The Series of Environmental Radioactivity Measuring Methods
DR-SITU(No.33)In-Situ Measurement Using Germanium
Detector

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

Sei-1

Table of Contents
Chapter 1 Introduction............................................................................................................................. 1
Chapter 2 Definitions of terminology...................................................................................................... 2
Chapter 3 Measurement equipment ......................................................................................................... 4
3.1

Requirements for the equipment........................................................................................... 4

3.2

Composition of the equipment ............................................................................................. 4

3.3

Example of the equipment specifications ............................................................................. 4

3.4

Equipment calibration........................................................................................................... 7

3.5

Measures against equipment contamination ....................................................................... 12

3.6

Other preparations .............................................................................................................. 14

Chapter 4 Measurement methods and spectral analysis ........................................................................ 15
4.1

Selection of the measurement site ...................................................................................... 15

4.2

Measurement ...................................................................................................................... 15

4.3

Recording ........................................................................................................................... 17

4.4

Spectral analysis ................................................................................................................. 19

Chapter 5 Calculation of radioactivity concentration and ambient dose rate ........................................ 21
5.1

Conditions for analysis ....................................................................................................... 21

5.2

Calculation of radioactivity concentration.......................................................................... 22

5.3

Calculation of ambient dose rate ........................................................................................ 32

Chapter 6 Interpretation of measured results ......................................................................................... 36
6.1

Implications of discrepancies between analytical and actual measurement conditions ..... 36

6.2Management of accuracy in measured result...................................................................... 38

Explanation
Explanation A Calculation of peak efficiency by simulated calculation ···························· 41
Explanation B Measurable ranges and measuring times for in-situ measurement ················· 45
Explanation C Example of spectrum by in-situ measurement ········································ 52
Explanation D Energy calibration in a nuclear disaster ················································ 56
Explanation E Vertical profile of radioactive materials in the ground································ 58
Explanation F Implications of discrepancies between analytical and actual measurement
conditions ·················································································· 76
Explanation G Directional characteristics of detector (angular dependence of peak efficiency) ·· 87
Explanation H Example of actual measurement························································· 89
Explanation I

Intercomparison measurement ·························································· 97

Appendix
Appendix 1
Appendix 2
Appendix 3
Appendix 4

Relationship between radionuclide activity and gamma-ray fluence rate 1 m above
ground ······················································································ 103
Relationship between dose rate and gamma-ray fluence rate 1 m above ground·· 115
Relationship between radionuclide activity and dose rate 1 m above ground ····· 137
References ················································································· 149

Chapter 1 Introduction
In situations where radioactive materials are released into the environment, such as accidents
occurring at nuclear facilities (hereafter, “nuclear accidents”), conducting an in-situ measurement*1
using a germanium (hereafter, “Ge”) semiconductor detector, which has excellent energy resolution,
enables the identification of the radioactive materials deposited on the ground. Further theradioactivity concentration and ambient dose rate attributable to the deposition can be obtained.
As an in-situ measurement entails the measurement of the ground surface in the field, it requires
approximately one-tenth of the time needed to conduct a measurement of a soil sample in a laboratory.
While it is difficult to obtain representative readings from a soil sample if the deposition of radioactive
materials is uneven over the ground owing to weather conditions, an in-situ measurement can facilitate
obtaining the result that indicates the average readings of the area measured. Therefore, it isparticularly useful when developing a distribution map of radioactive deposition over a wide area
following a release of radioactive materials into the environment. It was also employed in the effort tograsp the distribution of radioactive deposition following the nuclear accident that occurred at the
Fukushima Dai-ichi Nuclear Power Plant of Tokyo Electric Power Company Holdings (hereafter,“Fukushima Dai-ichi Nuclear Accident”), which was triggered by the major earthquake that hit easternJapan in March 2011.*2 The ability to identify the radioactive materials enables the estimation of thestate of emissions and furthermore, based on the half-lives of the identified radioactive materials, theprediction of the future ambient dose rate fluctuations to contribute toward accurate dose evaluation.
This document describes the calibration of a Ge detector for in-situ measurement and methods ofmeasurement and analysis based on the gamma rays of radioactive materials deposited on the ground.
It also includes points of caution in terms of in-situ measurement following a nuclear accident, which
differ from measurement during normal times, drawing on the post-Fukushima-Dai
ichi-Nuclear-Disaster measurement examples.
The analytical method is based on the HASL*3 method (H. L. Beck, et al.; HASL-258 (1972)) and has
drawn on the research published in ICRU*4 Rep. 53 (1994), and other studies. The HASL method isbased on certain assumptions in the analysis in terms of the vertical in-soil profile of radioactive
materials, topography of the surrounding areas, installation height of the detector, etc.; thus, it isnecessary to understand the impact on the analysis results if the actual conditions differ from theseassumptions. In this manual, this impact is clarified and further described in terms of the correction
method.
The detectability level is approximately 0.03 kBq/m2 of ground-deposited radionuclides through a 1-hmeasurement, and approximately 0.1 nGy/h of ambient radiation dose rate from those nuclides (seeExplanation B).
Because the in-situ measurement can serve to obtain the radioactivity concentrations in the ground andambient dose rates of naturally occurring radionuclides such as uranium series, thorium series, andpotassium-40, the methods to perform these are also explained.
――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1 “in-situ” means the field where things are taking place. In this document, “in-situ measurement” refers toan “in-situ measurement conducted using a Ge detector.”*2 “FY 2014 Commissioned Investigation of Radioactive Materials (collection of data on the distribution ofradioactive materials in relation to the accident at Tokyo Electric Power Company’s Fukushima Dai-ichi
Nuclear Power Plant and the development of transition models) Outcome Report, Area Research ofRadioactive Cesium Deposition Amounts” by Satoshi Mikami and Kimiaki Saito (2015)*3 Acronym for the Health and Safety Laboratory. Subsequently, it underwent changes to theEnvironmental Measurements Laboratory (EML) and is presently known as the National Urban Security
Technology Laboratory (NUSTL).
*4 Acronym for the International Commission on Radiation Units and Measurements.

