in [3]. The measured energy of gamma
rays ranges from 40 keV at minimum to 3 MeV or more, and the levels of a lower-level discriminator(LLD) and an upper-level discriminator (ULD) of a pulse-height analyzer shall be set within thisrange. If a strong noise were produced in the range of close to 30–40 keV, not only would themeasurement accuracy of low-energy gamma rays be decreased, but the resolution of the entire pulseheight would also be impaired; thus, measures of eliminating possible causes must be considered. The
LLD shall be set as low as possible without noise effects, and the ULD shall be set as high as possible
so as to obtain more information about cosmic ray components.
In [4], a monoenergetic gamma ray shall be used to check the shape of the entire pulse-height
distribution, the symmetry of the total absorption peak*1, etc., for confirming abnormalities, if any. In
addition, when an excessive signal input such as a pulse caused by cosmic rays is identified, anabnormal signal may be generated in the gamma-ray region owing to the amplification circuit andother transient characteristics; thus, an adequate inspection is required.
For [5], 662-keV gamma rays of 137Cs are used to determine the quality with the resolution of the total
absorption peak. Normally, approximately 7%–9% is the range of normal values, and any
abnormalities should be checked against the manufacturer's performance report, etc.
For [6], testing shall be conducted by distinguishing the characteristics of the electrical circuit only
from those of the NaI (Tl) crystal, including the light emission characteristics. The characteristics ofthe electrical circuit section shall be examined by using a pulsar*2 or similar, based on the relationshipbetween various outputs through an attenuator*3 and the channel numbers of the MCA corresponding
to them. Because the channel number corresponding to the pulse-height value of zero is different fromthat obtained with reference to the gamma ray energies, these relationships shall be accurately
determined in relation to the energy calibration of the pulse-height distribution (energy calibration will
be discussed later).
8.4 Primary processing of measured data
In preparation for energy calibration, the data of photo-peak channels corresponding to gamma-ray
energies are arranged, based on the pulse-height distribution data obtained by a reference source andthat obtained by this measurement. Because the relationship between the energy and the channel
number is almost linear, it can be determined that a normal measurement has been made if there is nolarge difference observed in the slope for each measurement.

*1

A peak that is produced in the gamma-ray spectrum where all the energy of the incident gamma rays isabsorbed by the detector*2
A device that electrically generates pulses at arbitrary energies (channels)*3
An attenuator is an electronic device that attenuates pulses to an appropriate level

46

8.5

Data acquisition for energy calibration

Perform energy calibration as necessary, in accordance with the energy calibration method that isestablished in advance. The energies of gamma rays used for calibration should be obtained in detail,over the entire energy range of the gamma rays to be measured; however, for the purpose of operation,
calibration is carried out at two points; at 1461 keV of 40K and at the energy level corresponding to thezero points, which are confirmed in advance. See Chapter 10 for details.
If a gamma-ray source can be used without special restrictions, the following methods will beapplicable.
Because of the limited number of reference sources available, a method of obtaining necessary
information from a few sources should be considered in advance. In terms of the calibration points, atleast one point should be selected in the range of 50–200 keV, and one point in the range of 1–2 MeV.
For this purpose, 0.122 MeV of 57Co is suitable as a calibration point on the low-energy side, and 1.46
MeV of 40K as a natural radionuclide or 1.33 MeV of a 60Co source is suitable as a calibration point on
the high-energy side. If an energy other than these is used to perform a calibration, a high-precisionenergy calibration shall be performed, with reference to Chapter 10.

47

Chapter 9
9.1

Response function used by NaI monitor for measured data

Description of a response function
To obtain the incident gamma-ray spectrum on the detector from a pulse-height distribution thatis output from the detector, knowledge of the response indicated by the detector when amonoenergetic photon is incident on the detector is required. If a pulse-height distribution P(h)is obtained while observing a photon with an energy spectrum N(E) (E is energy), this can begenerally expressed in the integral equation shown below. In this equation, ε(E) refers to theefficiency of the detector. K (E,h) or ε(E)K(E,h) in this equation is called a response function.
The former is obtained by normalizing the area of a pulse-height distribution group obtained at
the time of a monoenergetic photon incidence to 1, and the latter means the pulse-height
distribution group obtained at the time of a monoenergetic photon incidence.
P(h) = ∫ ε (E)K(E, h)N(E)dE(9.1)

A response function is determined by an experimental method or a Monte Carlo simulation. A
response function depends on the target radiation source and the arrangement of the detector used.
Because radiation in the environment is generally not isotropic, special consideration should be given
to directional dependence, when a small direction-dependent detector (spherical) or a cylindricaldetector is used.
9.2

Creation of a response function

A response function is obtained by measuring pulse-height distribution groups of the detector formonoenergetic photons with several different energies and dividing them by their corresponding
incident photon fluences. The irradiation conditions are that the source shall be placed in a direction
perpendicular to the axis of the cylindrical detector, with less contribution of scattered rays, and also
close to the parallel beam. The pulse-height distribution of arbitrary energies shall be obtained byinterpolation. A Monte Carlo simulation method is also available (see Explanation J.2).
The segmentation width (or channel) of a response function should be selected as appropriate
according to the purpose of the analysis, by setting it to 10, 20, 50, 100 keV, and so on. A response
function may also be used, as represented in a matrix of 10 rows and 10 columns, or 50 rows and 50
columns.
When a response function is represented as a matrix, a value obtained by adding each element of thecolumn (or the row) provides the energy efficiency corresponding to the column (or the row) when asingle photon is incident. The validity of the response function (the matrix) created can be checked bycomparing this value with the efficiency ηt, which is theoretically calculated using equation (9.2) for
the detector. It is desirable that they be nearly identical or the difference between them be as small as
possible.

=1− (

)

+2

(

)

+

(9.2)

where μ is the linear attenuation coefficient of the scintillator, and d is the diameter of the detector.
48

In addition, a P/T ratio can be calculated using a response function, so as to compare it with anexperimental value. In this case, an experimentally determined P/T ratio is reduced by approximately
5%–10% (see Appendix 1 “Calculation examples of a response function”), because an experimentallyobtained pulse-height distribution contains the contribution of scattered rays.
9.3

Application of a response function

To analyze a pulse-height distribution in which monoenergetic photons with different energies areobserved mixed at various intensities, and to determine the photon energy and intensity individually,
the stripping method (the peel-off method) is applied. This entails focusing on the total (energy)
absorption peak identified at the highest-energy position, applying the response function sequentially,
and then stripping from the highest-energy side. In doing so, the following methods can be used: amethod to subtract one after another in accordance with the response function and the peak on thehighest-energy side, a method to apply the least-squares fitting to minimize the square of thesubtracted residual, and a method to subtract from the highest-energy side on the basis of the totalenergy absorption peak and the P/T ratio.
In the case of environmental radiation observed when several types of monoenergetic photons andscattered rays are overlapped, the incident gamma-ray spectrum can be obtained by analytical methods
such as the inverse matrix method or the successive approximation method, in addition to the stripping
method. Among the abovementioned methods, the response matrix, in which a response function isrepresented as a matrix, is used.
The accuracy of the incident gamma-ray spectrum as an analysis result is determined not only bystatistical errors in a pulse-height distribution and the analysis method, but also by the compatibility of
the response function (or the response matrix) and the response generated by the measuring system inpractice.

49

Chapter 10 Processing of data on environmental gamma-ray spectrum measured by NaI
monitor
The processing of a pulse-height distribution obtained by a NaI monitor is, in many cases, performedby pulling the entire pulse-height distribution back to the incident spectrum and treating several
channels as a monoenergetic block, rather than focusing on a particular peak.
The details of the processing include the energy calibration of a measured pulse-height distribution,
conversion to an incident spectrum, and several analyses performed to use the results for specificapplications.
10.1 Energy calibration of the measured pulse-height distribution
In a measured pulse-height distribution, the relationship between energy and the pulse-height or thenumber of channels is not generally linear. If a linear relationship has been established, you may skip
this section. If this relationship is not established, the analysis result would be biased owing to the low
resolution of the detecting system. Therefore, this must be calibrated. To do this, the relationshipbetween the channel and the energy shall be clarified, and the pulse-height distribution shall be
converted into one usually having a constant energy width (channel width) per channel. To perform
this calibration process, the relationship between the channel and the energy of a pulse-height
distribution is converted using a polynomial that produces a best approximate point among several setsof points of the center channel of the peak in the spectrum and the energy corresponding to it. As apolynomial, a linear (straight line) type is usually used, which approximates three points of the energy
around a zero point (for example, 60 Kev of 241Am, 122 keV of 57Co, or the specific energy of an input
signal zero (for example, −17 keV)), 1461 keV of 40K, and 2614 keV of 208Tl of the thorium series(Figure 10.1). To further refine the calibration, a higher-order equation is used for the transformation,
which requires selecting a corresponding peak energy in the required number (if the nth expression isused, the required number is (n + 1) or more). Once the energy/channel relationship is obtained, eachchannel can be easily converted for the energy calibration. The converted spectrum is called thecorrected pulse-height distribution. In the case of a linear approximation, the channel width for apulse-height distribution is set to 10, 20, 50 keV, and so on, per channel, and a value of 100 or 200
keV may be selected, though it is not usual. Inaccuracy in this correction, if any, would directly
produce uncertainties in the final results of analyses; thus, precautions must be taken.

50Energy
(keV)

2614

1461

60
Aroundthe zero
point

40

K

208

Tl

ChannelFigure 10.1 Application example of linear approximation of three points

10.2 Conversion to an incident spectrum
Conversion to an incident spectrum is a process of data processing for obtaining an incident gammaray energy spectrum from the corrected pulse-height distribution, and is performed by applying a
response function as described in Chapter 9. In this case, that which is contained in a response
function must be clarified in advance. In other words, attention should be paid to the conditions underwhich the response function was created, the difference from the conditions under which the response
function was measured in the field, and especially, the incident direction of radiation. The conversion
methods are roughly divided into the following two: the stripping method (the peel-off method) andthe response matrix method.
(1) Stripping method
In the stripping method, a response function for a single peak corresponding to the high-energy
channel of the corrected pulse-height distribution is subtracted from the corrected pulse-height
distribution, and the value in the high-energy channel is reduced to almost zero; then, this operationis sequentially applied to the low-energy side. As a result, the response function for obtaining a
distribution in the low-energy region and the method used to introduce the constant at that time
affects the accuracy of the incident gamma-ray spectrum that is obtained after the processing iscompleted. Therefore, the shape and numerical values of the response function to be stripped must be
clarified. It is also necessary to provide the conditions used for creating the response function.
An example of the stripping method is given in Appendix 2.
(2) Response matrix method
The response matrix method is a solving method in which a specific photon energy is determined inadvance, a response function matching the corrected pulse-height distribution (a numerical matrix for51

each energy region) is created, and a matrix of each energy is solved by a simultaneous equation for
the actually measured corrected pulse-height distribution; two methods are used: one is to solve bycreating a conjugate matrix in advance, and the other is to solve by the successive approximation
method. In this method, as with the stripping method, it is desired that the measurement conditions
used when the response matrix is created be the same as those employed in the field. Differences
between the conditions, if any, would contribute to uncertainties in the results.
An example of unfolding by the successive approximation is presented in Appendix 3.
In applying both methods, the accuracy of the response function used is essential. When the detectedbody for which a response function is determined and the detected body actually used forenvironmental measurement are different, a difference may be produced in the result due to various
differences in the detected bodies, in particular the differences in the structure of a scintillator such as
the size, shape, weight, and casing, and differences in resolution. The difference in weight can becorrected as the difference in cross-sections. In converting the energy spectrum of incident gamma
rays, it may be necessary to further consider the contribution of cosmic rays from the higher-energy
region and the background derived from the detected body itself.
At present, response functions that can be applied to scintillation spectrometers with different
resolutions for each shape and size of the detector have been published, and they are incorporated into
a commercially available analysis software.
10.3 Representation of the results
There are many expressions for the analyzed measurement results. These include: (1) the total
counting rate value, (2) total count rate value for a corrected pulse-height distribution (3) gamma-ray
flux density (within the total energy range and specific regions), (4) irradiation dose rate (within the
total energy range and specific regions), (5) absorbed dose rate of NaI (Tl), and a graphical display
specified for those above. The statistical errors of the count values, precision, details associated withchannel calibration, functions used and their precision, dimensional structure of the detected body,
constant used, basic constants, data processing technique, processing program, and handling of the
background are also included. Of these, (1), (2), (3), and (4) mentioned above, and the ratios of (2) and(3) to (4), are often adopted as typical expressions of the results.
Note that the flux density of cosmic rays obtained from a high-energy region and its estimated doseare helpful for understanding the aspects of radiation in the environment. In doing so, if the manner inwhich components other than gamma rays in the spectrum are treated were not specified, the results
would not be interpreted clearly. If the relationship between the irradiation dose rate across the entire
energy region and the photon flux density or the counting rate is constant, the irradiation dose rate canbe immediately obtained from the counting rate by the photon flux density. In this case as well,because the incident gamma-ray spectrum is already obtained, the irradiation dose rate can be obtainedfor each energy region.
10.4 Calculation of dose contribution by radionuclide
In measuring a environmental gamma-ray spectrum method using a spectrometer, the dose
contribution to the environment for each radionuclide component is determined. A NaI (Tl)scintillation spectrometer is used only when limited types of artificial radionuclides (e.g., 131I, 137Cs,etc.) are present, because high resolution is not expected, unlike a germanium semiconductor detector.
On the contrary, in ordinary environmental radiation fields, the dose contributions by naturalradionuclides such as potassium, uranium, and thorium series are typically separated.

52

10.4.1 Use of count values in the peak area
This data analysis is performed by applying the peak area present in the spectrum, the energy andemission rate of the corresponding nuclides, and the detection efficiency of the detector.
Because it is generally not easy to select the peaks corresponding to the respective energies of theemitted gamma rays in a NaI (Tl) scintillation spectrometer, this is applied only to a limited number of
nuclides that emit relatively simple gamma rays.
10.4.2 Determination of dose contribution by nuclide by the matrix method
In a scintillation spectrometer, for the reasons already mentioned, it is not easy to separate the dose
contributions of each nuclide that emits gamma rays of various energies. Therefore, it is common thatthe energy regions covering a wide range are determined in advance, and a matrix analysis isconducted. This method is used for measuring normal environmental radiation, mainly composed ofnatural radionuclides; the method is explained below:
The dose contribution of natural radionuclides in the environment is usually generated by potassium,
uranium, and thorium series, and they are typically separated; however, in some cases this is followedby the contribution of artificial radionuclides.
To separate the dose contributions for these three main natural components, the energy regions
corresponding to each component are determined in advance, the contributions contained in eachregion per unit dose rate (dose rate, counting rate, flux density, etc.) are prepared as basic data for eachcomponent, and then the matrix is solved. For the energy regions employed in this method, thefollowing range is defined as a standard. An example of the matrix constant representing the dose rate
contribution between regions is illustrated in Table 10.1.
Potassium region (1.34–1.60 MeV)
Uranium series region (1.61–2.30 MeV)
Thorium series region (2.31–3.00 MeV)

53

Table 10.1 Contribution coefficients between regions for solving a matrix (the irradiation dose rate ineach region per μR/h; the numbers in parentheses represent the value in “cps” for a 3-in
spherical scintillator)
K-40 region

U series region0

Th series region

K-40

0.5 (1.5)

(0.0)

0

(0.0)

U series

0.106 (0.508)

0.255 (0.565)

0.024 (0.0236)

Th series

0.043 (0.689)

0.124 (0.674)

0.275 (0.343)(Note) These coefficients are based on the following measured data and are expressed in R,
which is the measurement unit used.
K-40: Standard source (KCl)

U series: Cave filled with radon in Houston

Th series: Kerala District in India

The sum of the respective values obtained using these coefficients must correspond to the measured
total dose rate value. However, differences in the matrix coefficients, counting errors, and conditions
used in the measurement environment may produce a difference of approximately 10%, and if a
difference exceeding this value is produced, it must be examined in detail. It is desirable to refrainfrom adopting the results when the difference between these items exceeds 10%.

54

Explanation

55

56

Explanation A

General matters on environmental gamma rays measured by a continuous
monitor

The monitor focused on in this manual is called “continuous monitor,” which can measure gamma raysin the energy range of tens of keV to several MeV among different types of ionizing radiation in thegeneral environment, continuously obtain count values with short time intervals, and output themeasured values in real time (to an analog or digital display). Many types of continuous monitors forenvironmental gamma rays are currently used, reflecting the historical background and owing totechnological advances; their characteristics and performance are quite different. However, regardlessof their type, the measurements obtained from these monitors are mainly the air absorbed dose rate
(μGy/h, etc.), ambient dose equivalent rate (μSv/h, etc.), or detector counting rate (cps, cpm, etc.)
acquired at the location of the detector.
Because continuous real-time data can be obtained using a continuous monitor, changes in dose rates
can be tracked with time, and along with meteorological observation data, the cause of these changes
can be clarified more easily. This an important characteristic of continuous monitors, when comparedwith a thermoluminescence dosimeter (TLD) or fluoroglass dosimeter (RPLD), which provide
integrated doses over a long time. In some cases, it is also possible to discriminate natural radiationfrom artificial radiation by analyzing continuous recording pattern, and in such cases, the contribution
from nuclear facilities can be estimated. In addition, many of the systems currently in use can acquireenergy information, which can be used for a more accurate discrimination.
However, it is difficult to install many continuous monitors because of the complicated mechanism,
considerable labor required for maintenance and management, high cost, high power consumption, etc.
Therefore, detailed dosimetry of the surrounding areas has generally been left to TLD and RPLD;
however, transportable monitoring posts and simplified electronic dosimeters may currently play thisrole.
Owing to the differences in the type of detector, structure around the detector, characteristics of thedevice, signal processing circuitry, and so on, differences may be produced between the measurements
taken by various types of monitors, even those taken at the same points concurrently. Many
problematic matters are identified in the vicinity of the detector, such as the contribution of cosmic
rays, shielding effects, directional characteristics, and differences in the energy characteristics. As amatter of special attention required for monitoring during normal periods, regarding the contributionof cosmic rays, for example, in the G(E) function method using a scintillator as a detector, high-energy
components including cosmic rays of 3 MeV or more are involved in the general use. However, thoseof 3 MeV or more are separated in the currently prevailing usage,; thus, a full understanding of thespecification details is important. The dose rate indicated by a monitor using a scintillator as a detector
may be considered to be lower than that indicated by one using an ionization chamber as a detector, by
a value corresponding to the normal contribution of cosmic rays. In addition, even if a detector of thesame model is used, depending on the differences in the production lots, the difference in thecontribution dose rate due to self-contamination makes it difficult to compare the measured values
with each other. With regard to the contribution dose rate due to self-contamination, one available
measuring method is to conduct in a basement or similar environment, where the contribution of theenvironmental gamma-ray dose rate and cosmic ray dose rate is reduced, and to evaluate from thepeak counting rate of 40K; another method is to use a representative value for each of the constituent
materials of the detector.
If the measured values are corrected considering these factors, the measured values of different
detector models can be compared; however, in the current situation, such corrections are not easy.
Therefore, when different detectors and different measuring devices are installed at the same point,
attention should be paid when comparing the measured values taken at the background level with eachother.
57

Explanation B
Explanation B.1

Principle, configuration, and characteristics of each monitor

Ionization chamber monitor

The ionization chamber detectors for continuous ionization chamber monitors currently in practicaluse include a metal spherical ionization chamber, in which pure argon gas or nitrogen gas is injectedwith high pressure. To enhance the sensitivity, the volume is increased, or the injected gas ispressurized. An ionization chamber monitor displays the dose rate as an analog output of a current–
voltage conversion by passing an ionization current through a high resistance, or as a digital output ofcharge by current integration at regular time intervals. An ionization chamber monitor has good energy
and directional characteristics, and it can have a wide range for dose rate measurements, depending on
the specifications. The energy characteristics and dose rate linearity of various ionization chambers areshown in Figures B.1 and B.2, respectively.