1

Chapter 2 Definitions of terminology
Terms, in particular, those specific to in-situ measurement, are defined below. For the basic terms usedin relation to the Ge detector, refer to the Radioactivity Measurement Series No. 7 “Gamma-ray
Spectrometry Using Germanium Detectors.”
In-situ measurement
A measurement taken on-site, out in the field. In Latin, “in-situ” literally means “in place.”
Response
The ratio of an indicated value to the amount to be measured. It is the inverse of a calibration
constant.
Angular dependence
The property that the response to gamma rays that enter the Ge crystal of the detector depends on
the angle of incidence.
Deposition amountThe amount of radioactive material that is deposited on, or penetrates, the ground of unit area
(Bq/cm2).
Mass depth
The depth measured from the ground surface by per-unit-area mass of the soil (g/cm2).
βWeight buffer depth (g/cm2). A parameter to indicate the vertical profile of radioactive materials in
the ground, which equals the mass depth at which the radioactivity concentration is 37% of theground surface (= 1/e). This expresses the degree of penetration, where greater values indicate
deeper penetration.
Scraper plate
A device to sample segments of soil vertically at any intervals, consisting of a metal frame to installon the ground and a metal plate for scraping the soil within the frame. By attaching a metal bar to
the metal plate, the depth is adjusted for sampling.
Infinite plane
A plain flat open ground that spreads wide with no obstacles.
Vertical profile
The distribution of radioactive material in the ground in the direction of depth.
Ground surface distribution
The state of radioactive materials distributed over an area of ground.
Exponential distribution
The state in which radioactive materials in the ground are distributed in such a manner that theydiminish exponentially in the direction of depth.
Homogeneous distribution
The state of radioactive materials distributed equally in the ground.
HASL
Acronym for the Health and Safety Laboratory. Subsequently, it underwent changes to theEnvironmental Measurements Laboratory (EML) and is presently known as the National Urban
Security Technology Laboratory (NUSTL).

2

ICRU
Acronym of the International Commission on Radiation Units and Measurements.

3

Chapter 3 Measurement equipment
3.1 Requirements for the equipment
The Ge detector used in the in-situ measurement is essentially the same as the ordinary gamma-ray
spectrometer used to measure environmental samples in a laboratory. For the details about thedetector and other measurement equipment, refer to the Radioactivity Measurement Series No. 7“Gamma-ray Spectrometry Using Germanium Detectors” and the IEC*1 Standards.
However, note that the following conditions must be met for conducting the measurements
outdoors, in addition to the requirements for ordinary spectrometers.
(1) The equipment must be easily portable and installable.
(2) It must be able to operate reliably in outdoor weather conditions.*2(3) It must be able to operate on batteries.
(4) The dewar bottle must be functional, light, robust, and of the smallest possible size so as ton notprovide a shielding effect to the detector.
3.2 Composition of the equipment
The equipment compositions are shown in Figures 3.1 and 3.2, illustrating the liquid nitrogencooling system and electric cooling system, respectively.
(1) Ge detector(2) Portable dewar bottle (for the liquid nitrogen cooling system only)(3) Multichannel analyzer (MCA)(4) Detector support rack(5) Personal computer (PC) (for the use of MCA control and data analysis)(6) Software(7) Other auxiliary devices
The electric cooling type is available in an all-in-one portable system including the detection unit,
cooling system, MCA, and control unit.
3.3 Example of the equipment specifications
An example of the specifications for in-situ measurement equipment is presented below. Note thatthe example is indicative of typical specifications and not intended to limit the use of otherequipment.
(1) Ge detector• Coaxial high-purity Ge detector*3――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1

Acronym for the International Electrotechnical Commission.
In particular, measures to protect the equipment from direct sunlight on extremely hot days are essential.
*3
Ideally, the diameter and length of the germanium crystal are equal (which minimizes the gamma-ray
incidence angular dependence of peak efficiency). The n-type detector is used for measuring low-energy
gamma/X rays, but those with an injection window made of beryllium, etc. are vulnerable and notsuitable for outdoor use. It is better to use a detector with an aluminum end cap.
*2

4

Relative efficiency at 25 cm
approximately 25%*4
Energy resolution half-width (FWHM) 2.3 keV or lower*5 against cobalt-60 1333 keV
Peak-to-Compton ratio
30–60:1*6
Measurable energy range 50–2000 keV*7 *8Battery operation time (with the electric cooling system only)
up to 8 h ofconsecutive operation
9• Crystal cooling timeat least 4 h*
• The detector can be installed in the downward-facing position.*10(2) Portable dewar bottle (for the liquid nitrogen cooling system only)• The bottle capacity
up to 7 L*11(3) Multichannel analyzer (MCA)*12• High-voltage power source with built-in amplifier• Amplifier gain approximately 2 to 2000*13• HV±10 to 5000 V• Detector over voltage protection
Must be installed with a feature to automatically shut down the high-voltage power before
the detector overheats.
• Spectral memory 4–8 kch*14• Integral nonlinearity 0.025% or lower• Differential nonlinearity 1% or lower• It can be run on batteries (portable type).
• If it has no display panel on the main part and is used with a PC to display the spectrum andcontrol the measurement, ensure that it can operate even when the PC is turned off.
(4) Detector support rack
• It must be able to support with stability the detector and dewar bottle facing downward 1 m
above the ground surface.
•
•
•
•
•

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*4 Electric cooling system detectors with the relative efficiency of 10–20% are commercially available. For

measurements in places with high ambient dose rates, detectors with low relative efficiencies may proveadvantageous as they are more unlikely to overload the signal processing.
*5 Unlike a liquid nitrogen cooling system, a Ge detector with an electric cooling system tends to increase
the half-widths owing to the device’s mechanical noise.
*6 As the peak-to-Compton ratio is calculated based on the spectrum measured inside the shields, thisserves only as a guide for in-situ measurements.
*7 Set this to roughly 30–2000 keV when measuring low-energy X/gamma rays.
*8 When measuring gamma rays of more than 2000 keV of natural radionuclides (thallium-208 of 2615
keV) etc., set the upper limit of the measurable energy range to 4000 keV.
*9 This depends on the size of the crystal and the performance of the freezing device. Thus, the coolingtime above is a reference guide.
*10 This is necessary to ensure that the liquid nitrogen does not leak, irrespective of the orientation in which
the detector is installed.
*11 Considering the ease of handling as well as the possibility that the dewar shields the rays, it is normally
3–5 L.
*12 Some detectors come with a built-in electric cooling system.
*13 The amplifier gain should be adjustable between 2000 and 4000 keV within the range of the spectral
memory.
*14 The energy per channel is normally at 0.5 keV/ch, and the channels to use are determined according to
the energy range of the target to be measured (2000 keV: 4 kch, 4000 keV: 8 kch).
When measuring gamma rays of energy 2000 keV or more of natural radionuclides (thallium-208 of
2615 keV) etc., use 8 kch.

5• Structurally, it should not pose obstacles between the detector and ground surface.
(5) Personal computer (PC) (for the use of MCA control and data analysis)*15• It must be compatible with MCA to perform measurements.
• It must be able to operate on batteries.
• It must have a display screen that is viewable in the daylight.*16(6) Software• MCA control software• Analysis software
It must be compatible with HASL-258 and ICRU Rep. 53 (more details are given in Chapter
5.)(7) Other auxiliary devices• Carrying cases for the detector, MCA, and PC• Cables (MCA-detector, MCA-PC connection cables, power cable, etc.)• DC/AC converter (from 24/12 V DC to 100 V AC) (optional)• Spare batteries for the detector, MCA, and PC• A dewar bottle to refill liquid nitrogen (20–30 L capacity) (for the liquid nitrogen coolingsystem only)• Other utensils required for the refilling of liquid nitrogen (for the liquid nitrogen coolingsystem only)• Gamma-ray radiation source for calibration• Jigs to set up the gamma-ray radiation source• Items to prevent contamination (plastic bags, etc.)――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*15 The PC is preferably water-proofed as it is used outdoors.
*16 As a PC’s LCD panel is often difficult to view in the bright daylight outside, it is advisable that a simple

light shade or similar be provided to improve the visibility.

6

Dewar containing liquid nitrogen

Preamplifier

High-voltage
power source

Main amplifier

PC for control
Ge detector

Figure 3.1 Configuration example of devices for in-situ measurement (liquid nitrogen coolingsystem)Main component*
High-voltage
power source
Main amplifier

Preamplifier

Freezer

Ge detector
PC for control*Personal Digital AssistantFigure 3.2 Configuration example of devices for in-situ measurement (electric cooling system)(All-in-one portable system including the detection unit, cooling system, MCA, and control unit)
3.4 Equipment calibration
The in-situ measurement involves a Ge detector installed 1 m above the ground, facing downward.
In this position, the gamma rays emitted by the radioactive materials deposited on the groundsurface reach the detector directly at angles between 0° and 90°, given that the detector axis ofsymmetry is 0°. For this reason, it is necessary to consider the detector’s gamma-ray angulardependence of peak efficiency. See Chapter 5 for the details about how to calibrate for angulardependence. In this Chapter, the angular dependence calibration is described in terms of thedetector’s peak efficiency, which is necessary for the calibration, and the procedures of thecalibration.