Air absorbed dose rate: approximately 0.4 μGy/h

2.5
Ａ(aluminum)（アルミ製）A
B
Ｂ(stainless（ステンレス製）
steel)C
Ｃ(stainless（ステンレス製）
steel)
D
Ｄ(stainless（ステンレス製）
steel)E
Ｅ(aluminum)（アルミ製）

Response

Response

2.0
1.5
1.0
0.5
0.0
10

100

1000

Energy(keV)(keV)
EnergyFigure B.1

Example of energy characteristics of various pressurized ionization chambers

58

Irradiation
photonenergy:keV
Irradiation
photonenergy:
8080
keV

2.0
A (aluminum)Ａ（アルミ製）B (stainlesssteel)
Ｂ（ステンレス製）
C
steel)Ｃ (stainless（ステンレス製）
D (stainlesssteel)
Ｄ（ステンレス製）
E
Ｅ(aluminum)（アルミ製）

Response
Response

1.5
1.0
0.5
0.0

0

0.4

2

30

4

100

6

Dose rate
Dose
rate(μGy/h)(μGy/h)(* Standardized on a comparison value of 0.4 μGy/h)Figure B.2

Example of dose rate linearity of various pressurized ionization chambers

As described in Chapter 4 of the main text, the factors affecting the measured values, such as theenergy characteristics of an ionization chamber dosimeter, are mainly the wall material and its
thickness, as well as the type of gas contained and its pressure.
For the purpose of dosimetry, it is ideal that the materials of a detector can be selected depending on
the dose to be measured. Ideally, a substance equivalent to air should be used to measure the absorbed
dose of air, and a substance equivalent to tissues should be used to measure the absorbed dose oftissues. An air-equivalent ionization chamber is most suitable for environmental gamma-ray
monitoring because it measures the air absorbed dose, but, in many cases, the sensitivity is insufficient
for low-dose-rate measurement. In an ionization chamber used as a continuous monitor, argon gas orsimilar is pressurized and injected into an aluminum or stainless-steel container to enhance thesensitivity. Because the atomic number and density of stainless steel and argon gas are higher thanthose of air, the mass–energy absorption coefficient becomes remarkably large in the low-energy
region, despite the high sensitivity, and the response increases. Further, in the low-energy region,
gamma rays are absorbed by the stainless-steel wall, resulting in a rapid decrease in the response. Theenergy characteristics of an ionization chamber can be expressed by reflecting these factors combined.

59

Explanation B.2
B.2.1

G(E) Function method by a NaI (Tl) scintillation monitor

Principles and configuration of the monitor

One of the methods of evaluating the absorbed dose and ambient dose equivalent from a NaI (Tl)scintillation spectrum distribution is to use a spectrum-dose conversion operator for the dose
calculation by directly applying a weighting function to the pulse-height spectral distribution of
gamma rays. This is the G(E) function method.
The G(E) function method uses a spectrum-dose conversion operator (G(E)) to evaluate dose rates
directly from the gamma-ray pulse-height distribution.
Assuming that the pulse-height distribution is P(E) and the in-situ dose rate is D, D is linked asfollows, through the spectrum-dose conversion operator G(E). The first line of the equation represents
the principle of the G(E) function method, and the second line represents the application of the method
used when the dose rate is calculated from the pulse-height distribution measured by a multichannel
pulse-height analyzer.
D= ∫

= ∑

where
D
P(E)
P(I)
E
I
Emin
Emax
Imin
Imax
G(E), G(I)( )• ( )( )• ( )(B.1): Dose rate (nGy/h): Pulse-height distribution (cpm/keV): Pulse-height distribution measured by a multichannel pulse-height analyzer(cpm/channel): Pulse-height (keV): Channel number: Lower limit of measurement of a pulse-height (keV): Upper limit of a dose rate evaluation (keV): Channel number corresponding to Emin
: Channel number corresponding to Emax
: Spectrum-dose conversion operator (nGy/h/cpm)

A method of performing a dose calculation based on the above equation using an analog signal ofpulse-height and an electronic circuit is called the DBM method, and a monitor corresponding to the
DBM method is called a DBM monitor. Recent advances in computer technology have made it
possible to perform small-scale and high-speed operations; thus, the method of digitizing apulse-height and converting it into a dose by multiplying directly by the G(E) function is now
prevalent.
Emin and Emax are the lower and upper limits of the energy range for dose rate evaluation, respectively,
and when environmental gamma rays are to be measured, the values of Emin = 30–50 keV and Emax =3000 keV are usually used. As can be seen from this equation, the G(E) function acts as a type ofenergy-dependent weighting function. This method has an advantage that the dose rate can beevaluated without converting a pulse-height distribution to an incident gamma-ray energy spectrum.
The G(E) function applied to each detector is determined separately, depending on the shape and size60

of the detector. An example of the G(E) function is shown in Figure B.3.
In the case of a NaI (Tl) detector, the most important factors that determine the response function and
the G(E) function are the shape and size of the NaI (TI) crystal, and the shape of the crystal container.
In the case of NaI (Tl) detectors, if the shape and dimensions are the same, the response characteristics
are also the same; therefore, once the G(E) function for one type of NaI (Tl) detector is determined,
the G(E) function can be directly applied to other NaI (Tl) detectors of the same type as well.
The G(E) function method was originally developed by considering the measurement of the
gamma-ray dose (rate) by a NaI (Tl) detector; however, it is also applicable to a detector that canobtain the energy information (pulse-height distribution) of incident radiation, as with a NaI (Tl)detector (References 9 and 10).

Figure B.3

G(E) function of a NaI (Tl) detector

B.2.2
Characteristics of the monitor(1) Stability[1] Detector
The most important aspect related to the changes in response in this type of monitoring is thestability of the detector and circuit with respect to the changes in ambient temperature in theenvironment. The instability of the detector against temperature changes can be divided into two
factors: the temperature dependence of the luminous efficiency of NaI (Tl) scintillation and the
decay time of light, and the temperature dependence of photoemission of the photoelectric surface
of a photomultiplier tube. The overlapping part of these two changes is usually the temperature
characteristic of the detector, and a combination of this with the pulse shaping time constant of theelectronic circuit (a digital type is less sensitive to temperature changes than an analog type) causes
the temperature dependence of the monitor.
[2] Electronic circuit
The temperature characteristics of a high-voltage power source are also important. In general, the61gain change of a photomultiplier tube used in radiation measurement is close to 10 times the rate ofvariation of the high voltage, and when stability within a range of ±1% of the gain change isrequired, the stability of the high-voltage power source must be less than ±0.1% (±1 V equivalent orless) of the gain change.
[3] Measures for stabilization
When gamma rays are monitored in an environment with a wide range of temperature fluctuations
throughout the year, special consideration should be given to the temperature characteristics;
stability should be secured at least within a range of ±(3–5%) throughout the year. Many of thedevices currently on the market take various measures for stabilization, such as thermal insulationof probes, isothermal treatment, and temperature compensation by an electronic circuit, andadequate stability is ensured throughout the year. However, this is the case of a fixed station or the
specifications corresponding to it; the fluctuation range becomes large when measures for
isothermal treatment cannot be applied owing to the nature of the system for a transportable
monitoring post, etc.
(2) Energy characteristics
The energy characteristics of a monitor using the G(E) function method are usually within ±20% in
the range of 0.1–3 MeV (Figure B.4), which is sufficient, considering the energy distribution ofenvironmental gamma rays. Note that because these characteristics also vary owing to the absorptionof gamma rays by the cover of the probe, an unnecessarily thick cover must not be used. Therefore,
the response at 60 keV is low, as shown in Figure B.4.

Air absorbed空気吸収線量率：約0.1～約1
dose rate: approximately 0.1–1μGy/h
μGy/h

2.0

A (3”φcylindrical)
Ａ（３″φ円柱形）
cylindrical)
B (3”φ（３″φ円柱形）
cylindrical)
C (2”φ（２″φ円柱形）
D (3”φcylindrical)
Ｄ（３″φ円柱形）
E(2”φcylindrical)
E （２″φ円柱形）

Response

Response

1.5
1.0
0.5
0.0
10

100

1000

Energy(keV)(keV)
EnergyFigure B.4

Energy characteristics of a NaI monitor using the G(E) function method(3) Response and counting errors
This type of measuring instrument exhibits a close relationship between the energy compensation
range and the response; when the compensation range is narrowed, the response (in this case, thecounting rate per unit dose rate) increases, while, when the compensation range is widened, the62response decreases. Regarding the energy range of environmental gamma rays, the maximum energy
range is 2.62 MeV for natural radiation, and no significant radiation of energy exceeding 3 MeV ispresent, except for gamma rays from 16N, which are attributable to the facility. Therefore, energy
compensation up to 3 MeV is sufficient.
The standard deviation of a measured value can be calculated considering the response andmeasuring time shown above. Figure B.5 shows an example of a calculation of measurement errors
in 10-min measurements taken with scintillators of various sizes.

2 in. φ × 2 in. NaI (Tl) detector

3 in. φ × 3 in. NaI (Tl) detector

Figure B.5

Statistical errors (1σ) in a 10-min measurement of environmental gamma rays by theG(E) function method (Reference 11)(Note) Because the figure is cited from Reference 11, the former unit R is used.

(4) Directional characteristics
NaI (Tl) scintillators have different directional characteristics depending on their shape, but the actual
characteristics vary further depending on the structure of the probe in use. Spherical scintillators havethe most excellent directional characteristics, and cylindrical scintillators develop exhibit their best
directional characteristics when the ratio of the diameter to height is 4:3. In the case of a scintillator
in which the diameter and height are equal, a difference of approximately 8% is produced in the
response between the axial and lateral directions. However, because of differences in the structures ofindividual probes in use, the precise directional characteristics should be determined bymeasurements. Figures B.6 and B.7 show examples of the directional characteristics when the axial
responses of a 3 in. φ × 3 in. cylindrical and 3 in. φ spherical scintillator are set to 1.0.

63

Figure B.6

Example of directional characteristics of NaI (Tl) scintillator (3 in. φ × 3 in. cylindrical)(Reference 11)

Figure B.7

An example of directional characteristics of NaI (Tl) scintillator (3 in. φ spherical)(Reference 11)(5) Contribution of cosmic rays and self-contamination[1] Contribution of cosmic rays
It is reasonable to treat cosmic ray components by dividing them into pulses of 3 MeV or less andthose of 3 MeV or more at the pulse-height. It is very difficult to carry out a separation valuation on64

the ground for the components classified as 3 MeV or less, because they overlap with theenvironmental gamma-ray spectrum. They are usually evaluated by taking measurements on a sea
or lake not affected by gamma rays from the Earth’s crust, or by extrapolating high-altitude
observation values taken by an airplane to the ground.
The cosmic ray component in the lowland area of Japan based on the data obtained by high-altitude
observations is approximately 2 nGy/h equivalent at 3 MeV or less, when a 3 in. φ × 3 in.
cylindrical NaI (Tl) scintillation detector is used. Components of greater than 3 MeV varydepending on the conversion factor, but they are typically equivalent to approximately 3 nGy/h(corresponding to 1.2 × 102 cpm). The cosmic ray intensity varies with latitude and altitude, butwhen a 3 in. φ × 3 in. cylindrical scintillator is used, it is calculated as 0.02 nGy/h per 1 cpm
counting rate at above 3 MeV.
[2] Contribution of self-contamination
With regard to the contribution of self-contamination, radiation from 40K contained in the glassmaterials of the detector is the only factor to be considered. The contribution of 40K is, for the
purpose of practical applications, calculated by multiplying the peak counting rate of 1.46-MeV
gamma rays of 40K measured in a shield, which is practically unaffected by 40K from outside, by
0.16 nGy/h/cpm or 0.04 nGy/h/cpm for scintillators of dimensions 3 in. φ × 3 in. or 2 in. φ × 2 in.,respectively (Table B.2). Usually, in the case of 0.9 nGy/h or less, and when glasses with little
potassium are used for the window surface of a NaI (Tl) scintillator and photomultiplier tube, thecontribution of 40K is rarely observed.
Table B.2

Contribution of contamination by 40K in a NaI (Tl) scintillation monitor

Dimensions ofdetector

Conversion factor from the total absorption peakcounting rate of 1.46 MeV(nGy•h−1/cpm)

2 in. φ × 2 in.

1.6 × 10−1

3 in. φ × 3 in.

4.0 × 10−2

B.2.3
Measurement and calibration(1) Considerations of stability
What is most important for this type of monitor is to consider the temperature characteristics andcircuit adjustment. The annual temperature of the environment varies very widely fromapproximately −10 °C to + 35 °C. Therefore, measures against temperature changes should be taken
when a monitor is purchased or before it is used.
Because the temperature characteristics also vary depending on the calculus time constant of thepulse amplifier, it should be used under the pulse processing condition in which a highest stability
can be secured in a range in which the time resolution is not impaired.
(2) Configuration of the monitor
The basic configuration of a monitor is shown in Figure B.8. However, when the discrimination
between natural radiation and facility radiation using information on the energy distribution and theseparation measurement of cosmic ray components are considered, the system shall be configured
with a multichannel analyzer (MCA) or a single channel analyzer (SCA) combined.
65

Detection unit
Heating enclosure

Measuring unit
ADC circuit

Main amplifier

RecorderNaI (Tl) detection unit

Counting rate circuit Dose
rate arithmetic circuit

High-voltage
power source

Printer
Display
unit
Ambient temperature
control circuit of thedetector

Figure B.8

To the circuits

External output

Low-voltage
power source

Block diagram of a NaI (Tl) scintillation monitor by G(E) function method(3) Data recording system
A data collection and processing device is one of the large components of auxiliary equipment of themonitor. In monitoring environmental gamma rays, a long-term continuous observation is usuallyconducted, and the results are analyzed to carry out a separate evaluation of dose contributions bycause. In many of the monitors currently in use, information such as digitally processed pulse-height
distributions, dose rates, and counting rates in arbitrarily set energy ranges are stored on electronic
media. In addition, the information mentioned above is output on a chart paper or something similar
as a backup.
(4) Determination of the measuring time
When a 3 in. φ × 3 in. cylindrical NaI (Tl) scintillation detector is used to obtain the cumulative
counting time for one data lot of measured values, a measuring time of several minutes is sufficient
to obtain adequate accuracy from the statistical changes in the count values. In determining themeasuring time, the temporal patterns of precipitation, fallouts, and the contribution of nuclear
facilities must be considered. In general, natural radiation does not decay faster than the effective
half-life of 35–40 min for descendant nuclides of radon, even when the dose rate rises rapidly owing
to rainfall, snowfall, etc., and the fluctuation is usually relatively slow. On the contrary, thecontribution of facilities usually indicates a rapid rise and fall and a short fluctuation period ofseveral minutes to tens of minutes. However, there are various patterns that cannot be determined as
attributable to facilities. It is desirable to use a device that can appropriately select the measurement
time, considering these characteristics. For example, several to approximately 10 min should beselected to obtain the cumulative counting time for one lot of data, such that the variation of naturalradiation and that of radiation attributable to facilities can be distinguished based on the difference influctuation periods. To reproduce the patterns more faithfully, one method is available in which the
cumulative time is set to approximately 1 to 3 min and a smoothing or average operation is carriedout in the subsequent data processing; it is relatively easy in the currently employed systems torearrange patterns after they are collected by the telemeter.
For cases where the device is used in a traveling survey system, a different concept from the above ispresented in Explanation F.
(5) Sensitivity adjustment and calibration
Because the sensitivity of a monitor using the G(E) function method varies as the gain of theamplifier changes, an accurate ratio of pulse-height to energy (e.g., 1.50 V/0.662 MeV) that issuitable for the measurement system must always be provided. Hence, the gain is adjusted prior tocalibration. This adjustment is generally performed by the manufacturers, and the method is outlined
here for reference.
66[1] Sensitivity adjustment
Set the device to the operating state and adjust the gain such that the pulse-height value
corresponding to the total absorption peak of 137Cs 0.662-MeV gamma rays and the total absorption
peak are established as specified.
At present, information on the pulse-height distribution spectrum can be obtained in many
continuous monitoring systems. This information can be used for gain adjustment. In addition, somesystems are currently available that can automatically perform the adjustment mentioned above byplacing, for example, a source of 137Cs in the vicinity of the detector. In this case, performed process
is not to adjust the gain but to raise or lower the applied voltage for the adjustment. This method ofadjustment is acceptable in many cases; however, when more a detailed adjustment is required, thegain needs to be adjusted.
[2] Calibration
There are two methods of calibration: A: a method using a reference gamma-ray source, and B: amethod using the spectrum-dose conversion operator method to evaluate a reference gamma-ray
source at an arbitrary location. In either case, the monitor is first set to its normal operating stateusing a reference source, prior to calibration.
A Method using a reference gamma-ray source
In this case, the conversion factor is obtained from the dose rate and the counting rate using theinverse square law of distance. The more scattered the rays, the more inaccurate the calibration
becomes; therefore, special attention should be paid to scattered rays during calibration andefforts should be made to maintain the accuracy (for the evaluation of scattered rays, see
Explanation D).
B Method of evaluating the reference gamma-ray dose using the spectrum-dose conversion
operator method
In this case, the effect of the errors in certified values and the scattered rays (these are includedin the evaluation) is small compared with the reference gamma-ray source method. However,
the error of the function used for the conversion and the difference in the conditions between thedetector used in determining the function and the calibrated detector, for example, the differencein absorption due to the difference in the thickness of the probe container, contribute to reducing
the accuracy of the evaluation.
Sensitivity matching and sensitivity confirmation after calibration are all carried out using thereference source that was used for sensitivity adjustment. These sources are preferably available
in more than one type. In addition, when a source is used, the amplifier output is provided to apulse-height analyzer to ensure that there is no drift in the total absorption peak channel.
(6) Measurable dose rate range[1]
G(E) function method
Monitors of environmental radiation are required to be capable of measuring as wide a dose-rate
range as possible, from natural radiation levels to high dose-rate levels in case of an accident. In a“continuous monitor,” which is a fixed type, a NaI monitor and an ionization chamber monitor are
installed together, with each monitor responsible for its respectively suitable dose-rate range; thus, awide range of dose rates of environmental radiation is covered.
The flux density per unit dose rate varies greatly depending on the energy of the gamma rays, and inparticular, the upper limit dose rate saturated owing to the temporal resolution of a pulse at low
energy levels is lower than that at higher energy levels. This is because counting loss is produced
during the dead time associated with the resolution of the pulse amplifier. This characteristic is
related to the detector size, photon energy, and the counting rate (dose rate). The relationship67

between the air absorbed dose rate and the detector is shown in Figure B.9. When the photon energy
is around 100 keV in the cylindrical detectors of dimensions 2 in. φ × 2 in. and 3 in. φ × 3 in., amaximum of approximately 2 μGy/h would be needed in order to achieve an accuracy ofapproximately 10% for the response relative to the air absorption dose rate. The level of switching
to an ionization chamber monitor may be determined by referring to the photon energy and thecounting rate, rather than the dose rate.

Irradiated photon
energy: approximately 100keV
keV
照射光子エネルギー：約100

2.0

A (3”φcylindrical)
Ａ（３″φ円柱形）
cylindrical)
B (3”φ（３″φ円柱形）
cylindrical)
C (2”φ（２″φ円柱形）
D (3”φcylindrical)
Ｄ（３″φ円柱形）
E (2”φcylindrical)
E（２″φ円柱形）

Response
Response

1.5
1.0
0.5
0.0
0

1

2

3

4

5

6

Dose rate (μGy/h)Figure B.9

Example of dose rate dependence in a NaI (Tl) scintillation monitor

Explanation B.3 Silicon semiconductor detector
B.3.1
Principles and configuration of the monitor
The measurement principle of the silicon semiconductor detector is similar to that of the ionizationchamber dosimeter. However, because the information obtained is a pulse, a dose rate can becalculated by multiplying the pulse by a factor used to convert into a dose rate. Note that, owing to themeasurement principle mentioned above, it is not possible to obtain energy information as with theionization chamber dosimeter.
B.3.2
Characteristics of the monitor(1) Dose rate linearity
Owing to the characteristics of the detector, it is suitable for measuring a high dose rate. Therefore,
this offers a measurable dose rate range up to hundreds of mGy/h, which is difficult to measure witha NaI (Tl) scintillation detector.
An example of the dose rate linearity of 10 μGy/h to 100 mGy/h is shown in Figure B.10.
The dose rate linearity of a silicon semiconductor detector is good within a range of ±10% for theirradiation values at all dose rates, and there is no decrease in sensitivity owing to counting loss,
which is observed in a NaI (Tl) scintillation detector.

68

1.20
1.15

Response

1.10
1.05
1.00
0.95
0.90
0.85
0.80
10

100

1000

10000

100000

Irradiation value (μGy/h)Figure B.10 Example of the dose rate linearity in a silicon semiconductor detector(2) Directional characteristics
The directional characteristics of a silicon semiconductor detector vary, depending on theconfiguration of the measurement system, shape of the detector, etc. A graph showing the directional
characteristics of a system is presented in Figure B.11 as an example. This graph shows the
directional characteristics, with the irradiation direction from the effective center of the detector to
the vertical upward direction set to 0°, including the directions up to ±90° in steps of 30°. The
response of each angle is within a range of ±10%, indicating a good response. In addition, this iswithin a range of ±20% of the standard of JIS Z 4325: 2008 and meets the standard.
1.20* Irradiation direction
Vertical

1.15

Response

1.10

Front

1.05

Rear

(View from the right side)

1.00
0.95
0.90
0.85
0.80
-120

-90

-60

-30

0

30

60

90

120

Angle ( )Figure B.11 Example of the directional characteristics in a silicon semiconductor detector

69

(3) Energy characteristics
Figure B. 12 shows the response of each energy normalized by the response of 137Cs 662-keV gamma
rays. The air absorbed dose rate for each energy is approximately 1000 μGy/h.
The response of each energy has decreased by an energy of approximately 50 keV, which isconsidered to be due to the detector cover acting as a shield. Additionally, an excessive response isobserved around 80 keV, which is affected by the photoelectric effect. The overall energy
characteristics are mostly within the range of the standard of JIS Z4325: 2008, except for 50 keV,which was affected by the detector cover acting as a shield.

1.80
1.60

Response

1.40
1.20
1.00
0.80
0.60
0.40
0.20
10

100

1000

10000

Energy (keV)

Figure B.12 Example of the energy characteristics in a silicon semiconductor detector

Explanation B.4

CsI (Tl) scintillation detector

B.4.1 Principles and configuration of the monitor
As with a NaI (Tl) scintillation monitor, the G(E) function method is used to evaluate the absorbed
dose and ambient dose equivalents. In the monitor using a CsI (Tl) scintillation detector, the range ofmeasurable energy varies depending on the type of photodetector it is combined with. The
specifications should be selected in accordance with each application and the features of the monitor.
Table B.3 lists the features of typical scintillators and photodetectors used in gamma-ray detectors.
The main feature of a CsI (Tl) scintillation detector is its higher density compared with that of a NaI(Tl) scintillation detector. This means that the detector has a high sensitivity despite its small size, anda very compact system configuration is possible by combining a photodetector with a compatible
wavelength sensitivity. However, the attenuation constant is larger than that of a NaI (Tl) scintillation
detector. This means that after a gamma-ray photon is incident on the detector and the scintillator
emits light, a long time is required for the light to attenuate. For this reason, when measuring at a highdose rate (a high counting rate), incident photon counting loss occurs during the attenuation. Therefore,
because of possible underestimation, attention should be paid to the handling of the measured value
70when the dose rate (counting rate) is high. Especially in case of an emergency, because it is assumed
that the dose rate is high owing to the contribution from nuclides of the low-energy side, attention
should be paid to the setting of an upper limit of the measurable dose rate, depending on the ratio of
each nuclide that accounts for the contribution of the dose rate. However, owing to advances inelectronic technology, some systems are currently available that have improved the counting
characteristics by reducing the time required for in-circuit processing.
It should be noted that the CsI (Tl) scintillation detector described here is utilized in a system forobtaining spectral information, while in some systems for other applications, energy information
cannot be obtained, and the energy range to be measured may be different from that described here.
Table B.3

Phosphor

NaI(Tl)CsI(Tl)a)

Densitya)(g/cm3)

3.67
4.51

Features of a scintillator and a photodetector
Attenuation
constanta)

Deliquescenta)

(ns)
230
1000

Maximum

Maximum sensitivity

fluorescentwavelength of awavelengtha)

photodetector (nm)

(nm)
Yes
No*3

PMTb)*1

MPPCc)*2

420

500

410
565

Reference 12; b) Reference 13; c) Reference 14*1 PMT: Abbreviation for a photomultiplier tube
*2 MPPC: Abbreviation for multi-pixel photon counter “MPPC” is a registered trademark owned byHamamatsu Photonics Co., Ltd.
*3 Some references suggest that it is slightly deliquescent.

Because this is assumed to be used as a continuous monitor (particularly as a traveling survey system),
this section describes a typical example of a monitor that can also measure the low-energy side, whichis essential in an emergency. The following descriptions are cited from the report on the KURAMA-II
traveling survey system (Reference 15).
When a photon is incident on the CsI (Tl) crystal, the optical signal is amplified via multi-pixel photon
counter (MPPC), a multi-pixel semiconductor photodetection element, which enables highmagnification at low voltage. Then, the pulse-height data or dose rate data processed in the circuit on
the substrate can be output via a USB cable, and when directly connected to a personal computer, thepulse-height spectrum data can be acquired together with the dose rate using a dedicated software
package. Power is supplied via the USB, and the energy range of gamma rays to be measured is from30 to 2000 keV. The configuration of the measuring instrument is shown in Figure B.13. In this system,
the G(E) function method is adopted for calculating the dose rate, with the specification of calculating
the dose rate while compensating the energy.

71Gamma rays

Figure B.13 Example of the configuration of a CsI (Tl) scintillation detector

B.4.2

Characteristics of the monitor(1) Dose rate linearity
Owing to the structure or nature of the measuring instrument, the sensitivity decreases by the effectof counting loss during a high dose rate (a high counting rate), as with the NaI (Tl) scintillation
detector. In the range of 0.2 μSv/h (the ambient dose equivalent rate) to 30 μSv/h at 662 keV of 137Cs,
the response change is approximately ±15%. In addition, when the dose rate increases further, theindicated value decreases by 30% at 100 μSv/h, and by approximately 50% at 200 μSv/h. Figure
B.14 shows a graph of the dose rate linearity.
With the improvement of electronic circuits, the processing speed has increased, and the dose rate
linearity has been improved as a result. The latest characteristics must be determined from thespecifications of the employed equipment.

Sealed source of 137Cs
Irradiation device of 137Cs

Response

After dead time is corrected

Dose rate (μSv/h)

Figure B.14

Dose rate linearity in a CsI (Tl) scintillation detector72(2) Variation in the indicated values
When the variation in the indicated value is evaluated from the mean value (the indicated value) andthe coefficient of variation by repeatedly measuring at each dose rate with a measuring time of 3 s, avalue of less than 15%, which is the standard of JIS Z 4333: 2014, is obtained at a dose rate of
approximately 0.2 μSv/h or more. The relationship between the mean value (the indicated value) andthe coefficient of variation is shown in Figure B.15.

＾Coefficient of variation

ExperimentCalculation

Indicated value (μSv/h)

Figure B.15 Variation in the indicated value in a CsI (Tl) scintillation detector (calculated as 40 cpm
per 0.01 μSv/h)73

(3) Incident directional characteristic
When normalized by the response in the direction of 0°, the sensitivity slightly decreases from therear of the detector (180°) to the direction in which the substrate for data processing and transmissionis placed (270°), but the response change in all directions is in the range of approximately −23%

Response (normalized at 0°)

through +10%. Figure B.16 shows a diagram representing the directional characteristics.

Angle of incidence θ (°)

Figure B.16 Directional characteristics of a CsI (Tl) scintillation detector (this directional
characteristic exists when the detector is placed in a container dedicated to the travelingsurvey system KURAMA-II.)74

(4) Energy characteristics
Figure B.17 shows the relationship between the gamma-ray energy (effective energies are 60 keV of241

Am, 340 keV of 133Ba, 662 keV of 137Cs, and 1250 keV of 60Co) and the response when gamma-ray

photons of each energy are irradiated in a measuring system combined with a CsI (Tl) scintillation
detector and the G(E) function method. Simulation results obtained from Monte Carlo calculations
are also presented. Although a slight decrease in the response of 241Am is observed because theenclosure acts as a shield, relatively good energy characteristics are exhibited at other energies; thus,
this effect is not necessarily considered when using the detector.

Response

ExperimentCalculation

Gamma-ray energy (MeV)

Figure B.17 Energy characteristics in a CsI (Tl) scintillation detector

75Explanation C

Reference gamma-ray source

Explanation C.1 Application of a reference gamma-ray source
A reference gamma-ray source is a gamma-ray source that has a certified irradiation dose rate at a
specified distance (normally 1 m) from the source and well-defined traceability; this is used tocalibrate the monitor.
Explanation C.2 Reference gamma-ray sources and dose rates
Gamma rays in the environment have an energy spectrum ranging from tens of keV to several MeV,
and the dose rate is approximately 50 nGy/h. Considering these facts, the five nuclides, 241Am, 57Co,
133
Ba, 137Cs, and 60Co, described in the main text were selected as the reference gamma-ray sources forthe calibration of a continuous monitor, based on the conditions of having appropriate half-lives andeasy availability. The sources of 241Am, 57Co, and 133Ba were intended for checking the energy
characteristics in the low-energy region, and the radioactivity of approximately 10 MBq was specified.
The reference gamma-ray sources are easy to move when used in a certified device with a display, andthus are convenient for calibrating a monitor installed in the field.
Explanation C.3 Accuracy of calibration of dose rates
For reference gamma-ray sources of 137Cs and 60Co, the accuracy of dose rate calibration isapproximately ±10%, except in special cases, and approximately ±30% for those of 241Am, 57Co, and133
Ba.
Explanation C.4 Obtaining a reference gamma-ray source
Reference gamma-ray sources are supplied from the Japan Radioisotope Association. The Japan
Radioisotope Association receives high-precision primary references from the National Institute ofAdvanced Industrial Science and Technology, and it certifies and supplies reference gamma-ray
sources.
Explanation C.5 Certification of dose rates for gamma-ray sources
C.5.1
Certification by the comparison method
Reference gamma-ray sources are certified by a specific secondary measurement standard calibrated
by a national standard. However, in some cases, the dose rates of other gamma-ray sources can becalibrated by the reference gamma-ray source at the facility and used as a practical reference
gamma-ray source. In this case, the reference gamma-ray source and the gamma-ray source to becalibrated should be of the same nuclide. An intermediary detector is not particularly restricted, aslong as it has good stability and energy characteristics. As shown in Figure C.1, take a certain distance
L from the detector, and use Qs as the indicated value of the reference gamma-ray source. Next,
assume that the indicated value is Q, when substituted by the gamma-ray source to be calibrated.
Assuming that the dose rate at 1 m from the reference gamma-ray source is Ẋ s, the dose rate Ẋ at 1 mof the gamma-ray source to be certified is obtained by the following equation.
̇=

̇

(C.1)

Because the above equation is independent of L, a distance can be selected according to the dose rate.
Measurements should be repeated at least twice and the results should be averaged. When the
gamma-ray source is approximately 10 MBq, the monitor currently in use can be used as anintermediary detector.
76

C.5.2
Certification by the reference measuring instrument
A gamma-ray source whose dose rate is certified by a high-precision ionization-chamber-type
dosimetry meter, which is the reference measuring instrument, can be specified as a reference
gamma-ray source.
See Explanation D for the contribution of scattered rays.

L

Reference gamma-r ay
so urce or gamma-r ay
so urce to b e cer tified

I n ter med iar y d et ec to r

P ap erst a ndFigure C.1

Example of the comparison method for certifying the dose rate of a gamma-ray source

77Explanation D

Countermeasures against scattered rays during calibration

The dose rate ̇ at the time of calibration by a reference gamma-ray source is the sum of the dose rate
̇ d due to direct gamma rays from the source and the dose rate ̇ b due to scattered gamma rays fromthe surroundings.
̇ =

̇d +

̇b

(D.1)

η = Is/(Id − Is)(D.2)

The contribution of scattered gamma rays ̇ b/ ̇ d depends on the distance between the source anddetector, the height of the source and detector from the ground surface (or the floor surface), thesurrounding objects, etc.; within a certain range, the closer the source to the detector, the higher thesource and detector are from the ground surface, and the fewer and farther the surrounding objects, thesmaller is the contribution of scattered gamma rays.
To conduct a calibration, the extent of the contribution of scattered gamma rays should be known.
Additionally, the energy of the scattered gamma rays is generally considerably lower than that of thedirect gamma rays, and the effect of the scattered gamma rays may be seemingly large depending on
the characteristics of the detector. Calibration should be preferably conducted in a state where thecontribution of scattered gamma rays is small.
A continuous monitor may be installed in various situations, but it is assumed in this case that thedetector is installed at a certain height above a wide ground surface. The calibration direction (a line
connecting the source and the detector) is supposed to be parallel or perpendicular to the groundsurface (Figure D.1).
Figures D.2 and D.3 show data for the contribution of scattered gamma rays when the calibration
direction is parallel to the ground surface. The curve depicted in Figure D.2 is obtained
semi-experimentally for the case where the ground surface is concrete, and a 60Co source is used.
Experimental results obtained using a 60Co source under various conditions are also shown in thisfigure. Table D.1 also includes the results for differences between the source and detector heights, aswell as the results for a 137Cs source.
In addition, according to the results of a measurement taken by placing gamma-ray sources at a heightof 1 m from the detector on the roof of the monitoring post, the contribution ratio of scattered gamma
rays was 2.3% for an iron roof and 2.5% for a concrete roof when a 60Co source was used, and 2.6%
and 3.5%, respectively, when a 137Cs source was used.
Available methods for measuring the contribution ratio of scattered gamma-ray sources include theshadow shield method and the spectrum measurement method, among others. Here, the shadow shield
method, which is based on a simple principle, is explained.
In the shadow shield method, as shown in Figure D.4, a lead block is placed between a radiation
source and a detector, and scattered gamma rays from the surroundings are measured by blocking thedirect radiation. The lead block should be thick enough to attenuate the direct radiation (to a level ofapproximately 1/1000). The support base of the lead block should be of a light material, such as
foamed polystyrene (so that it would not act as a scatterer).
Assume that the indicated value without a lead block is Id, and the indicated value with a lead block is
Is; then, the contribution ratio η of the scattered gamma rays to the direct gamma rays is given by:

The scattering from the support base of the lead block should be checked. If it is not negligible, obtainη by the following equation. In other words, if Io is the value indicated when only the support base isplaced, it is expressed as (Is and Id are the same as in equation (D.2)):
78

η=

(

)

(D.3)

Vertical direction

Horizontal directiondSourced

Detector
Sourceℓ

DetectorhhFigure D.1

Placement of the source and the detector during calibration

Co s our ce

Det ect or

Ba r e gr oun d, con cr et e

Dose rat e ofdirect r adiation

Dose rat e ofscattered r ays

Va l ue of Ch i l t on

S p e c t r u m m e a su r i n g m e t h o d
Bar e gr ou nd
Co n cr et e
S hadow shi el d method
S i m p l y p a v e d su r fa c e

Figure D.2

Semi-experimental results of Chilton, et al. (the curve) and the measurement results

79

Co source Detector
WallFigure D.3

Dose rate of
direct radiation

Dose rate of
scattered rays

Bare ground

Contribution of scattered rays in the presence of light-weight block walls

80

Dose rate of
scattered rays
Dose rate of direct
radiation

Table D.1 Contribution of scattered gamma rays by concrete or similar materials
d
h

h
ℓ

0.25

0.5

1.0

2.0

4.0

8.0

16.0

32.0

64.0

0.134
0.132
0.125
0.118
0.113
0.109
0.106
0.105
0.105

0.118
0.102
0.090
0.082
0.077
0.074
0.072
0.071
0.071

0.083
0.068
0.059
0.053
0.049
0.047
0.046
0.045
0.045

0.053
0.042
0.036
0.032
0.030
0.029
0.028
0.027
0.027

0.109
0.115
0.113
0.109
0.105
0.102
0.099
0.098
0.097

0.110
0.099
0.088
0.081
0.076
0.072
0.070
0.069
0.069

0.083
0.069
0.059
0.053
0.049
0.047
0.046
0.045
0.044

0.054
0.043
0.037
0.033
0.030
0.029
0.028
0.028
0.027

137

Cs (0.662 MeV)

1
2
4
8
16
32
64
128
∞

0.002
0.017
0.052
0.087
0.115
0.133
0.144
0.150
0.161

0.007
0.026
0.060
0.094
0.119
0.135
0.146
0.152
0.162

0.024
0.049
0.081
0.109
0.129
0.142
0.150
0.155
0.164

0.062
0.088
0.113
0.132
0.145
0.152
0.156
0.158
0.163

0.107
0.127
0.138
0.142
0.143
0.142
0.141
0.140
0.140

60

Co (1.25 MeV)

1
2
4
8
16
32
64
128
∞

0.001
0.010
0.030
0.052
0.070
0.083
0.091
0.095
0.103

0.004
0.015
0.035
0.057
0.074
0.085
0.093
0.097
0.105

0.014
0.029
0.050
0.069
0.083
0.093
0.099
0.103
0.111

0.038
0.057
0.076
0.092
0.103
0.110
0.114
0.117
0.121

0.073
0.093
0.108
0.116
0.119
0.120
0.119
0.119
0.120

h: height of the detector, ℓ: height of the source, d: horizontal distance between the source and the detector(Reference 16)

Lead block(placed nearly in the middle)(Scattered rays)

SourceDetector

Scatterers (walls, floors, bare ground, etc.)

Figure D.4

Layout of the shadow shield method

81

Explanation E

Considerations of continuous measurement using a transportable
monitoring post

Explanation E.1

Key points to be considered

As described in the main text, when continuous measurement is carried out using a transportable
monitoring post, the following key points are to be considered.
[1] Understanding of the measurement method[2] Temperature characteristics[3] Power source[4] Selection of the installation site
[5] Maintenance and management
Of these points, [3] and [4] are described in the text; thus, [1], [2], and [5] are explained here.
Explanation E.2

Understanding of the measurement method

A transportable monitoring post was originally commercialized for the purpose of having one device
to temporarily act as continuous monitor (a NaI monitor and an ionization chamber monitor) as an
installed station, at a site where continuous monitoring is required in case of an emergency. Thismeans that a wide range of measurable dose rates is required, ranging from the normal low dose rate toa high dose rate for which an ionization chamber monitor is responsible. Therefore, each manufacturer
addresses this problem using a different measuring method and a different detector. Table E.1summarizes the various measurement methods.
Dose rate measurements can be taken over a wide range, depending on which combination of detector
and measurement method is selected. However, the features of such combinations are diverse, andsome of their effects are advantageous, whereas others are disadvantageous. When purchasing a device,
the specifications should be examined, considering the necessity and cost of acquiring energy
information. In addition, the features of the device must be understood, when used, and data should be
evaluated considering the specifications.
Explanation E.3

Temperature characteristics

As mentioned in the text, a temperature controller is not usually equipped. Essentially, a temperature
controller is designed to suppress the change in response due to temperature characteristics, using atemperature compensation circuit. However, these have not yet successfully followed rapid changes intemperature due to direct sunlight, etc. Figures E.1 and E.2 show examples of the temperature
characteristic test results taken by the manufacturer before shipping, and Figures E.3 and E.4 show theeffect of temperature changes during continuous measurements when the device is installed outdoors.
Note that because the devices used for Figures E.1 and E.2, and Figures E.3 and E.4 are not from thesame manufacturer, they cannot be completely compared.
In the NaI detector shown in Figure E.1, the sensitivity tends to be enhanced with a rise in temperatureand decrease when the temperature falls. In the silicon semiconductor detector shown in Figure E.2,82the sensitivity tends to decrease when temperature rises, and increase when the temperature falls. A
similar trend is observed in Figures E.3 and E.4, which show the effect of temperature changes duringcontinuous measurements. However, while the change during temperature characteristic testing isapproximately ±5%, the response to the effect due to a sudden temperature change outdoors may
contribute to a maximum change of approximately 30%.
Therefore, the following points should be considered for the temperature change in a transportable
monitoring post: (1) the tendency of changes in sensitivity is exactly opposite between a NaI detector
and a silicon semiconductor detector, and (2) changes in sensitivity may be greater when exposed to alarge temperature change outdoors, than during temperature characteristic testing (temperature shock).
Explanation E.4

Maintenance and management

As mentioned above, diverse variables are identified because of a “transportable” type. If this system
is installed for a long period of time and continuous measurements are carried out, the following
maintenance methods will be helpful.
(1) Comparative measurement
To confirm that a transportable monitoring post is operating properly, perform measurements at thesame ground level in the vicinity of the detector of the transportable monitoring post, using anothersound measuring instrument (such as a survey meter) that has been calibrated and inspected, andcompare the measured values. However, if there is significant nonuniformity in the distribution ofambient dose rates, caution should be exercised in the comparison.
(2) Inspection and calibration
Inspection and calibration are effective for maintaining the integrity of measuring equipment, but inthe case of a transportable monitoring post susceptible to temperature, inspection and calibration
should be carried out quarterly, considering the seasonal temperature conditions.

83

Table E.1

Various measuring methods and their features
Upper row: Low dose rate areas
Lower row: High dose rate areas

Detector

NaI

NaI(Tl)scintillation detector

NaI(Tl)scintillation detector
NaI +semiconductor
Silicon semiconductor
detector

Semiconductor

Silicon semiconductor
detector

Measuring
methodFeatures• G(E) function
method• DBM method• Energy characteristics are good
• Energy information can beobtained• Pay attention to the lowering ofsensitivity owing to counting loss

• Measurements up to a high dose
rate are enabled using the samemodel of detector• Current system• Dose rate dependency is good
• Energy information cannot be
obtained• Energy characteristics are good
• G(E) function • Energy information can bemethodobtained• DBM method • Pay attention to the lowering ofsensitivity owing to counting loss
• Measurements up to a high dose
rate are enabled• Energy information cannot be• Pulsed system
obtained• Note changes in response owing
to switching of a detector• Low dose rates can also bemeasured by adding elements• Because it is sensitive to cosmic
rays, caution should be exercised
during measurement of a low• Pulsed systemdose rate• Energy information cannot be
obtained• Measurements up to a high dose
rate are enabled84

Reference at 20 °C of temperature
characteristic (%)

Temperature of a constant-temperature bath (°C)

Figure E.1

Example of temperature characteristic test results of aReference at 20 °C of temperature
characteristic (%)

transportable monitoring post (NaI)

Temperature of a constant-temperature bath (°C)

Figure E.2

Example of temperature characteristic test results of atransportable monitoring post (silicon semiconductor)

85

Temperature (°C)

Air Dose Rate (μSv/h)

Temperature (°C)
"Air Dose Rate (μSv/h)"

Figure E.3

Effect of temperature changes during continuous measurement
by a transportable monitoring post (NaI)

Temperature (°C)

Temperature (°C)

Air Dose Rate (μSv/h)

Air Dose Rate (μSv/h)

Figure E.4

Effect of temperature changes during continuous measurement
by a transportable monitoring post (silicon semiconductor)

86Explanation F
Explanation F.1

Considerations of the traveling survey system

Overview

Monitoring using a traveling survey system is a measuring method for efficiently conducting
monitoring over a wide area while moving by using a vehicle loaded with measuring equipment,
which has a function of measuring the environmental gamma-ray dose rate.
From the data obtained by this method, it is possible to know the dose rate distribution mainly on andalong the road, as well as to detect a point with a locally high dose rate (a hotspot) with relative ease.
These data are helpful in assessing the exposure dose across a region, and in facilitating thedecontamination of hotspots.
Explanation F.2

Considerations of conducting monitoring

The following points should considered when conducting monitoring using a traveling survey system.
F.2.1

Measuring time and intervals for collection

Because this monitoring is carried out while moving, unlike an installed monitor, the point of
measurement (the target range) is always changing. To correctly grasp changes in the point of
measurement (the target range), the interval of collecting measured data should be short. This leads tothe acquisition of the mapping resolution to be used in preparing a dose rate distribution map and thedata in which the points are subdivided to find hotspots.
In the monitoring currently conducting by a traveling survey system, the vehicle is not traveling at avery low speed that would impede the traffic of other vehicles, but it is running at a legal limit. Table
F.1 presents the relationship between the elapsed time and the travel distance when traveling at a speedof 40 km/h.

87

Table F.1 Relationship between the elapsed time and the travel distance
Elapsed time (s)Travel distance (m)
1
11
3
33
5
56
10
111
20
222
30
333
60
667
(When travelling at a speed of 40 km/h)
Although the measurement ranges are not necessarily the same, if the detector is installed at a heightof 1 m above the ground in a direction perpendicular to the ground surface, the contribution ratio of
the gamma-ray dose rate from within a radius of 10 m is 60–80%, as shown in Figure F.1. From Table
F.1 and Figure F.1, if the elapsed time is 1–3 s (the measuring time and the interval of collection), thetravel distance is short, and subdivided data can be obtained.
Because there are various installation conditions in each traveling survey system, it is desirable toknow in advance the extent of the contribution ratio to each installation condition.

Figure F.1

Contribution ratio from the surroundings with regard to the ambient dose equivalent at 1
m above ground when natural radionuclides are uniformly distributed in the ground and134

Cs and 137Cs are exponentially distributed (β: 1.0 g/cm2) (Reference 17)88F.2.2

Measurement accuracy

In “F.2.1 Measuring time and intervals for collection,” it was stated that a measuring time of 1–3 s is
sufficient to obtain subdivided data, but if the measuring time is short, the measurement accuracy
naturally decreases. Explanation B.4 shows the coefficient of variation for measurements repeatedly
conducted at each dose rate for a measuring time of 3 s. As a result, it is understood from theexperiences acquired after the Fukushima Dai-ichi NPS accident that, in a would-be target area formonitoring with a traveling survey system, if its dose rate is approximately 0.2 μSv/h, measurements
could be conducted using a coefficient of variation of approximately 15%.
To establish measurement conditions for conducting a monitoring using a travelling survey system,
refer to the description above.
In addition, Figure F.2 shows an example of a comparison of the results from measurements taken atcertain fixed points along the road, under the conditions of 3-s measurement times, for reference. Fromthese results, a good correlation is exhibited between the traveling survey results and the air dose rateof 1 m above ground at some fixed points, and the measured values of 1 m above ground at these fixed
points are approximately 1.2 times higher than those of the traveling survey results. In other words, themonitoring results by the traveling survey system, whose target areas include an area on the road, are
apt to show lower values than the fixed points in the surrounding areas, owing to the “weathering
effect” and the like, which is relatively large on the road. The impact of this effect depends on thenumber of days elapsed since the occurrence of a nuclear disaster. It must be understood that measured
values contain various information as mentioned above.
Incidentally, the following explanation in this section is quoted from “Improvement of measurement
infrastructure and application to actual measurements using a KURAMA-II travelling survey system”(Shuichi Tsuda, et al.), JAEA-Technology 2013 - 037 (Reference 15).

89

Air dose rate at 1 m above ground (Aug. 14 through Sept. 7, 2012)

Traveling survey (Aug. 20 through Oct. 12, 2012)

Figure F.2

Relationship between the measurement results by a traveling survey system(mean values in a 100-m square mesh) and the dose rate around the travel route(Reference 15)

F.2.3

Effect of shielding by the vehicle body

When a monitoring with a vehicle is conducted, the effect of shielding by the vehicle body should berecognized. This section explains the effect of shielding. In addition, when a detector is installed in a
vehicle, the correction coefficient for inside and outside the vehicle must be calculated, which is also
explained.
When a measurement is performed with the detector installed in the vehicle, the measured value isusually lower than that taken outside the vehicle. This is due to the shielding by the vehicle body.
Figure F.3 shows an example of the results of comparing measurements taken inside and outside thevehicle.

90

Dose rate outside the vehicle (μSv/h)

Dose rate inside the vehicle (μSv/h)

Figure F.3

Relationship between dose rates measured inside and outside the vehicle (Reference 15)

From the comparison results, a good correlation is considered to be obtained between the dose rates
inside and outside the vehicle. However, when the dose rate decreases, a larger variation may occur.
Because the magnitude of the influence on the variation in dose rates varies depending on the type of
vehicle and the measurement environment, the influence on the variation in dose rates should beunderstood in advance according to the conditions used for measurement. Figures F.4 and F.5 show theresults of comparisons made between vehicle types, for reference.
To confirm the relationship between dose rates inside and outside the vehicle, select multiple points
with different dose rates, where radioactive materials are considered to be deposited (distributed)
broadly and uniformly, take measurements by installing a detector in the predefined location in thevehicle, measure the dose rate outside the vehicle at a height of 1 m above ground, and compare themeasured values.
In addition, this method of confirming the relationship between the dose rates inside and outside thevehicle can also be applied to cases at a normal environmental level. However, when the measurement
is conducted at the points with low dose rate, the correction coefficient for inside and outside thevehicle can be utilized by increasing the statistical accuracy of the measurement; for example, byincreasing the measuring time, although it depends on the size of the detector.
The influence of the shielding effect on the gamma rays from the downward direction, owing to thedifference in the position where an in-vehicle detector is installed, can be confirmed by setting a
gamma-ray source at the lower part of the vehicle body and performing measurements for the cases
both when the vehicle body is present and absent. In addition, by acquiring a large set of data, not only
from one position, but from various positions where the gamma-ray source is set, the detector can beinstalled at an appropriate position in the vehicle, where the shielding effect is low.
91

Dose rate in a station wagon (μSV/h)

Dose rate in a sedan (μSV/h)Figure F.4

Comparison of dose rates between vehicle types (a station wagon and a sedan)

Dose rate in a minivan (μSV/h)(Reference 15)

Dose rate in a sedan (μSV/h)Figure F.5

Comparison of dose rates between vehicle types (a minivan and a sedan) (Reference 15)

92

Explanation G

Analysis of measurement results

The time variation of the monitor readings is summarized in Table G.1. In other words, the causes ofthe variation in environmental gamma rays are roughly classified into natural phenomena,
radionuclides for medical and industrial purposes, and those attributable to nuclear facilities. The
analysis of measurement results is to extract these variations, examine their causes, determine thecauses of variations, and perform a dose assessment. While the pattern of variations in measurement
results, usually obtained by continuous measurement, is characterized by the rate and periodicity ofvariations, the cause of a variation is often known by its correlation with other information.
However, monitor readings involve essential variations, apart from the following, which are
attributable to statistical phenomena, and they are roughly classified into those related to the meter andthose related to the radiation source (the former includes pure counting statistics).
The analysis of measurement results includes the selection of populations, calculation of mean values
and standard deviations, extraction of outliers, etc.; the selection of populations is particularly
important. In recent years, it has become feasible to determine the outliers by analyzing energy
information, such as a gamma-ray spectrum, obtained concurrently with the measured dose rate.
Explanation G.1

Selection of a population

In selecting a population, the following matters should be considered.
G.1.1

Measuring time and time constant

The measuring time is a time interval that is arbitrarily set to calculate the dose rate from informationsuch as the measured and acquired counting. This is the measuring time that is described in thismeasurement method.
Apart from this, the time constant is defined for the radiation measurement equipment. The time
constant is important when the instantaneous value (a value obtained in real time) being measured isread by the rate meter. The survey meter is a typical measuring instrument, and the NaI monitor andthe ionization chamber monitor, which are used as continuous monitors, also have specified time
constants, in addition to the measuring time. However, the time constant is not directly involved in themeasuring time.
A measured value reaches 95% of the original value at a specified position after a period equal to threetimes the time constant elapses, and it reaches 99% after a period equal to five times the time constantelapses. Figure G.1 shows the relationship of the time change between the time constant and themeasured value, using a survey meter as an example.
Many measuring systems are designed to select several settings for a time constant. If the setting of atime constant is short, the measured value can sense the original dose rate of the position as describedabove. However, a short setting of the time constant means that the variation in measured values
increases, because of a lack of information, such as the counting, obtained by the measuring
instrument at a specified interval of the time constant. In addition, a long setting of the time constantcan suppress the variation in the measured values to be relatively small, in contrast to the above, but
the time until a measured value is obtained increases. Table G.2 presents the variation in the measured
values for each time constant at each dose rate.