7

3.4.1 Energy calibration
Measure any given gamma-ray radiation source (ideally, use cobalt-57, cobalt-60, etc.) and adjust
the amplifier gain and analog-to-digital converter (ADC) zero level such that the total absorption
peak is aligned with the target channel.
Use the software to prepare a relational function of the gamma-ray energy and channel and save thecalibration formula thus obtained in a file.*17
3.4.2 Peak efficiency calibration
It is desirable that the peak efficiency calibration*18 be conducted regularly using a standard point
radiation source, etc. to obtain the efficiency curve. It is also recommended that resolution be
calculated and used to check that these values do not change significantly over time.
3.4.2.1 Calibration using standard point radiation source
3.4.2.1.1
Necessary devices(1) Ge detector and a measurement circuit(2) Standard point radiation source
The nuclide should be selected such that it covers the gamma-ray energy range of themeasurement target.
Table 3.1 illustrates an example of nuclides used for calibration and their gamma-ray
energies.
The magnitude of the radiation source should be approximately 100 kBq. At this level ofradioactivity, it can be used with a simple registration with the authorities as an approved
device with a certification label. It is required that an “application for Approved Devices with
Certification Labels” be submitted to the Nuclear Regulatory Commission within 30 days ofstarting to use it.
The application form of “Approved Devices with Certification Labels” can be obtained fromthe vendor of the radiation source. Alternatively, it can be downloaded from the Nuclear
Regulatory Commission website.
(http://www.nsr.go.jp/activity/ri_kisei/shinsei/shinsei1-1.html)――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*17 Most MCAs can save an energy calibration result taken in advance and use it to display the horizontal

axis of the spectrum in gamma-ray energies.
*18 It is also possible to obtain the peak efficiency by simulation computing (see Explanation A).

8

Table 3.1

Example of nuclides used for calibration and their gamma-ray energies

Nuclides Gamma-ray energy (keV)
Emission ratio
Half-life
241
Am
59.5
0.359
432.6 yrs.
133
*
Ba
81.0
0.355
10.55 yrs.
57
Co
122.1
0.856
271.7 days
139
Ce
165.9
0.799
137.6 days
133
Ba
356.0
0.621
10.55 yrs.
137
Cs
661.7
0.851
30.08 yrs.
54
Mn
834.8
1.000
312.2 days
88
Y
898.0
0.937
106.6 days
60
Co
1173.20.999
5.27 yrs.
22Na
1274.50.999
2.60 yrs.
60
Co
1332.51.000
5.27 yrs.
88
Y
1836.1
0.992
106.6 days
133
*Because the peaks of Ba of 79.6 keV (emission ratio: 0.026) overlap, the total of the emission ratio
is shown here (cited from the Evaluated Nuclear Structure Data File (ENSDF) (March 2016)).

9

(3) Jigs to fix the detector and radiation source
The jig must be able to hold the Ge detector and radiation source at a set distance (1 m ormore) and to alter the angle between the line formed by the radiation source and detector andthe detector axis of symmetry (changeable by 10°–15°).
Figure 3.3 illustrates an example of the jig used in a measurement.
When installing the radiation source, pay attention to its orientation such that it does not renderitself as a shield. The revolution axis of the jig should be aligned with the center of thedetector.

Platform to fix
radiation sources

Center line of thedetector

Fix the radiation source,
ensuring that it (acrylic plate,etc.) does not cause ashielding effect in terms of its
orientation and direction.

Align the revolution axis withthe center line of the detector.

Figure 3.3

Example of a jig

10To correct the angular dependence, measure the detector’s angular dependence of peakefficiency N(θ)/N0.*19 The angular dependence of peak efficiency fluctuates with the gamma-ray
energies. Thus, it is necessary to measure the dependence with a number of gamma-ray
energies.
3.4.2.1.2
Measurement procedures*20(1) Fix the point radiation source on the jig and run a measurement until the net peak countreaches approximately 10000 (multiple point radiation sources may be measured
simultaneously).
(2) Repeat the measurement by varying the angle (between 0° and 90°).
(3) From the measured results, obtain the net peak count rate (s−1) of the target nuclide.
(4) Calculate the gamma-ray fluence rate (cm−2s−1) at the detector’s location based on thenumber of gamma rays per second emitted from the standard radiation source used and the
distance between the detector and the radiation source (see 5.2.1 (1)).
(5) Obtain the peak efficiency per unit fluence rate by dividing the net peak count rate (3) by
the fluence rate (4).
(6) Express the peak efficiency for each gamma-ray energy by the function of the angle θ(standardize by giving θ = 0° the value of 1.0 and apply the least square).
(7) Using the function thus obtained, calculate the angular dependence correction term Nf/N0
for each gamma-ray energy through Equation (5.6) in Chapter 5.
3.4.2.1.3
Angular dependence
If the ratio between the length (L) and the diameter (D) of the Ge crystal close to 1, the detector
is less angle-dependent in terms of the peak efficiency outside the low-energy area (> 200 keV),and the angular dependence correction term becomes close to 1 (see Figures 3.4 and 3.5).
If the detector has an L/D ratio between 0.9 and 1.1 and the low-energy gamma/X rays below
200 keV are excluded from the measurement, then the angular dependence correction term
Nf/N0 will probably be within the range of 0.9 to 1.1. In this case, the angular dependence
calibration is not mandatory, and it can be calibrated solely with the irradiation from the 90°
direction.

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*19 See Chapter 5, Equation (5.6)*20 Procedures (3) to (7) may be automatically performed using the device vendor’s software.