93

Table G.1

Variations due to natural phenomena

Cause of variations

Pattern of variations

Time variation in monitor readings
Frequency of variations

Rainfall andsnowfall• Gradual variation during rainfall• Increases and decreases are complicatedly
mixed

Regional differences are produced
approximately 100 times per year

Lightning

Increases rapidly and decreases with ahalf-life of approximately 30 min

Regional variations are identified(often in winter on the Sea ofJapan side)

Snow cover

Shielding effect of snow coverRegional differences are identified

Other
meteorologicalphenomena

Diurnal period due to the inversion layer

Common in winterAbsorption of radiation by ground surface
moisture

Atmospheric nuclear
explosion test

Variation is observed a few days after the
test, and the increase is almost proportional
to the elapsed time.

Nuclear facility

Not constant; especially on the leeward axis,
an increase in dose rate and a short period ofvariation are observed.

Characteristics of themeasuring instrument

Mainly due to temperature change

Failure of themeasuring instrument

An excessive or underestimated value isindicated

Increase in amount≈ 100 nGy/h≈ 50 μGy/y

In some cases, approximately tensof nGy/h at maximum
Reduced by approximately 10 to
30 nGy/h≈ 10 nGy/h

Reduced by approximately 2
nGy/h
The shorter the number of days
elapsed, the greater the increase,and in the case of a few days
elapsed, the increase may be
several times the normal level.

Daily and yearly temperaturechanges94

In some cases, it may range froma few percent to 10%, depending
on temperature.

Relative value

Relative value

Time (× time constant)Figure G.1

Time (× time constant)

Relationship of time change between the time constant and a measured value

Table G.2

Variation in measured values for each time constant and dose rate
τ: time constant
Dose rate (nGy/h)Coefficient of variation (%)τ (3 s)τ (10 s)τ (30 s)

10

77

42

24

20

54

30

17

50

34

19

11

100

24

13

8

200

17

9

5

500

11

6

3

1000

8

4

2

2000

5

3

2

5000

3

2

1

10000

2

1

1

20000

2

1

1

(1 in. φ × 1 in. cylindrical NaI (Tl) scintillation survey meter)

G.1.2

Time interval

The cumulative counting time for one lot of data or the time interval for a data printout must becommensurate with the rate of change in the dose rate of the event in question. The time interval used
for natural phenomena such as rainfall ranges from several minutes to tens of minutes. When thevariation based on nuclear facilities is considered, an interval of as short as a fraction or less isapplied; however, in the case of fallout events, an interval of as long as several times or more isapplied (note that natural phenomena include one-day and one-year cycles). Therefore, the data
extraction interval for continuous measurement shall be specified as 5, 10, or 15 min as a standard, andif a short interval is required, 1, 2, 3, or 5 min shall be applied.
95

In the continuous measurement currently being conducted in Japan, the statistical error of the counting
obtained at this standard interval is 2–5%, which is negligible for the variation to be detected.
However, in the case of a few minutes or less, counting statistics (variations in the numerical values ofthe measurement results) should be considered.
These data are used to determine hourly values for subsequent analyses.
G.1.3
Quantity of data to be included in a population
The more data that are contained in a population, the closer it approaches to a true distribution; thus, it
is desirable that more than dozens of data points are contained. It is easy to obtain such a quantity of
data, except in the case of a population that consists only of precipitation data.
Populations are roughly divided into normal times, precipitation, snow cover, immediately after anuclear test, and others.
(1) During normal times, it may also be useful to separate them by specific time and month,considering daily and annual variations.
(2) Precipitation can be classified into thunderstorm, light rain, typhoon, snowfall, and the nature of
the front, but it is advisable to classify it into thunderstorm, other precipitation, and snowfall. In
addition, because the radiation level decreases during snow coverage, a separate population
should be defined for snow coverage.
(3) If an increase is observed immediately after a nuclear test, a different population should be
prepared. Usually, because the fallout in the early stage of a nuclear test is attenuated at t−1.2,
where t is the elapsed time since the test was conducted, a separate population should be used forthe contribution of the fallout by setting a time interval of a fraction of t, for example, bycollecting the data every one day when five days have elapsed since the test, every two days
when 10 days elapsed since the test, and so on, until the data fit within three times the standard
deviation of the normal value.
Explanation G.2 Extraction of outliers and investigation of causes
An outlier is defined as a value that exceeds three times the standard deviation in the distribution
specific to each population, and the following points should considered when investigating the causeof an outlier.
Normal values, with regard to the distribution of observed values during no rainfall, are usually
expressed as a normal distribution (the distribution of observed values during no rainfall tends tospread around the bottom of the distribution in practice, owing to variations in soil moisture andatmospheric radioactivity (approximately 2 nGy/h at 3.7 Bq/m3 of 222Rn)). However, during rainfall,
the high-dose-rate side is distributed almost in an exponential relationship. The standard deviation σ′during rainfall is generally larger than the standard deviation σ during no rainfall.
The probability that observed values during no rainfall exceed the mean ± 3σ is approximately 0.3%,
and the probability that observed values during rainfall exceed the mean ± 3σ' is approximately 2%.
Even given these considerations, thunderstorms can be cited as a natural phenomenon that still
exceeds the range of three times the standard deviation.
When a monitor gives an excessive (or underrated) indication value, the process of extracting theoutlier and investigating its cause usually starts with inspecting the monitor.
If it is confirmed that the monitor is operating normally, an examination should be made to determine
whether it may be attributed to any natural phenomena such as precipitation. In doing so, compare the
distribution with that of a population consisting of the data in normal times including precipitation,etc., or that of a population consisting of only the data of precipitation, if such data are available andsufficiently accumulated. The contribution of precipitation to the annual dose in Japan isapproximately 10 μGy.
96

It is also conceivable that cosmic ray neutrons cause an error in the software of the semiconductor
device.
If the causes have not yet been determined through the above processes, the influence of the nuclear
facility should be suspected. In this case, examine the weather conditions (wind direction, wind speed,
atmospheric stability, etc.), and review the correlation between the indicated values of multiple
monitors located in different directions relative to the stack position of the facility, in order todetermine whether the indicated values of the monitors located in the downwind direction are
excessive. Needless to say, it would be preferred if information on the operational status of the facility
could be obtained.
When it turns out that the cause is a plume from the stack, a relatively short cycle of fluctuation isoften observed; therefore, it is useful for determining the causes to know the fluctuation pattern;
however, to do so, the time interval of data extraction must be shorter than that for standard values. In
addition, when the outlier is attributable to the facility, the energy distribution of gamma rays is
usually different from that of normal times, and energy information may act as an influential factor.
If the cause is still unknown after the above procedure is executed, records must be kept anddocuments must be prepared for later examination.
Explanation G.3 Examples of variations in the environmental gamma-ray dose rate
G.3.1
Variations in dose rates due to natural changes in weather conditions, etc.
The discussion above is based on a statistical approach for the variation in the environmental
gamma-ray dose rate, but in this section, some examples of variations typically observed in practice
are given for reference.
It is well known that the environmental gamma-ray dose rate increases temporarily during rainfall andsnowfall.
This is due to the following two phenomena: one is called “washout,” in which the radiation dose rate
temporarily rises when descendant nuclides of radon, an inert gas floating in the atmosphere, fall on
the ground surface along with raindrops or snow; the other is called “rainout” in which, at the stage ofgenerating a cloud that brings about rainfall or snowfall, the cloud takes in radon and descendantnuclides of radon on the upward current, and raindrops containing radon and its descendant nuclides
fall on the ground surface, thereby temporarily raising the dose rate.
In addition, when there is a large amount of rainfall or snowfall, a temporary decrease in the dose rate
may be observed (in some cases, the effect of snow shielding may be prolonged), because gamma rays
emitted from the earth crust are shielded by the moisture retained on the ground surface. Figures G.2
and G.3 show examples of variations observed in the dose rate due to meteorological phenomena.
After the Fukushima Dai-ichi NPS accident, it has been confirmed that, at the site with extremely highradioactive cesium concentrations deposited on the ground surface, the decrease in the dose rate due tothe shielding effect of the ground surface water is more remarkable than the increase in the dose rate
due to the fall of radon, etc., even in a small amount of rainfall. In the case of variations caused bynatural phenomena in addition to those caused by weather such as rainfall, the changes in response due
to changes in temperature have been confirmed. Refer to the description in Explanation E.3 forexamples that depend on temperature changes.

97Bi-214 (natural)

RainfallRainfall (mm)Air dose rate (μSV/h)

Air dose rate

Date and time

Figure G.2(These data are calculated as 1 μGy/h = 1 μSv/h.)

Example of an increase in the air dose rate due to rainfall

Bi-214 (natural)

RainfallRainfall (mm)Air dose rate (μSV/h)

Air dose rate

Missing of rainfall due to snow cover* During 12:00 of Feb. 8 to 21:00 of Feb. 9, measurement of
rainfall was missed due to snow cover.

Figure G.3

G.3.2

Date and time

(These data are calculated as 1 μGy/h = 1 μSv/h.)

Example of a decrease in the air dose rate due to snow cover

Variations in dose rates due to changes in the surrounding environment

Changes in the surrounding environment can bring about various effects on the measured values of acontinuous monitor.
During monitoring in normal times, a temporary decrease in the dose rate due to the shielding effect ofvehicle(s) parking near the continuous monitor, changes in the dose rate due to the construction of new
concrete buildings, changes in the dose rate due to asphalt laying, etc. may be produced. Figure G.4
shows examples of such variations.
It can be seen from this figure that the dose rate also varies with changes in the surrounding98environment, such as (1) soil (lawn), (2) asphalt, (3) concrete, and (4) other buildings. This is causedby the concentrations of natural radionuclides contained in materials used in the surrounding
environment. Therefore, even if the surrounding is “(1) soil,” it does not necessarily exhibit the same
trend in dose rate, and the dose rate also varies depending on the concentrations of naturalradionuclides contained in the soil. Incidentally, granite (granitic rocks) is often used for stonemonuments and similar structures built in shrines and temples that fall under “(4) other buildings.” Itis generally known that such granitic rocks contain relatively high concentrations of naturalradionuclides, and this effect may result in a slightly higher dose rate.

On the groundTokyo Station

Akasaka-mitsuke
StationOmotesando Station

Sangenjaya Station

Dose rate (nGy/h)Approx. 11 km
Yaesu Underground
Shopping District

SubwayStationGamma-ray dose rate

Cosmic ray dose rate

Time

Approx. 3 km

Meiji-jingu Shrine

On the turf

Otorii

Otorii

Gamma-ray dose rate

Jingu-bashi

In front of themain shrine

Harajuku gate

Harajuku gate

Dose rate (nGy/h)Yoyogi Park

Cosmic ray dose rate

Time

Figure G.4

Examples of variations in dose rate due to changes in the surroundings (Reference 18)

In addition, after the Fukushima Dai-ichi NPS accident, possible outcomes are a decrease in the dose
rate due to decontamination, temporary increase in the dose rate due to temporary placement of
decontamination waste in the vicinity of the continuous monitor, and temporary increase or decrease in
the dose rate due to the installation and operation of a plant facility for treatment of decontamination
waste liquid near the continuous monitor.

99

G.3.3

Variations in dose rate due to other factors(1) Radioactive materials for medical use
At present, in Japan, radioactive materials are utilized in medicine by administering to the body andconfirming their behaviors. As a result, some cases were reported in the past in which the dose rateincreased when a person to whom medical radioactive material had been administered passed
through in the vicinity of a continuous monitor. Table G.3 lists the major gamma-ray emitting
nuclides used in medical applications. In addition, Figure G.5 shows a case of detection due to the
influence of a person to whom medical radioactive material had been administered.
Figure G.5 shows the gamma-ray spectrum detected when a user with medical 67Ga administeredapproached a continuous monitor (a NaI monitor). 67Ga emits photons of multiple energies, as
explained in Table G.3. In Figure G.5, a total of four peaks around 95 keV can be seen. In such a case,the nuclides corresponding to the respective gamma-ray photon energies can be confirmed for
identification.
(2) Nondestructive inspection
For nondestructive inspection, industrial gamma-ray equipment is used for gamma-ray transmission
testing. The major gamma-ray emitting nuclides in wide use are 192Ir (0.317 MeV, etc.), 60Co (1.332
MeV, etc.), 169Yb (0.198 MeV, etc.), and others.
In addition, X-ray generators are sometimes used and a continuous monitor that is sensitive to X-rays
often detects the effects. Figure G.6 shows an example of detecting the effects in a nondestructive
inspection.
Some industrial X-ray generators conduct nondestructive inspections by irradiating X-rays in alow-energy region (tens of keV to 300 keV). The spectrum shown in Figure G.6 is obtained when
X-rays in this low-energy region are detected. The energy of X-rays depends on the tube voltage
applied to the X-ray tube in the X-ray generator, and the value varies depending on the purpose ofeach usage, among other factors. In addition, the energy distribution of the bremsstrahlung used forthis nondestructive inspection is called a continuous spectrum, which has a wide energy distribution
without a sharp peak such as that of gamma rays.
It should be noted that, unlike the bremsstrahlung, “the characteristic X-ray,” which is often
mentioned when performing a gamma-ray spectrometry, shows an energy distribution with a peaksimilar to that of gamma rays.
Apart from the major variation factors described here, various other factors have been identified.
In environmental radiation monitoring, it is necessary to investigate the factors from the data
showing variations and to confirm whether the contributions are due to artificial radionuclides.
To do so, the physical conditions at each point must be constantly acquired, and efforts should bemade to obtain the accompanying weather data as much as possible. In addition, if a system capableof acquiring energy information is provided, the factors of variations can be quickly identified byclosely monitoring the information along with the dose rate.

100Table G.3

Examples of major gamma-ray emitting nuclides used as radiopharmaceuticals(References 19 and 20)

NuclideHalf-life

Energy (MeV) and emission rate of major gamma
rays

51

Cr

27.70 d0.320 - 9.92%
0.0050 - 19.7% V-Kα

67

Ga

3.261 d0.0933 - 39.2%
0.185 - 21.2%
0.300 - 16.8%

81

Rb

4.576 hDaughter

81m

Kr

0.0086 - 50.3 %Zn-Kα

0.0096 - 6.8 %Zn-Kβ0.190 - 64.0%

81m

Kr

0.466 - 23.2%
0.510 - 5.3%

99

Mo

99m

Tc

111

In

65.94 h0.181 - 6.0%

Daughter 99mTc

0.739 - 12.1%

6.01 h

0.141 - 89.1%

Daughter 99Tc

0.0184 - 6.1 % Tc-Kα

2.805 d0.171 - 90.2%
0.245 - 94.0%

123

I

133

201

I

Xe

Tl

Cd-Kα

0.0261 - 14.5%

Cd-Kβ

13.27 h0.159 - 83.3%

Daughter

0.0275 - 70.7% Te-Kα

123m

0.0310 - 16.0% Te-Kβ

8.021 d0.284 - 6.1%

Daughter

0.364 - 81.7%

131m

Xe

0.637 - 7.2%

5.243 d0.0810 - 38.0%

Te

131

0.0232 - 67.8%

72.91 h0.0310 - 40.3%

Cs-Kα

0.0350 - 9.4%

Cs-Kβ

0.167 - 10.0%
0.0708 - 73.7%

Hg-Kα

0.0803 - 20.4%

Hg-Kβ

0.00999 - 46.0%

Hg-L

101

Count number― When approaching― Background

Energy [keV]
Figure G.5

NaI spectrum obtained when a person with medical radioactive material (67Ga)
administered approaches the detector

Measurement of the spectrum
for 6 min during nondestructive
inspection
Measurement of the spectrum for 6 minwithout nondestructive inspection

Figure G.6

Example of detection of the effect by X-ray generator during nondestructive inspection

102

Explanation H

Integrity check of a continuous monitor by comparative measurement

Explanation H.1 Overview
Continuous monitors constantly perform measurements under conditions such as wearing weather,changing temperature, high humidity, and in some cases, even in poor environments. However,considering the significance of environmental radiation monitoring, the integrity of the measurement
system must be maintained even under poor conditions. In addition, it is essential to maintain the
integrity not only for the purpose of monitoring any abnormalities attributable to nuclear facilities, butalso of utilizing the data obtained from continuous monitors for the exposure dose assessment after an
accident, in order to help in assessing the exposure dose of the residents living in the vicinity. In thissection, some examples are given to confirm whether an appropriate measurement value has beenobtained as a representative of the relevant field, rather than identifying a clearly unsound situation
including unmeasurable conditions and physical damages.
Explanation H.2 Method
To confirm whether the measured values obtained by the installed continuous monitor appropriately
indicate the dose rate in the field, take measurements for comparison at the same place, using anothermeasuring device that is properly adjusted. However, it is not easy to remove a continuous monitor
that has already been installed. Therefore, the comparison and confirmation can be performed byanalyzing data obtained through a comparative measurement using the following method.
(1) Selection of measurement points
Draw a straight line along at least three axes, such that it passes through the center of the continuous
monitor installed. Select measurement points on this straight line, at an interval of approximately
0.25–0.5 m from the center of the continuous monitor.
(2) Measurement
Use a sound NaI (Tl) scintillation spectrometer with gain adjustment and electrical zeros confirmed
and adjusted, and take measurements at the measurement points selected in (1). In doing so, a type ofthe measuring device that is capable of assessing the dose rate contribution of cosmic rays, with
self-dose required, should be employed.
The selection of the points (1), and the results of the measurements taken at these points (2) are
shown as examples in Figures H.1 and H.2.
(3) Analysis
The following explains how to perform an analysis based on the measurement results obtained.
[1] Calculate first, second, and third approximate equations for each axis.
[2] From each of the calculated approximate equations, obtain the value of the point where thecontinuous monitor is installed, through interpolation.
[3] For each axis, adopt an approximate equation whose error in the dose rate is smallest, which
is obtained through interpolation at the point where the continuous monitor is installed.
[4] From the approximate equation adopted for each axis, obtain a weighted average of the values
at the point where the continuous monitor is installed.
[5] Subtract the dose rate of the cosmic ray contribution and the self-dose from the obtained
weighted average value, and calculate the pure environmental gamma-ray dose rate at the
point where the continuous monitor is installed.
[6] Add the dose rate of the cosmic ray contribution and the self-dose, which are included in themeasured values of the continuous monitor, to the calculated environmental gamma-ray doserate, and compare it with the measured values of the continuous monitor.
103

Explanation H.3 Availability of comparative measurement
Comparative measurements are conducted on the assumption that the environmental gamma-ray dose
rate does not fluctuate while measurements are being taken in the vicinity. Therefore, they must be
carried out in a situation where no variations in the dose rate due to rainfall or similar phenomena andno changes in the surrounding environment (movement of vehicles, etc.) are observed.
In addition, as a condition for comparison, the dose rate of the cosmic ray contribution and self-dosemust be evaluated. Therefore, the specifications of the measuring equipment should be fully
understood.

Generator

Axis A
Building

Panzer mast

IonizationchambermonitorAxis BNaI monitor

Axis C
Figure H.1

Example of measurement points during comparison and schematic diagram (circled
numbers indicate the numbers of the measurement points of each axis)

104

Dose rate (nGy/h)Axis A

Distance from the detector (m)Dose rate (nGy/h)Axis B

Distance from the detector (m)Dose rate (nGy/h)Axis CDistance from the detector (m)Figure H.2

Measurement results at measurement points shown in Figure H.1

105

Explanation I
Explanation I.1

Contribution ratio of the dose rate during an accident

Overview

The main contributions of the air dose rate are the gamma rays from natural radionuclides present in
the Earth’s crust in normal times and the natural radionuclides contained in surrounding structures.
However, in case of an emergency due to a nuclear accident, the contribution ratio will change. Thissection explains the contribution ratio that was observed when dose rates increased during the
Fukushima Dai-ichi NPS accident.
Explanation I.2

When a radioactive plume passes throughFigure I.1 is a trend graph showing the air dose rates from March 15, 2011, when the Fukushima
Dai-ichi NPS accident occurred, to March 31, 2011.
On or around March 15, when the radioactive plume shown in Figure I.1 is believed to have passed
through, it can be seen that artificial gamma-ray-emitting nuclides such as 133Xe (approximately 5days), 132I (approximately 2 hours), 131I (approximately 8 days), 132Te (approximately 3 days), 134Cs(approximately 2 years), and 137Cs (approximately 30 years) are present in the order of theircontribution to the dose rate under the meteorological conditions with no rainfall. The periods inparentheses indicate the physical half-lives. Except for 134Cs and 137Cs, most of these nuclides haveshort half-lives of ten days or less.
Figure I.2 shows the contribution ratio of each nuclide to the dose rate rose on or around March 15,
when radioactive plumes released into the atmosphere by pressure vents or hydrogen explosions were
supposed to pass through at the Fukushima Dai-ichi NPS of TEPCO.

Air dose rate

RainfallRainfall (mm)Air dose rate (μSV/h)

Bi-214 (natural)

Date and time

Figure I.1(These data are calculated as 1 μGy/h = 1 μSv/h.)

Air dose rate and contribution dose rate of each nuclide (March 2011) (Reference 21)106Natural, 1.5%

Figure I.2

Example of dose rate contribution ratio when a radioactive plume passes throughAs can be seen from this figure, the dose rate contribution ratio of 133Xe (81 keV) accounts for morethan 90% of the total when a radioactive plume passes through. Therefore, in case of an emergency,
especially at the early stage of an accident in which a radioactive plume passes through, the energy
characteristics in the low-energy region should be acquired in advance, because of the high dose rate
to be produced by the low-energy gamma-ray-emitting nuclides.
Explanation I.3

During (or after) fallout

As explained in Explanation I.2, it is conceivable that a radioactive plume passed through during theearly stage of the accident. After that, as shown in Figure I.1, radioactive cesium and others, which had
been radioactive nuclides suspended in the atmosphere, were deposited on the ground surface together
with raindrops owing to the rainfall of March 21, (that is, the fallout). Figure I.3 shows thecontribution ratio of the dose rate when this fallout was generated.

Natural, 2.5%

Figure 1.3

Example of dose rate contribution ratio during fallout
107

The contribution ratio, after 60 days have elapsed since the fallout is shown in Figure I.4.

Natural, 3.5%

Figure I.4

Example of the dose rate contribution ratio after 60 days have elapsed since fallout

Figure I.3 shows that the contribution ratios of radioactive cesium and radioactive iodine are large, andFigure I.4, showing the condition after 60 days have elapsed since the fallout, tells us that thecontribution ratio of radioactive cesium is overwhelmingly large. This is because the release from thepower plant has stopped, and the contribution ratio of the radionuclides with relatively longer
half-lives has increased, because they have been attenuated by the physical half-life and theweathering effect of the deposited radionuclides.
These results show that, in case of an accident, the situation is divided into three stages: (1) when aradioactive plume passes through, (2) when fallout is generated, and (3) after the release has stopped,and dose rate contribution ratios commensurate with each situation are identified. Therefore, it isextremely important for those who are engaged in continuous monitoring to understand the evaluationof the measured values corresponding to the gamma-ray energy of each contributing nuclide and thatthe data contain such information.

108

Explanation J
Explanation J.1

Creation of a response function

Meaning of a response function

When an unfolding is performed to observe the output pulse-height distribution from the detector andobtain an energy spectrum of radiation incident on the detector (this method focuses on photons), theshape of the response from the detector to monoenergetic radiation (the response function) should beknown. The agreement between the response function and the response of the detector considerably
affects the analysis result.
Assuming that the spectrum of the incident photon is expressed as N(E) (E is energy) and the numberof pulses observed between the pulse height h and h + ⊿h is P(h) ⊿h, then, given that the response
function is K(E,h) and the efficiency of the detector is ε(E), the following equation is established:P(h)⊿h = ⊿h ∫( ) ( , ℎ) ( )(J.1)

For a monoenergetic photon, N(E) is regarded as a delta function. In this case, ε(E) is a constant, andthe above equation changes as follows:P(h)⊿h = ⊿h ∫( ) ( , ℎ) ( − )P(h) = ε( ) ( , ℎ)

(J.2)
(J.3)

A response function is obtained by dividing the pulse-height distribution by ε(E).
Further, if ε(E)K(E,h) is collectively expressed as R(E,h) in equation (J.1), the pulse-height
distribution P(h) relative to the spectrum N(E) of the incident photon is expressed as follows:P(h)⊿h = ⊿h ∫( , ℎ) ( )(J.4)

R(E,h) is the probability that an incident photon with energy of E produces a pulse of height h in thedetector.
When the pulse-height distribution obtained by measuring a monoenergetic photon is divided by theflux density incident on the detector, a pulse-height distribution when monoenergetic photons of unit
flux density are incident on the detector is obtained, and this represents R(E,h); thus, it is easy tounderstand the meaning of the latter response function.
In equation (J.1), K(E,h) represents a distribution in which the area of the pulse-height distribution is
1 for every different photon energy E.
A response function is specific to the radiation source, as well as the configuration and shape of thedetector. Table J.1 illustrates some examples of relatively common response functions or response
matrices. Apart from those shown in the table, many other response functions have been created.
Heath developed and compiled a method of interpolating a response function for any energy based on109

a response function that is experimentally determined using a specific radiation source and a specificdetector configuration (References 22, 23, 24, and 25).
Explanation J.2

Creation of a response function

A response function is specific to the arrangement of the detection system and the source. There are
two methods to determine a response function: (1) the experimental method and (2) the Monte Carlosimulation method. The former is the method represented by Heath, as mentioned before. In the lattermethod, additional conditions such as the equipment in use and its geometry must be considered;however, such procedures are cumbersome and a simulation is often performed under the condition
that photons are incident on a bare detector.
(1) Experimental creation method
This method starts with obtaining an output pulse-height distribution of the detector for photons ofvarious monoenergetics. This is advantageous because the contribution to the pulse-height distribution
under the conditions specific to the detection system used, such as forward- and backward-scatteredrays, can all be included in the response function. If the entire detection system, including the shield,
is always kept in the same condition, they can be regarded as one detector. To evaluate a photon
spectrum in the natural environment, a group of pulse-height distributions by monoenergetic photons
should be measured under the condition of a broad incident beam. Photon radiation sources ofmonoenergetic photons commonly used to create a response function are listed in Table J.2. It is notnecessary to measure all of these sources. In taking measurements, a radiation source that is structured
to emit as few scattered rays as possible should be used.

110Table J.1

Shape

(a)
3 in. φ × 3 in.
cylindrical

Response functions and response matrices of a NaI (Tl) detector

Producer or userorganization

Number ofrows andcolumns

Conditions (energy range, bin width,
Referencestarget, etc.)( I ) Heath-

0.335–2.75 MeV; interpolate a response
function of any energy

a)

(II)

Zerby–Moran-

0.679–6 MeV; perform a Monte Carlocalculation for an arbitrary pulse-height
distributionb)(III) Berger–Seltzer-

2 - 20 MeV; perform a Monte Carlocalculation of a pulse-height distribution
for each 1 MeVc)(IV) Kyoto Univ.,
Engineering

20 × 20

0–1.44 MeV; 72 keV/bin, equal division;for transmission spectrum studyd)(V) Nagoya
Institute ofTech., Lab.

22 × 22

0–2.61 MeV; unequal division by
monochromatic energy peak width; forthe natural environmente)(VI) Nagoya
Institute ofTech.

46 × 46

0–9.2 MeV; 200 keV/bin, equal division;for the environment in the vicinity of thereactorf)(VII) JAERI-

Response matrixes of arbitrary width canbe calculated for gamma rays of up to 10
MeV.; intended for various shapes andsizes of detectorsg)( I ) RIKEN

Continuous,with aninterval of
10 keV

0–3.0 MeV; 10 keV/bin, equally divided
BG (intrinsic background, cosmic ray
distribution is subtracted); for the naturalenvironmenth)(II) JAERI

60 × 60
295 × 295

0.05–3 MeV; 50 keV or 10 keV/bin,equal division; for the naturalenvironment
0–0.1 MeV; 20 keV/bin, 0.1–3 MeV 100
keV/bin; for the natural environmenti)(b)(III) Reactor of
3 in. φ spherical
Kyoto Univ.

34 × 34(IV) Nagoya
Institute ofTech., Lab.

22 × 22

0–2.614 MeV; unequal division by
monochromatic energy peak width; forthe natural environmentk)(V) Nagoya
Institute ofTech.

22 × 220.01–3 MeV; equal division of alogarithmic pulse-height axis; for thenatural environmentl)( I ) Hitachi

14 × 14m)(II) JAERI

60 × 60

0–1.4 MeV; 100 keV/bin, equal division;
inverse matrix; for the study of scatteredrays
0.05–3 MeV; 50 keV/bin, equal division(c)
2 in. φ × 2 in.
cylindricalj)n)Referencesa): Reference 22 R.L. Heath; IDO-16880-1 (1964)b): Reference 26 C.D. Zerby and H.S. Moran; ORNL-3169 (1961)c): Reference 27 M.J. Berger and S.M. Seltzer; Nuclear Instr. Meth., 104, 317–332 (1972)d): Reference 28 T. Hyodo and F. Makino; Memoi. Fac. Engin., Kyoto Univ., 14, 291 (1962)e): Reference 29 湊進; 名工試報告, 27, 384(1978)f): Reference 30 Y. Nakashima; Thesis. Fac. Engin., Nagoya Univ., (1980)g): Reference 31 森内茂,堤正博,斎藤公明; 保健物理, 42, 71–83, (2007)h): Reference 32 M. Okano; Natural Radiation Environment III, Symposium Series DOF51(CONF-780422) 867 (1980)i): Reference 33 S. Moriuchi; Personal Communicationj): Reference 34 I. Urabe, T.T. Sujimoto, K. Yamazaki and K. Katsurayama; Jour. Rad.

111

Res., 19, 163 (1978)k): Reference 35 S. Minato; Personal Communicationl): Reference 36 明野吉成: 卒業論文; 名大工, 原子核工学科, Mar. (1982)m): Reference 37 石松健二; 日本原子力学会誌, 4, 24 (1962)n): Reference 38 S. Moriuchi; Personal Communication

112

Table J.2

Nuclides and gamma-ray energies that can be used to create a response function (thecircles indicate commonly used nuclides)
Nuclide Energy (MeV)

Nuclide Energy (MeV)○ 241Am

0.060

○

129m

Te

57

Co

85

Sr

0.513

0.109

91m

Y

0.555

0.122

○ 137Cs

0.662

Sc

0.155

95

Nb

0.764

Ce

0.166

○ 54Mn

0.835

Ru

0.213

92

0.930

Xe

0.250

86

○ 203Hg

0.279

51

0.320

41

Ar

1.290

115m

In

0.335

42

K

1.510

195

Au

0.411

26

Al

1.780

7

0.478

24

Na

1.369, 2.754

○

47
139
97
135

○

Cr

Be

Nb
Rb

○

65

Zn

1.080
1.114

The resulting group of pulse-height distributions for monoenergetic photons is normalized relative tothe number of incident photons, taking into account the efficiency of the detector. That is, the groupof pulse-height distributions obtained by measurement is normalized such that every area becomes 1,
and then the efficiency ε (E) = 1 − exp – ( )ℓ is applied to each pulse-height distribution,

considering the energy of each incident photon. Here, μ(E) is the linear attenuation coefficient of thedetected body for the photon of energy E, and ℓ is the average passing distance of photons in thedetected body. The response function for an incident photon of arbitrary energy can be obtained bygraphical interpolation, using the group of pulse-height distributions for monoenergetic photons thatwas previously obtained.
Heath developed a graphical interpolation method and created a computer program to perform this; its
principle is as follows. When a group of pulse-height distributions for monoenergetic photons
measured under the same condition is arranged in three dimensions and covered with a smooth curved
surface, a response curved surface is formed.
The total energy absorption peak of the response curved surface thus obtained is interpolated by thefollowing Gaussian distribution.

=

exp

−

(

)

(J.5)

In this equation, y0 is the peak height of the Gaussian distribution and b0 is an amount related to thehalf-width, which can be determined experimentally as a function of the energy Eγ of the incident
photon. Next, with respect to the pulse-height distribution due to the Compton effect, the energy of
the Compton edge, that is, the maximum energy generated by the Compton effect, is obtained fromthe following equation.
113

=

−

/

1+

(J.6)

In addition, the energies of the peak Eb due to backscattering, the single escape peak EPS of theannihilation photon due to the reaction of pair production, and the double escape EPD are determined
as follows:

=
=

=

/

1+

(J.7)

− 0.511(MeV)(J.8)− 1.022(MeV)(J.9)

The pulse-height distribution is divided into five intervals of L1, L2, ..., L5 according to the positionsof EC, Eb, EPS, and EPD (see Figure J.1). Naturally, in the case of Eγ < 1.02 MeV, it can be divided into
three intervals of L1, L2, and L3 (see Figure J.2). The pulse-height distribution is expressed by thefollowing polynomial in each interval.

ｇ( ) =++∑

sin(J.10)

That is, the pulse-height distribution g(X) of the interval Li can be expressed as above. X is the peakvalue of the pulse. In this equation, M is the number of data points in the interval, and N is the numberof terms in the development, which is usually approximately M/2. The entire pulse-height distributionof the measured monoenergetic photons is fitted using the above equation. The coefficients a, b, andbK are interpolated with respect to Eγ for each division interval Li. By using the obtained values of a,b, and bK and the previously obtained b0 and y0, the pulse-height distribution of the incident photons
of arbitrary Eγ relative to the detector can be calculated by equations (J.4) and (J.10).
To display these groups of pulse-height distributions in a matrix, i, j, and the element Rij of the
response matrix relative to any given energy Ej can be determined by the following equation:

=∫

( ,

)

(J.11)A response matrix should be created to have an appropriate number of rows and columns, according
to the shape of the photon spectrum, the analysis method to be applied, the data processing capability,
etc. When analyzing a continuous spectrum, it is not necessary to have many rows and columns,
whereas, when a spectrum consisting of a large number of monoenergetic photon groups is expected,
many rows and columns are needed.

114

Counting rate (an arbitrary scale)

Backscattering rays

Escape peak

Scattering gamma rays at
the Compton edge due tothe materials in the vicinity
Contribution of annihilation
photons generated due to pair
production in the shield

Compton edge

Small-angle
scattered ray from aβ absorber

Pulse-height value

Response of a 3 in. × 3 in. NaI (Tl) detector to 1.78-MeV gamma rays

Counting rate (an arbitrary scale)Figure J.1

Pulse-height value

Figure J.2

Response of a 3 in. × 3 in. NaI (Tl) detector (in the case of below 1.02 MeV)

115

(2) Monte Carlo simulation (References 39 and 40)
When a photon enters the detector, a secondary electron is generated through interaction with thedetected body. The secondary electron expends its energy in the detected body. The main interaction
between the photon and the detected body is caused by the action of photoelectric effect, Compton
effect, and electron pair generation. The Monte Carlo simulation uses random numbers to track thehistory of an incident photon in the detected body. A secondary electron is generated every time aphoton collides, and the energy of the secondary electron is lost through ionization and excitation.
However, when the incident photon has an energy of several MeV or less, the range of the secondary
electron generated in the detected body is short; thus, all the energy of the secondary electron istreated as absorbed by the detected body. However, when the photon energy is high, secondary
electron tracking is required. The scattered photons are tracked again until they lose all their energy inthe detected body, fall below a certain cutoff energy, or escape from the detected body.
When the energy of the generated secondary electrons is large, the Monte Carlo simulation program
becomes more complex, including two subprograms, one that track photons and one that tracks
secondary electrons.
Figure J.3 shows an outline of a simple simulation (without electron tracking) for determining a
response function using the Monte Carlo simulation.
First, the following is given as input data: the shape of the detector (radius and length), density,
sections of energy, number of rows and columns of the response matrix, resolution of the detector,
value of the attenuation absorption coefficient owing to each interaction, constant associated withelectron tracking (if required), number of histories, and so on. Because each process is probabilistic,
random numbers (a uniform random number that is usually (“0”,“1”)) are used to track each process.
As the initial condition for a photon, determine the position and direction of incidence relative to thedetector. Next, determine the distance that the photon travels through the detected body and thecoordinates of the interaction point before interaction occurs. If the interaction point is outside thedetected body, return to the beginning and start tracking the next photon. If the interaction point is
inside the detected body, determine whether interaction is a photoelectric effect, a Compton effect, oran electron pair generation. When a photoelectric effect occurs as the interaction, determine thesection of energy for recording the amount of the total energy of the photon, which is assumed asabsorbed by the detected body, and proceed to track the next photon. When a Compton effect occurs,
determine the energy and scattering direction of the scattered photon. In this case, the difference
between the energy of the original photon and the energy of the scattered photon is specified as theabsorbed energy. The scattered photon is further tracked in the scintillator as before, to determine thetravel distance and the coordinates of the next interaction point, until they fall below the cutoff energy
or escape from the detector.