11

1.6
1.2

1.4
L/D=1.02

1.0

L/D=0.59

0.8

Nf/N0

Nf/N0

1.1

L/D=1.31

1.2

1.0
0.9
0.8

0.6

0.344 MeV
0.662 MeV
1.408 MeV

0.7

0.4
10

100

1000

0.5

0.6 0.7

0.8 0.9 1.0 1.1

Gamma-ray energy (keV)

1.2 1.3 1.4

L/D

Figure 3.4 Relationship between
gamma-ray energy and angulardependence correction term
(Nf/N0) (cited from ICRU Rep.
53)

Figure 3.5 Relationship between the
L/D ratio and angulardependence correction
term (Nf/N0) of a Ge
detector (cited from
HASL-300)

3.5 Measures against equipment contamination
3.5.1 Anti-contamination measures
If an in-situ measurement is conducted with contaminated equipment, the measured results may not be
reliable. If the measurement will be conducted at a place with radioactive contamination, or
contamination is not verified in that place, then anti-contamination measures should be taken inadvance. As preventive measures, it is effective to cover the equipment with plastic bags, etc. (seeFigure 3.6). To prevent the equipment from being contaminated during transfer, it is recommended to
double the plastic layers for transportation, as long as this does not interfere with the measuringperformance, use the detector’s cover, etc. It is also important to cover the interior of the transit
vehicle and change the plastic protections frequently to avoid introducing contamination into thevehicle.

Figure 3.6

Example of protecting the equipment

12(1) Protection on the detector
Cut out a large sheet of vinyl plastic to the shape of the detector and cover it, using sealants asnecessary, so that the main component of the detector is not contaminated (see Figure 3.7). It is
important to remember that necessary arrangements are made to enable the handling of the device fortransportation, cable connection, and inlet of liquid nitrogen (only for devices with the liquid nitrogencooling system). If the main component has air intake/ventilation openings, these should be fitted withfilters to prevent dust from entering the device.
In summer, the in-situ measurement may be under strong sunlight, in which case the temperatureinside the plastic bag will increase and affect the detector itself. Therefore, provide a sun shade, etc. asnecessary.

Liquid nitrogen cooling type)Figure 3.7(Electric cooling type)Protection example on the detector(2) Protection of MCA, etc. (except where MCA is mounted inside the main detector)
Using plastic bags, etc. that are larger than the MCA and the PC, protect these devices, with sealant if
necessary, to prevent contamination. The protection must be able to accommodate certain
arrangements with the MCA, as it must be connected to the detector and the PC for control with cables,
as well as attaching a power cable. As the MCA releases heat during operation, the protection must not
be hermetic, but must have some ventilation possibilities. Similar to the detector, it requires some
protection against sun light.
(3) Protection of the cables
For applying protection to the cables, long plastic tubes are useful. Cables are most susceptible todamage while being handled; therefore, the protection must be as robust as the system with two layers,
etc. to prevent breakage and tearing. Pay attention not to allow the protection to place loads on theconnections.
(4) Protection of the detector support rack
For protecting the legs of the detector support rack, long plastic tubes are useful as in the case of thecables. Because these legs will be in constant contact with the ground during an in-situ measurement,
extra measures must be taken such as doubling the plastic layers, using thick plastic bags, etc.

13

3.5.2 Decontamination methods
If the equipment becomes contaminated during an in-situ measurement, etc., decontamination isnecessary. If the contaminated location/parts are identified, wipe off the contaminant. If wiping is noteffective, consider the possibilities of replacing contaminated parts.
To verify the equipment contamination, perform regular background measurements under the same
conditions and ascertain the background levels.
3.6 Other preparations• Start cooling the detector in advance, considering the time needed for cooling it before starting themeasurement. It is desirable, where it is operationally possible, to keep it cool at all times toprevent the degradation of the Ge crystal. If it is stored at the ambient temperature for a long period,
it is necessary to check its efficiency regularly.
• Ensure that the batteries are fully charged (for the detector, portable MCA, and PC). In summer, thebatteries may run out significantly more quickly. Spare batteries will help to extend themeasurement duration in the field.
• If the equipment includes devices that cannot be run on batteries, a mobile generator may be used
as a power supply. However, in this case, verify in advance that the noise does not interfere withthe measurement.
• If using automotive batteries to supply power, it is necessary to verify in advance the batteries’
stability as a power source and the device specifications.

14

Chapter 4

Measurement methods and spectral analysis

4.1 Selection of the measurement site*1
When calculating the radioactivity concentration, the conditions of the place where the measurement istaken has a significant bearing on the analysis results. Thus, it is necessary to consider the following
points when selecting the measurement site.
• For the measurement, it is desirable that the conditions of the measurement site should be
similar to that of the analysis (infinite plane), that is, a flat and wide area (area, ideally a 30-m
radius, and at least 10 m radius).
• The radioactivity concentration measured by the in-situ measurement is the amount depositedon the ground. When comparing with the amount of fallout, etc. the site should be a grassland or
bare ground where radioactive materials are considered to be held and where humaninterventions have not been made, instead of covered surfaces such as tarmacked ground.
• To avoid making the transit vehicles become gamma-ray shields, they should be parked awayfrom the measurement site.
However, note that the site does not have to be an infinite plane if the measurement is for theidentification of nuclides or calculation of the ambient dose rates.
It is practically difficult to find an ideal location that satisfies all the points raised above in the realenvironment for selecting the measurement site. The in-situ measurement is based on certain
assumptions in the analysis in terms of the topography of the surrounding areas, vertical profile ofradioactive materials in the ground, installation height of the detector, etc.; thus, it is necessary tounderstand the impact on the analysis results if the actual conditions differ from these assumptions(see Chapter 6).
4.2 Measurement
4.2.1 Installation of the equipment(1) Set up the Ge detector in the center of the selected measurement site, positioning the device
on a rack facing downward, such that the distance between the ground surface and the
crystal’s geometric center is 1 m, and having studied the ground surface contour, adjust thelegs of the rack to make the detector parallel to the ground surface.
(2) Place the portable MCA and PC at least 3 m away from the detector*2 and connect these withcables. If a table is used for the MCA and PC, it should be of the minimum size to minimize
the shielding effect on the gamma rays that arise from the ground.
(3) Take measures to stabilize the equipment temperature within the range of guaranteed
operation if the weather conditions necessitate it, to prevent the gain fluctuation due toextreme temperatures.*3(4) If necessary, the rack legs can be fixed to the ground to prevent a fall.
4.2.2 Detector startup and measurement procedures(1) Turn on the MCA and PC, and apply a high voltage to the Ge detector. Leave the setup for awhile to warm up. The time required for the warm-up depends on the devices involved.
Know the timing specific to the devices used in advance.
――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1 See Explanation F*2 When taking a measurement at a site that is possibly contaminated, apply anti-contamination measures(see 3.5.1).
*3 Provide sun shades in summer, and thermal insulation covers in winter.

15

(2) Perform a measurement to run equipment adjustments. Normally, a measurement picks up
peaks deriving from natural radionuclides such as potassium-40 (1461 keV), which are used
to adjust gains. In situations following a nuclear accident, however, it is necessary toconsider using artificial radionuclides for the gain adjustments (see Explanation D). During
the measurement, it is recommended that the detector and MCA be operated on batteries toensure electric stability.
(3) Preset the measuring time and start measuring. The measuring time should be approximately
30 min to 1 h, which must be judged by visually checking the measured spectrum (seeExplanation B).
(4) Check the dead time.
(5) Once the measurement commences, do not approach the detector (to avoid shielding of the
gamma rays rising from the ground).
4.2.3 Verifications to be conducted during the measurement
Normally, the gamma-ray spectrum obtained through an in-situ measurement contains peaks ofnatural radionuclides. Therefore, the spectrum can be checked during the measurement using these.
Confirm the peak of potassium-40 (1461 keV) to verify its position (center channel) and FWHM in
the spectrum. If the peak center is off the true center by two channels or more, or the FWHM issignificantly changed, an insufficient warm-up or equipment faults are suspected. Take a second
measurement or run an equipment inspection.
4.2.4 Procedures for finishing the measurement(1) Check if the measuring time has reached the preset value.
(2) Verifying that the data collection has been terminated, save the spectrum in a data file andnote the file name.
(3) Take an overview of the spectrum. If unexpected peaks, noises, etc. are found, investigate
their causes and arrange necessary measures, such as taking a second measurement.
(4) If an analysis is conducted on the in-situ measurement site, refer to “4.4 Spectral analysis.”
Steps (5) to (7) below can be omitted if it is difficult to secure the time for warm-up before themeasurement at a subsequent measurement site, provided that no problems are caused in transferring
the equipment.
(5) Shut down the high-voltage power source and turn off the MCA.
(6) Close the PC and turn off the power.
(7) Disconnect the MCA, detector, and PC and remove the respective connecting cables.
(8) Dismantle the detector from the support rack carefully and place it into a carrying case.
(9) Move to the next measurement site.

164.3 Recording
Take records of the following items in terms of the measurement sites and measurements. Table 4.1
illustrates an example of such a recording format.
4.3.1 Records of the measurement site
(1) Take notes about the descriptions of the surrounding areas.*4(1) Topography (level ground, slanted, etc.)(2) Land use (sports ground, part of precincts of temple, cultivated area, noncultivated area,etc.)(3) Surface conditions (grassed, bare ground, turfed, sand dunes, vegetable farm, orchard,
tarmacked, etc.)(4) Type of the soil (sandy, loamy, clay, etc.)*5(5) Breadth of grasslands, etc.
(6) Condition of the soil(7) Structures in the surrounding areas (distances between the measurement point and thebuildings, size of the building, materials (timber, concrete, etc.))(2) Take photographs of the surrounding areas.
(3) Measure the dose rate near the measurement site with a survey meter and record the readings.
*6(4) If a GPS*7 is available, record the coordinates of the measurement site.
(5) Weather conditions (take detailed notes about precipitation, and if possible, record the winddirection, velocity and ambient temperature)
4.3.2 Records about the measurement(1) Date and start time of the measurement(2) Name of the person who conducted the measurement(3) Detector and measurement equipment (model, serial number, etc.)(4) Data file name of the spectral data measured(5) Center channels, etc. of the main peak*8(6) Characteristics of the spectrum (note if there are irregular forms or increase in FWHM, etc.)――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*4 The descriptions are about the area of approximately 30 m radius around the measurement site.
*5 The soil classification is described in the Radioactivity Measurement Series No. 16 “Methods ofenvironmental sampling.” Take notes of the details if the soil is of a special kind (humic soil, etc.).
*6 This is for the purpose of ensuring that the measurement point is not on a locally high/low-dose-rate

spot.
*7 Global positioning system*8 See Table 4.3 for the radionuclides usually detected and their gamma-ray energies.