116

Input data

Number of histories
Initial condition (incident position
and incident direction of the photon)

Distance photon travels in the scintillator(determine an interaction point)
Boundary condition (whether inside

OutsidePhoton escapes from the detector

or outside the detector)
InsideDetermine theinteraction

Photoelectric effectElectron pair
production

Compton effect
Scattered photon(E > Ec)

Ec: cutoff energy

Photon was absorbed

Figure J.3

Tracking the annihilation photon

Flowchart of creating a response function by the Monte Carlo simulation method(simplified)Now, add the energy applied to the secondary electron before it reaches the cutoff energy or theenergy applied to the secondary electron before it escapes from the detected body, and record the sum
in the corresponding energy section. When an electron pair is generated, the difference between theenergy of the original photon and 1.022 MeV is taken as the absorbed energy. Assuming that two
photons of 0.511 MeV are generated in the directions separated by 180° at the interaction point, track
each photon separately, add the energy absorbed in the detected body, and record the sum in thecorresponding energy section.
Determine the energy absorbed by the detected body for each photon in the above process, andspecify a bin in the energy section for recording, which was initially given as the input data by theabsorbed energy, considering the energy resolution of the detected body. A response function isdetermined by repeating the above process for the number of histories, for each of the energy sections
initially given.
117

Explanation J. 3

Problems in creating a response function

Unlike the case in which a response function is experimentally created by fixing the detector and theradiation source, many contributions of radiation from various directions are present when measuring
environmental radiation; thus, many matters should be considered to create a response function.
In the Monte Carlo simulation, it is a greatly burdensome work to create a program that includes all of
the following: (1) absorption and scattering by the reflection from the aluminum case of the detector
and the MgO contained therein, (2) scattering by the scintillator window and the light guide, (3)scattering by the photoelectric surface of the photomultiplier tube, (4) escape of K X-rays from iodine,(5) directional dependence, (6) pulse-height resolution, (7) efficiency of the detector, (8)peak-to-Compton ratio, and so on. In addition to these factors, the nonlinearity between the amount oflight emission and the secondary electron energy should also be considered. This is summarized asfollows:
The contribution mentioned in (1) is due to the attenuation of the incident primary photon flux and thebuildup by the scattered ray. That is, in the case of a monoenergetic photon, the primary photon flux
decays according to exp − ′( ), but some fraction of it becomes incident on the detected bodyas a scattered ray. Here, ( ) is the linear attenuation coefficient of the absorber (the case, etc.), and̅ is the average traveling distance of photons in the absorber. Therefore, to evaluate an accurateenergy spectrum, not only the attenuation of the primary photon flux but also the contribution of such
scattered rays must be included in the response function. (2) denotes the effect of a glass of thickness
several millimeters or similar, which is present at the exit of the emitted light in the detected body.
When a primary photon is incident from the front of the detected body, the fraction of photons
transmitted through the detected body, which is backscattered at the front and absorbed again by thedetected body, should be considered. When a monoenergetic photon is measured, the peak observed
around the range of 180 keV is contributed by such scattered rays. The contribution mentioned in (3)is due to the component in which the photon transmitted through the body as described in (2) isfurther backscattered on the photoelectric surface of the photomultiplier tube and reabsorbed by thedetected body. The contribution of this component is considered to be smaller than that of (2). The
contribution mentioned in (4) is related to the structure of the detected body, and it is important for
low-energy photons when an interaction occurs on the surface of the detected body. In the case of (6)and (8), not only the detected body, but the characteristics of the whole system including themeasurement system, are involved.
For accurately evaluating the contributions of (1) through (5), there is no other way than either
incorporating it into a Monte Carlo simulation program or correcting it with an experimental
analytical method. In the former case, because of the structure of the detected body, a simplified
boundary condition is inevitably established owing to partially overlapped aluminum plates andnonuniformity of the reflective material. In the latter case, a single low-energy source is required
because the lower the energy of the photons, the more likely that these contributions are not negligible.
Because the case is generally thin, it is sufficient to consider the contribution of the scattered rays by
the case only once. Directional dependence is related to the structure of the case and is particularly118remarkable in the low-energy region. In summary, a form Rm close to the actual system can beobtained by multiplying the response matrix R created for the bare detected body by the matrix Ms
representing the contribution of the absorptive scattering and the matrix D for correcting thedirectional dependence to modify the response matrix.
ｍ = D × Ms × R(J.12)

However, the errors in the result are only a few percent, even if corrections are not made for (1)through (5).
Corrections for (6) through (8) are important for adapting the response function (matrix) created in thebare detected body to the actual system, or for adapting the response function already published to theemployed detection system. To correct the resolution, it is sufficient to adapt only a portion of thetotal absorption peak to the resolution of the system to be used, considering the peak-to-Compton
ratio, because the distribution due to the Compton effect is continuous. Because the response function(matrix) thus corrected, whose area (element) is statistically calculated for each incident photon
energy, should produce the total efficiency, the details can be remodified in comparison with the total
efficiency ε (E) = 1 − exp − ( )ℓ , which is theoretically obtained using the attenuation
coefficient of the detected body.

Regarding a bare 3 in. φ NaI (Tl) detector, an example of an unevenly divided response function(matrix) composed of 22 rows and 22 columns, corrected by calculating the effect of the MgOreflector and the Al case by the plate approximation for the response function (matrix) obtained by aMonte Carlo simulation, is illustrated in Table J.3. The number of histories for each energy is 50000.
Additionally, if the numerical value of each element is divided by 1000, a response to amonoenergetic single photon incidence can be obtained.

119

Table J.3 Example of a 22 × 22 response matrix of a 3 in. φ NaI (Tl) scintillator for a uniformly incident gamma-ray field
E

0.1

0.2

0.4

0.5

0.1

979.7

0.0

0.2

27.0 916.6

0.3

71.7

17.3 759.6

1.6

0.0

0.4

80.8

66.4

15.6 593.6

10.1

0.0

0.5

92.1

73.7

58.1

17.8 473.4

13.8

0.0

0.6

83.3

73.0

66.7

51.5

20.6 381.8

17.8

0.0

0.7

73.5

66.4

60.7

59.2

45.8

22.6 318.8

17.8

0.0

0.8

63.4

60.5

55.7

55.0

56.6

40.6

23.0 270.0

21.5

0.0

0.9

54.6

52.4

49.1

48.2

47.8

52.6

34.6

26.5 234.8

23.2

0.0

1.0

46.0

44.7

44.3

42.5

41.0

44.3

49.3

32.7

27.0 210.3

25.0

1.1

42.5

40.3

37.3

38.6

37.7

38.6

39.9

48.9

31.1

27.4 188.1

1.2

36.8

34.0

33.8

34.0

34.2

34.4

34.9

38.2

48.0

1.32

31.8

30.4

29.2

28.3

29.4

29.2

28.5

30.2

1.465

26.8

27.8

25.0

23.9

24.8

24.3

24.3

1.615

22.4

21.3

21.9

21.3

21.2

24.6

1.765

18.4

18.4

18.4

18.4

18.4

1.970

15.1

15.1

15.1

15.1

2.205

11.8

11.8

11.8

2.410

9.9

9.9

2.615

8.3

2.860
3.100

E

0.3

0.6

0.7

0.8

0.9

1.0

1.1

1.2

1.32

1.465 1.615 1.765 1.970 2.205 2.410 2.615 2.860 3.100

24.6

0.0

28.7

28.5 168.2

26.1

0.0

35.3

43.0

35.1

17.3 180.9

11.8

0.0

25.2

26.5

30.3

38.4

41.9

24.3 169.7

10.5

0.0

20.6

22.6

22.1

23.5

29.4

33.5

58.1

27.0 152.4

11.8

0.0

18.4

21.3

21.7

18.6

19.7

20.0

25.7

44.7

57.7

29.8 143.6

12.7

0.0

15.1

15.1

15.1

15.1

18.2

19.2

16.4

17.5

26.3

43.9

53.7

34.0 157.2

1.5

0.0

11.8

11.8

11.8

11.8

11.8

11.8

11.8

14.9

21.7

22.1

24.6

33.5

47.8

72.4 140.0

5.3

0.0

9.9

9.9

9.9

9.9

9.9

9.9

9.9

9.9

9.9

9.9

21.7

27.0

21.1

26.8

84.6

50.0 126.5

8.6

0.0

8.3

8.3

8.3

8.3

8.3

8.3

8.3

8.3

8.3

8.3

8.1

11.8

15.6

31.6

21.5

50.4

70.3

46.7 122.4

7.0

0.0

7.0

7.0

7.0

7.0

7.0

7.0

7.0

7.0

7.0

7.0

7.0

7.2

9.2

11.8

16.0

23.0

45.8

37.7

68.0

53.3 131.6

2.6

5.9

5.9

5.9

5.9

5.9

5.9

5.9

5.9

5.9

5.9

5.9

5.9

7.2

8.8

9.2

9.4

35.3

20.9

32.5

70.4

0.7

120

73.0 105.0

Appendix

121

122

Appendix 1

Calculation Examples of a Response Function

The components of a response function consist of a Compton distribution component generated from theprimary incident photon, a KX-ray escape component, an annihilation gamma-ray spectrum component,
and the total absorption component. However, the detector actually relays a shape specific to each probe,
because of the influence of the NaI (Tl) crystal, as well as scattered gamma-rays generated from the casethat encloses it, and from the photomultiplier tube. Although it is ideal to obtain a response function in astrict sense by means of a standard gamma-ray-based experiment in which absorption and scattering
components are not included, it is impossible to obtain such a source in practice. Even if this was possible,
high accuracy cannot be expected under normal experimental conditions. However, in considering adetermination method that does not include such an error factor, it is possible to employ a theoretical
calculation method based on the actual shape of the detector, anticipating that the determination of a
response function with high accuracy would be feasible. In this case, given that it is assumed that the
spectral analysis described here should be performed by means of a simple calculation code without
impairing accuracy, the shape of the detector is modeled so that any response functions can be calculated
by a semi-experimental method.
Based on the assumption that the spectral distribution considered here is supposed to consist of a Compton
component, a KX-ray escape component, an annihilation gamma-ray spectral component, and a total
absorption component, it is determined by calculating each component first and then synthesizing a result.
Normalization is performed using the peak efficiency, which is obtained based on an experiment on the
total absorption component and other components, as well as the overall efficiency.
The approximation is outlined as follows:[1] Compton spectral distribution
Synthesis by means of a rectangular distribution and an exponential function.
[2] Spectral distribution produced by the KX-ray escape of iodine
Spectrum synthesis involving an escape using the escape rate of the KX-ray of iodine.
[3] Spectral distribution by means of annihilation gamma-rays
This is synthesized by calculating a probability based on the response function of 0.51 MeV.
[4] Total absorption component
This is synthesized as a δ function, including the total absorption component produced byannihilation gamma-rays.
A complete response function is obtained by superimposing [1] to [4]. Figures 1.1 and 1.2, and Table 1.1
show examples of the response functions that are actually calculated from a 3 in. φ spherical NaI (Tl)detector.

123

(Count/50 keV)/(γ/cm2)(Count/50 keV)/(γ/cm2)

Channel number

Channel number

Figure 1.1

Calculation examples of a response function

124

(Count/50 keV)/(γ/cm2)(Count/50 keV)/(γ/cm2)

Channel number

Channel number

Figure 1.2

Calculation examples of a response function

125

126

(for a 3 in. φ spherical scintillator, 50 keV/bin)

Table 1.1 Examples of Response Functions Generated by Synthesis

127

128

129

130

131

132

133

134

Example of Data Processing of NaI Monitor by Stripping Method*1

Appendix 2This data processing is a calculation procedure to analyze the pulse-height distribution obtained using a 3 in.
φ spherical NaI (Tl) scintillator, to obtain information on gamma-rays and cosmic rays in the range of 0–3
MeV.
It can be generally classified into (1) the process of converting measurement results to 10 keV/ch, (2)selection and introduction of input data necessary for processing, (3) calculation performed based on the
input data, and (4) display of calculation results, and others.
(1) Process of converting measurement results to 10 keV/chThis conversion process standardizes the energy calibration, i.e., the relationship between the channel andthe energy of the pulse-height distribution obtained (i.e., creation of a uniform channel width among allenergy regions). The calibration is performed by interpolating the numerical values between channels.
Usually, three points of a zero point, 40K (1461 keV), and 208Tl (2614 keV) are used to make correctionsin the first order. However, to correct the distortion in the higher orders, fitting is performed in the nthorder using more than (the order + 2) the number of sets of the relationship between the channel and theenergy of the measured pulse-height distribution. The interpolation of the numerical values between
channels is determined by means of proportional distribution. Given that the result after conversion to 10
keV/ch is used as the basis for data processing performed in and after (2), the correctness of thiscalibration affects the accuracy of the result.
(2)

Input data required for processing

Considering the input data required for processing, the P/T ratio of the detector (NaI (Tl) scintillator) is
the only measured data necessary for this analysis. This value is obtained using a point source ofmonochromatic gamma-rays, as shown in Figure 2.1 and Table 2.1.
In addition, the data of the overall counting efficiency required for processing is obtained by calculating
the literature values for specific constants such as the active cross-section of gamma-rays and thescintillator.
(3) Calculation processing based on input data
The processing of the measurement results obtained from conversion to 10 keV/ch includes (a) subtractionof the Compton contribution and (b) application of peak efficiency. Peak efficiency (PE) can be
determined from the following equation:
PE = 1 −

(

2

)

+2

(

1

)

+

1

(

)

using the P/T ratio (PTTR), the diameter “d” of the spherical scintillator, as well as the attenuation
coefficient μ of the NaI (TI) scintillator. The peak efficiency values calculated for the 3 in. φ spherical NaI(Tl) scintillator are shown in Table 2.2. The peak area count Np and the count Nc of the Compton
continuum are as shown in the following equation:
=
*1

(1 −

)/(

)

As this is quoted from the Radioactivity Measurement Series No. 20 "Environmental Gamma-ray
Spectrometry" (1990), the former units of “R” and “rad” are used.
135

Nc is obtained from Np using the preceding equation and Nc is subtracted by distributing it evenly to eachchannel of the Compton continuum.
This corresponds to a rectangular approximation of the Compton continuum (Figure 2.2). In practice, thechannel corresponding to 3 MeV (10 keV width) is first regarded as the total absorption peak (the area Np
count) of the photon of energy Eγ, and Nc obtained from this Np is distributed evenly to the corresponding
channels from the Compton edge Ec (MeV) (= Eγ/(1 + 0.511/2 Eγ)) to 0. Subsequently, this operation isrepeated sequentially from the high-energy side to the low-energy side. In this way, the spectrum
excluding the Compton contribution is obtained and the photon spectrum expressed as the gamma-ray flux
density is determined by applying the peak efficiency of each energy and the attenuation rate of the
gamma-rays caused by the scintillator case for each channel. Furthermore, the exposure dose rate for eachenergy is also obtained by applying the conversion factor of the photon flux density-exposure dose to eachenergy.
(4) Reliability of the results
The errors produced by this method include statistical errors such as the counting errors, including
calculation processes, and systematic errors, such as counting losses, errors in conversion to 10 keV/ch,
P/T ratio, the subtraction method of a Compton continuum (differences between a rectangle and the actual
shape), selection of basic constants, and differences in background data.
(5) Analysis and representation of results
Normally, the range of 0–3 MeV is treated as the information on gamma-rays, and the range of 3 MeV ormore is mainly treated as the information on cosmic rays.
i)

Analysis and representation of the region of 0–3 MeV[1] Representation of the results of converting the measured data to 10 keV/ch.
[2] Representation of the spectrum in which a Compton contribution is subtracted.
[3] Representation of the results of calculating the flux density using the photon peak efficiency forthe aforementioned spectrum, i.e., the representation of the photon spectrum.
[4] Representation of the results for converting the photon spectrum to the exposure dose spectrum.
[5] For [1], [3], and [4], the results of integrating the values of 3 Mev–10 keV into the low energy
region individually.

The standard representations are as follows. Representations of [1], [2], [3], and [4] are generally madeusing a semilogarithmic plot. For the integration result of [4], a linear expression in percentage is used
for integral values up to the specified energy (generally, up to 30 keV) set to 1. It is also possible torepresent gamma ray and cosmic ray doses together with data numbers as shown in Figure 2.3.
ii)

Treatment of the region above 3 MeV

The information on the region above 3 MeV is mainly based on cosmic rays and in a 3 in. φ spherical
NaI (Tl) scintillator, the cosmic ray dose rate equivalent can be obtained by multiplying the counting
rate above 3 MeV by the constant (N3) of 2.12 (μR/h)/(cps).
However, in an environment in which gamma-rays with energy above 3 MeV are present (in thevicinity of accelerator facilities, reactor facilities, etc.), high-energy gamma-rays may be overestimatedas cosmic ray contributions. For measurements recorded at such a site, it is necessary to use an energy
region with little influence due to artificial radiation, for example, a region of 10 MeV or more, tocalculate the cosmic ray dose. In this process, N10 = 3.5 (μR/h)/(cps) should be used as the136

aforementioned constant. Although the ratio (N10/N3) of both of the constants varies as 1.3–1.87
depending on the cosmic ray components in the environment, 1.65 is used here.