17

1.

Table 4.1 Example of recording format
Records of the measurement site
(Describe the required details or put a circle on applicable items. If more space is needed, write in
the Remarks section.)(1)
MeasurementsiteN:(3) CoordinatesE:(2)
Measurement
point
°
°

′
′

(4) Weather
conditions(precipitations
in particular)

(5)
Snow
cover
(cm)

• Sports ground• Precincts of a temple,
etc.
•Sandy-Loamy• Flat ground / slanted ground(8) Type of(6) Topography(7) Land use • Cultivated land ••Clay• Other ()
the soil
Non-cultivated land
•Other (• Other()• Grassland • Bare ground ••Wet(9) GroundTurfed land • Sandy land •(10) Breadth
(11)
•Dry
surface
Farmland • Orchardof grasslands,
Condition•Otherconditions• Tarmacked surface
etc.
of the soil(
• Other ()•Yes / No (within 30 m)• Distance from the buildings:(12) Structuresmin the(13) Dose rate(14)Photograph•Yes / No• Type of the buildings:
surrounding
measurement
taken
Timber; Concrete; Other
areas
(
)
• Number of stories:))(15) Remarks

2. Records of the measurement(Describe the required details or put a circle on applicable items. If more space is needed, write in
the Remarks section.)(1) Measurement start(2) Measurement[month]/[date]/[year]
date
start time(4) Name of theperson who conducted
the measurement(6) File name of thespectral datameasured(8) Main peak center
212Pb
channel(For other nuclides, (239 keV)
describe in the
Remarks section)(9) Detected artificial
radionuclides[hour]:[minute](3)
Measuringtime

seconds(5) Detector/equipment No.

Detector No.
Measurement equipment No.

(7) Irregularities in the
spectrum• None• Yes (

Peak (ch)
214Pb

214Bi

228Ac

40K

(352 keV)(609 keV)(911 keV)(1461 keV)(10) Remarks

18)
FWHM (keV)
40K

4.4 Spectral analysis
The analysis of the spectrum obtained by a Ge detector will focus on the total absorption peak basedon the monochromatic photons found in the spectrum. Although it is possible to perform automatic
analysis using the software provided by the measurement equipment manufacturer, it is necessary toverify the software’s analytical conditions before putting it to use. To avoid errors in quantification due
to changes made in the analytical conditions, it is important to ensure that the same analyticalconditions are applied both to the equipment calibration and in-situ measurements. When alteringcertain analytical conditions, it is recommended that the impact of the condition difference on
analytical results be understood in advance.
4.4.1

Analytical proceduresSpectral measurement

Peak search

Energy calibration

Use the peak of known energy found in the measured spectrum.

Identification of
nuclidesCalculation of peak area

Calculation of radioactivity concentration and ambient dose rate

4.4.2 Peak search, identification of nuclides, and calculation of peak area
The peak search, identification of nuclides, and calculation of peak area are performed in the samemanner as in the case of the general gamma-ray spectrometry. Refer to the Radioactivity
Measurement Series No. 7 “Gamma-ray Spectrometry Using Germanium Detectors” for operation
procedures.
Table 4.2 lists the radionuclides generally detected under normal circumstances.
Table 4.2 Radionuclides normally detected through in-situ measurements

Naturally occurring radionuclides
Uranium series214

Pb 214Bi

Artificial radionuclides

Thorium series

Other

Tl 212Pb 212Bi 228Ac

40

208

K 7Be

137

Cs

As many peaks of artificial radionuclides are expected to be detected in a nuclear accident, data
analyses for nuclide identifications and peak area calculations must be conducted carefully.

19

An example of an in-situ measurement spectrum is shown in Explanation C.
4.4.3 Energy calibration
Energy calibration needs special precautions for in-situ measurements, as it is conducted outdoors,which means that it is exposed to temperature fluctuations and the measurement must becommenced relatively quickly after turning on the power of the equipment. There is a risk thatpeaks may move owing to the variation in gains caused by changes in temperatures subsequent to
the lab-based energy calibration using gamma-ray radiation sources. Therefore, conduct thecalibration again in the field using the known peak detected in the spectrum obtained through an
in-situ measurement. In the in-situ measurement, it is often the case that certain naturally occurring
radionuclides are detected in the spectrum. Use the peak of such a nuclide to carry out the energy
calibration. Table 4.3 illustrates an example of these nuclides and their gamma-ray energies. Whena nuclear accident occurs, the background radiation level rises and some radionuclides in Table 4.3,
especially those in the low-energy region, become unsuitable for calibrations. In such a case, it may
be necessary to perform the energy calibration using gamma-ray energy of the artificial
radionuclides emitted through the accident, as seen in Table 4.4 (see Explanation D).
Table 4.3

Naturally occurring radionuclides usable for energy calibration and their gamma-ray
energies
Gamma-ray energy (keV)

Emission ratio*

239
352

0.434
0.369

583
609

0.306
0.469

Ac
K

911
1461

0.290
0.107

Tl

2615

0.359

Nuclides
212

Pb
Pb

214
208

Tl
Bi

214
228
40
208

* 238U and 232Th series show the per-decay emission rates of the parent nuclides(cited from ICRU Rep. 53, Table 3.4)