3 in. φ x 3 in. cylindrical NaI (Tl)
Source distance 10 cm+ 3 in. spherical Nai (Tl)P/T ratio

Source distance 30 cm

Photon energy (MeV)
Figure 2.1

P/T ratio of NaI (Tl) scintillator

137

Table 2.1 P/T ratio of 3 in. φ spherical NaI (Tl) scintillator

keV

0

10

20

30

40

50

60

70

80

90

0 1.0

1.0

1.0

1.0

1.0

1.0

1.0

1.0

1.0

1.0

100 0.99

0.985

0.98

0.97

0.962

0.955

0.945

0.94

0.933

0.925

200 0.916

0.907

0.9

0.89

0.882

0.872

0.862

0.853

0.843

0.83

300 0.82

0.806

0.795

0.785

0.77

0.76

0.746

0.735

0.725

0.713

400 0.703

0.691

0.682

0.67

0.66

0.65

0.641

0.631

0.621

0.612

500 0.604

0.596

0.586

0.578

0.57

0.562

0.555

0.547

0.541

0.535

600 0.527

0.522

0.516

0.51

0.505

0.5

0.495

0.49

0.485

0.48

700 0.475

0.47

0.466

0.461

0.457

0.453

0.449

0.445

0.441

0.438

800 0.434

0.431

0.428

0.425

0.421

0.417

0.414

0.411

0.408

0.405

900 0.401

0.398

0.396

0.393

0.391

0.387

0.385

0.382

0.379

0.376

1000 0.374

0.371

0.369

0.366

0.364

0.361

0.359

0.357

0.355

0.353

1100 0.35

0.348

0.346

0.344

0.342

0.34

0.338

0.336

0.332

0.33

1200 0.33

0.328

0.326

0.324

0.322

0.32

0.318

0.316

0.314

0.312

1300 0.31

0.308

0.307

0.305

0.304

0.302

0.3

0.298

0.296

0.295

1400 0.294

0.292

0.291

0.29

0.289

0.287

0.285

0.284

0.283

0.281

1500 0.28

0.278

0.277

0.276

0.275

0.273

0.272

0.271

0.27

0.269

1600 0.267

0.266

0.265

0.264

0.263

0.261

0.26

0.259

0.258

0.257

1700 0.255

0.254

0.253

0.252

0.251

0.25

0.249

0.248

0.247

0.246

1800 0.245

0.2446

0.243

0.242

0.241

0.24

0.239

0.238

0.237

0.236

1900 0.235

0.234

0.233

0.232

0.231

0.23

0.229

0.228

0.227

0.227

2000 0.226

0.225

0.224

0.224

0.223

0.222

0.221

0.22

0.22

0.219

2100 0.219

0.218

0.217

0.216

0.215

0.214

0.213

0.212

0.211

0.211

2200 0.21

0.209

0.209

0.208

0.208

0.207

0.207

0.206

0.205

0.204

2300 0.204

0.203

0.203

0.202

0.202

0.201

0.2

0.199

0.198

0.197

2400 0.197

0.196

0.196

0.195

0.195

0.194

0.193

0.193

0.192

0.191

2500 0.191

0.19

0.19

0.189

0.189

0.187

0.187

0.186

0.186

0.185

2600 0.185

0.184

0.184

0.183

0.182

0.181

0.181

0.18

0.18

0.179

2700 0.179

0.178

0.178

0.177

0.177

0.176

0.176

0.175

0.175

0.174

2800 0.174

0.173

0.173

0.172

0.172

0.171

0.171

0.17

0.17

0.169

2900 0.169

0.168

0.168

0.167

0.167

0.166

0.165

0.165

0.164

0.164

138

Table 2.2 Peak efficiency of 3 in. φ spherical NaI (Tl) scintillator

keV

0

10

20

30

40

50

60

70

80

90

0 1.00

1.00

1.00

1.00

1.00

1.00

1.00

1.00

0.999

0.999

100 0.988

0.982

0.976

0.964

0.954

0.945

0.932

0.924

0.913

0.900

200 0.887

0.873

0.861

0.845

0.831

0.815

0.799

0.784

0.768

0.749

300 0.734

0.716

0.700

0.685

0.666

0.652

0.634

0.620

0.607

0.592

400 0.579

0.565

0.554

0.540

0.528

0.516

0.506

0.495

0.484

0.474

500 0.466

0.457

0.447

0.438

0.430

0.421

0.414

0.406

0.400

0.393

600 0.386

0.380

0.374

0.368

0.363

0.358

0.353

0.348

0.343

0.338

700 0.333

0.328

0.324

0.319

0.315

0.311

0.307

0.304

0.300

0.296

800 0.293

0.289

0.286

0.283

0.280

0.276

0.273

0.270

0.267

0.264

900 0.261

0.258

0.256

0.253

0.251

0.248

0.246

0.243

0.241

0.238

1000 0.236

0.234

0.232

0.229

0.227

0.225

0.223

0.221

0.219

0.218

1100 0.215

0.213

0.212

0.210

0.208

0.206

0.204

0.202

0.200

0.198

1200 0.197

0.196

0.194

0.192

0.191

0.189

0.187

0.186

0.184

0.182

1300 0.181

0.179

0.178

0.177

0.176

0.174

0.173

0.171

0.170

0.169

1400 0.168

0.166

0.165

0.164

0.163

0.162

0.160

0.160

0.159

0.157

1500 0.156

0.155

0.154

0.153

0.153

0.151

0.150

0.150

0.149

0.148

1600 0.147

0.146

0.145

0.144

0.144

0.142

0.141

0.141

0.140

0.139

1700 0.138

0.137

0.136

0.136

0.135

0.134

0.134

0.133

0.132

0.131

1800 0.131

0.130

0.129

0.128

0.128

0.127

0.126

0.126

0.125

0.124

1900 0.123

0.123

0.122

0.121

0.121

0.120

0.119

0.118

0.118

0.118

2000 0.117

0.116

0.116

0.116

0.115

0.114

0.114

0.113

0.113

0.113

2100 0.112

0.112

0.111

0.111

0.110

0.109

0.109

0.108

0.108

0.107

2200 0.107

0.106

0.106

0.106

0.105

0.105

0.105

0.104

0.104

0.103

2300 0.103

0.102

0.102

0.102

0.102

0.101

0.100

0.09970.0991

0.09862400 0.0985

0.09790.09780.0972

0.0972

0.0966

0.09600.09590.09530.09482500 0.0947

0.0942

0.0941

0.0935

0.09350.09240.09230.09180.0917

0.09122600 0.0911

0.09060.09050.08990.08940.08880.08870.08820.08810.08752700 0.0875

0.08690.08690.08640.08630.08570.08570.08520.08510.08462800 0.0845

0.08400.08400.08340.08340.08290.08280.08230.08220.08172900 0.0817

0.08110.08110.08050.08050.08000.07940.07940.07880.0788139

np: Peak countFigure 2.2

Eγ:

Peak energy

Ec:

Energy at a Compton end

R:

P/T ratioW:

Channel widthnc :Count/channel

Subtraction method for a Compton continuum140Cosmic ray dose rate = 3.33 gamma-ray dose rate = 4.92 (μR/h)

141

Figure 2.3 Representation example of spectrum analysis results using a 3 in. φ spherical NaI (Tl) scintillation detector

Photon energy (MeV)

Dose rate for each gamma-ray energy(μR/h)

Energy spectrum of gamma-ray
flux density

Spectrum in which a Compton portion issubtracted

Integrated dose of less than 3 MeV(percentage)

Spectrum converted to 10 keV/ch

Dose rate(percent)

142

Table 2.3 Representation example of spectrum analysis results using a 3 in. φ spherical NaI (Tl) scintillation

Description of Table 2.3* DATA NO.

Data number

The measured data number is expressed as six numbers consisting of ABCDEF. A indicates the data for theyear 197A or 198A, and BC is the order of the tape in the year A; thus, ABC indicates the tape number.
The data number of the ABC tape is assigned to DEF.
* "B"

Meaning of this representation

This indicates that the calculation is performed using the data obtained by subtracting the background
specific to the detector, or specific data as the background.
* "C"

Meaning of this representation

The background subtraction, in addition to the aforementioned, is a calculation performed by obtaining thecontribution of cosmic rays, etc. from the count above 3 MeV (ABOVE COUNT) and using the subtracted
result.
* "D"Meaning of this representation

This indicates that the calculation was performed using the raw data without subtracting these backgrounds.
CHANNEL

Number of channels

The number of channels used in the measured data; either 1024 or 2048 is mostly used in this program.
KEV/CH

Channel width

Indicates the channel width for the numbers shown in NO. Usually, 10 keV/ch is used.
TIME

The time of recording

Indicates the time for which the recording is normally performed. The first two digits indicate the hour, andthe following digits indicate the minutes containing the decimal point, respectively. In some cases, therecorded data number is set to ADEF.XY, indicating that the data is collected in the number XY by DEF.
LAST-CH

Last channel

Indicates the last channel used when data analysis was performed. A channel corresponding to 3 MeV isusually indicated.
LIVE-TIME

Measurement time

Indicates the time (s) for which measurement is made using 0 channels of the spectrum. In some cases, thistime is added separately and in other cases, a number is used which is obtained by multiplying the numberof 0 channels several times (to correct counting losses). In these cases, the 0 channels of the spectrum (thetotal counting hours) are shown in parentheses.

143

ABOVE-CNT

Excessed countTotal count values ranging from the last channel for which spectrum analysis was performed (LAST CH) tothe last channel actually used.
COS-DOSE (μR, H)

Cosmic ray doseThis is calculated as cosmic ray dose (ABOVE CNT/LIVE TIME) x K, which is calculated from ABOVE
CNT. The value 2.12 is used for K. If a different value is used, it is shown in the last term.
GAMMA (μR, H)

Gamma-ray dose

The gamma-ray dose is generally calculated from the spectrum up to 3 MeV.
(CH)
This is the lower limit of the energy region employed in the preceding dose calculation. If this value is 3, it
indicates that integration up to 30 keV has been performed. For this parameter, 2 - 5 is usually used. In thisexample, a value of 10 keV is used.
WEIGHT

Weight of the detector

For the weight of the detector, (NaI (Tl)) was used. By predefining this selection, the average absorbed
dose rate of the scintillator is obtained from the sum of the product of the counting rate corresponding toeach energy of the pulse-height distribution and the energy (Σi Ni × Ei).
E.CH

Channel width

Indicates the channel width of the measured data. The measured spectrum is corrected to 10 keV/ch using
an nth order equation and the coefficient of the first order equation obtained in advance is shown.
RESULTS

Calculation results

DATA is the counting rate (cps) of 10 keV - 3 MeV.
FLUX is the total flux density (γ/cm2) of 10 keV - 3 MeV.
DOSE is the exposure dose rate (μR/h), obtained by integrating the values of 10 keV - 3 MeV. ADOS is theaverage absorbed dose rate (μrad/h) in the scintillator.
PER D
This is obtained by dividing each dose of RESULTS by the value of the exposure dose; however, theDOSE value (this should be 1.00) is obtained by dividing the value of the channel region set to (CH) × 10
keV - 3 MeV by the value of 10 keV - 3 MeV. Therefore, the value is not always 1.00.
0 - 50, 50 - 100, 100 - 250... up to 3000
This value is a percentage between the energy regions of 0 - 50 keV, 50 - 100 keV, 100 - 250 keV, 250 500 keV, 500 - 1000 keV, 1000 - 2000 keV, and 2000 - 3000 keV, calculated for each result, respectively.

144

NO.

Channel number

Data is output as the value of each channel number (10 keV/ch) but the content to be output can be selected.
Generally, when data is output on one page, the values of 1 - 100 KeV are output for every 10 keV/ch, andthe values of 100 keV - 3 MeV are output for every 100 keV.
* DATA (cps)The counting rate value (cps/10 keV) for each channel (10 keV) is calculated for a specified channel bycorrecting the measured data to 10 keV/ch using the nth order equation.
* PEAK (cps)This is the result of subtracting the Compton contribution from DATA (cps). In the case of "B" and "C,” acalculation is made by subtracting the background spectrum, which is separately predefined.
* FLUX (γ • cm-2 • s-1)
This is the true photon flux density obtained by dividing the PEAK (cps) result by the peak efficiency ofthe detector.
* DOSE (μR/h)
The exposure dose rate calculated from the above photon flux density. The organ dose rate, the effective
dose rate, etc. can be calculated by multiplying the result by the coefficient determined from the energy
absorption coefficient of a material.
SUM OF DATA, FLUX, DOSE
Individual results are integrated from 3 MeV to the specified channel in the low energy region.

Note 1)Note 2)Note 3)

Note 1)

It is possible to calculate up to five sets of RESULTS for the output, including a comparison withother measured values such as ionization chambers whose characteristics are known as a result ofcorrecting the absorption for a scintillator case, and the values such as the genital gland dose. In thiscase, the values other than DATA will change.
Note 2)The underlined content is used in the graphical display of the results.
That is, the data number; representation of C, D, B; cosmic ray dose rate C =; and the gamma dose
rate G =.
Note 3)
Items marked as * are also graphically displayed. Each graph is shown as semilogarithmic, and SUM
OF DOSE is shown as a linear graph in percentage. Among these numerical values, the underlined
data is made into a successive graph.
145

Appendix 3 Example of unfolding of NaI (Tl) scintillation spectrum by the method of
successive approximation*1
There are various methods for converting the pulse-height distribution obtained by a NaI (Tl) scintillation
spectrometer into a photon energy spectrum (the response correction). In this document, we describe an
example of analyzing the data measured by a 3 in. φ spherical NaI (Tl) scintillator using the method of
successive approximation, based on a response matrix.
Appendix 3.1

Outline of the analysis method(1) Energy calibration
Energy calibration is performed with the aid of a computer to determine the number of channels of thetotal absorption peaks of 208Tl (2.61 MeV) and 40K (1.46 MeV) among the environmental gamma-ray
pulse-height distributions, and by using the obtained two peak channels and their corresponding energies,
and the energy typically associated with the "zero point."(2) Adjustment of the input data
The pulse-height distribution is divided every 20 keV for values below 100 keV, and every 100 keV forvalues above 100 keV to conform to the response matrix division to obtain a corrected pulse-height
distribution.
(3) Subtraction of background components
Cosmic ray components and the detector-specific background are subtracted. Cosmic ray components are
obtained by the coincidence method with a plastic scintillation spectrometer. In addition, thedetector-specific background is obtained from the measurements taken in an iron chamber.
(4) Response correction
The pulse-height distribution is corrected using a response matrix and converted to an incident gamma-ray
spectrum. (The response matrix used at this time is shown in Table 3.1. The details of the method are also
given in Appendix 3.2.)(5) Calculation of dose rates
The energy spectrum is expressed in the roentgen unit, by multiplying the incident gamma-ray spectrumby the energy absorption coefficient of air that corresponds to each energy. In addition, the exposure dose
rate at the measurement location can be obtained by integration across the entire energy range (details of
this method are given in Appendix 3.3).

Appendix 3.2

Energy analysis by the method of successive approximation

The probability of energy E being absorbed when one photon of energy E0 is incident on the detector is R
(E0, E). When the energy distribution of the incident photon is X (E0), the distribution Y (E) of theabsorbed energy E can be expressed as follows:
Y(E) = ∫
*1

(

,

)•

(

)

(Attachment 3-1)

As this is quoted from the Radioactivity Measurement Series No. 20 "Environmental Gamma-ray
Spectrometry" (1990), the former unit of “R” is used.
146

Here, when the incident energy and the absorbed energy are equally divided into n equal parts, Ej (j = 1,
2,..., n) is incident from the equation (Appendix 3-1), and the distribution Y (Ei) of the absorbed energy Ei(i = 1, 2,..., n) can be expressed:
Y(E ) =•• ∆(Attachment 3-2)

Further, for a wide range of incident energies Ej (j = 1, 2,..., n), we have:Y(E ) = ∑•• ∆(Attachment 3-3)

The preceding equations can be used. In the equation (Attachment 3-3), if n = ∞, the equation of (Appendix
3-1) is applied.
X (Ej) (j = 1, 2,..., n) represents the energy spectrum of the radiation incident on the detector, and Y (Ei) (i =1, 2,..., n) is the pulse-height distribution observed on the detector. Rij represents the response function of
the detector relative to the incident radiation of energy Ej.
The equation (Attachment 3-3) can be expressed as follows, using the square matrix R composed of n rows
and n columns, and the n-dimensional column vectors X and Y:
Y =R• X

(Attachment 3-4)=

(Attachment 3-5)

= R•

(Attachment 3-6)

Next, a pulse-height distribution Y0 is taken as a first approximation of the spectrum to be obtained.

Y1 is obtained by multiplying this term by the response matrix R.

Next, for X2 of the second approximation, the ratio of the corresponding elements of X1 and Y1 ismultiplied by the corresponding element of Y0. Thus, the first element of X2 is given by:
(

(

) = (

)

)

• ( )

(Attachment 3-7)

is obtained from the preceding equation. Y2 is then obtained by further multiplying X2 by R.
=R•

Similarly:
(

(

) = (

(Attachment 3-8)

)

)

• ( )

(Attachment 3-9)

147

If such procedures are repeated 50 times, a value converged within the range of up to 10-1 is obtained.
Appendix 3.3

Dose conversion

Response correction is performed by the successive approximation method, and the dose conversion fromthe incident gamma-ray spectrum is performed using the following equation:
D = 1.724 ×

(

•

•

•

•

)•

(

⁄ℎ )

(Attachment 3-10)

where,
Ei: the center energy of the ith histogram (eV)W

: the W value of air (33.85 eV/ion pair)r: the radius of the scintillator (3.81 cm)

μi: the linear attenuation coefficient of air (cm-1) relative to the photon of energy Ei

Ni: the number of incident photons/⊿ E (items/(100 keV • s)) entering the ith histogramT: measurement time (s)

148

149

the central energy of a photon (MeV)

the central energy of a photon (MeV)

Table 3.1 Response matrix of a 3 in. φ spherical NaI (Tl) scintillation detector

Appendix 4

References

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2） JIS Z8103：2000 計測用語
3） Radiation protection instrumentation‐Transportable, mobile or installed equipment to measure
photon radiation for environmental monitoring

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5） 柴道勝, 下村遼平, 武石稔, 三枝純, 福島環境におけるモニタリングカー車内外空間線量率換算手法の検討, 日本放射線安全管理学会, 第 14 回学術大会（2015）
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151

Revision history
Established in July 1977
Revised in March 1996
Reviced in December 2017152