Table 4.4

Artificial radionuclides usable for energy calibration in a nuclear accident and their
gamma-ray energies
Nuclides
134
Cs
137

Cs
Cs

134

110m

Ag
Ag

110m
110m

Ag

Gamma-ray energy (keV)
605

Emission ratio
0.975

662
796

0.849
0.851

885
1384

0.729
0.243

1505

20

0.131
(cited from ICRU Rep. 53, Table A.1)

Chapter 5 Calculation of radioactivity concentration and ambient dose rate
5.1 Conditions for analysis
The in-situ measurement enables calculation of the radioactivity concentration (radiation per unit
area: Bq/cm2) of the radioactive materials deposited on the ground and ambient dose rate. The
HASL analytical method developed by Beck et al. (hereafter, the “HASL method”) has been usedworldwide for the evaluation of such data, and it is employed by the ICRU. Therefore, this manual
also adopts the HASL method for conducting analyses.
The HASL method uses certain assumptions to calculate the radioactivity concentration andambient radiation dose rate, such as the topography, vertical profile of radioactive materials in the
ground, and the detector installation height. Table 5.1 lists these assumptions.
Table 5.1 Assumptions used in in-situ measurement
Condition

Description

Topography of thesurrounding areas
Vertical profile ofradioactive materials in
the ground
Height of the detector

A flat and wide, infinitely open area
(infinite plane)
The distribution expressed by Equation (5.1)
1 m from the ground surface

The radioactivity concentration A(Z) at mass depth Z in the ground is expressed as follows:

A(Z ) A0 exp(

Z

)
(5.1)

A(Z)
Z

A0
β

: Radioactivity concentration (Bq/g) at mass depth Z: Mass depth (g/cm2)
The depth measured from the ground surface by per-unit-area mass ofthe soil.
: Radioactivity concentration (Bq/g) at the ground surface: Weight buffer depth (g/cm2)
A parameter to indicate the vertical profile of radioactive materials in
the ground. It expresses the degree of penetration, where greater
values indicate deeper penetration.

The value of β is the mass depth at which the radioactivity concentration is 37% of that at the
ground surface (= 1/e); it is almost zero for ground surface distribution and infinite for anunderground homogeneous distribution in the ground. The exponential model is simply anapproximation, but it suggests relatively realistic values after a certain time has elapsed since theradioactive deposition.
As time passes, the distribution may present peaks at certain depths owing to the transition or
diffusion of the radioactive materials, or the distribution may cease to be exponential as the land issubjected to use, erosion, or decontamination. However, even in such cases, a practically valid β
can be employed to associate the measured results on the ground with the radiation sources in the
ground (see Explanation E).

21

The radiation (deposition amount) per unit area Aa is expressed as follows:

Aa

A0

(5.2)
2

Aa : Radioactivity per unit area (Bq/cm )5.2 Calculation of radioactivity concentration
5.2.1 Analysis of radioactive materials deposited on the groundFor evaluating the radioactive materials deposited on the ground, the radioactivity concentration
Aa (radiation per unit area: Bq/cm2) can be calculated by the following equation.

Aa
Aa
Nf
Nf/Aa

Nf /
=

Nf
Aa

(5.3)

: Radioactivity per unit area (Bq/cm2): Peak count rate of energy E (s−1) in in-situ measurement: Efficiency in in-situ measurement

Nf/Aa is calculated by the following equation:

Nf

N0 N f φ
Aa = φ N0 Aa

N0

:

φ

:

(5.4)

Peak count rate (s−1) of the gamma ray of energy E that enters the detector at
the angle (0°) to the detector axis of symmetry
Fluence rate (cm−2s−1)(1) N0/φ
N0/φ indicates the peak count rate (s−1) at the fluence rate (cm−2s−1) at the angle of thedetector axis of symmetry (0°). This value is dependent on the detector and calculated by the
gamma-ray radiation source measured for each detector.
• Install several radiation sources with varying energies at distances of 1 m or more along thedetector axis of symmetry.
• Perform the measurement and calculate N0.
• φ is calculated by the following equation:

22

S a
r 2 e μa x e μh y
φ= 4
S : Radioactivity of the gamma-ray radiation source (Bq)
A : Gamma-ray emission ratio

(5.5)

μa
: Attenuation coefficient in air for gamma rays (cm−1)

x : Distance from the radiation source to the detector front cap (cm): Attenuation coefficient of gamma rays in radiation source holder (in the
ground) (cm−1)
y : Distance through radiation source holder that gamma rays (in the
ground) pass on the way to detector (cm)*1r : Distance from the radiation source to the detector’s effective crystal
center (cm)1) If the gamma-ray energy is over 1 MeV, the detector’s effective crystal
center is the geometric center of the Ge crystal.
2) If the gamma-ray energy is less than 0.1 MeV, the detector’s effective
crystal center is the Ge crystal face.
3) Where the energy is in the range between 1) and 2) above, use thefollowing equation to obtain r.
μh

1 1
r= μ•

μ

e μd (μd 1)d1 e-μd+ 0 +x: Attenuation coefficient in Ge crystal for gamma rays (cm−1)

d : Ge crystal thickness (cm)d0

: Distance from detector cap to crystal face (cm)An example of the calculation to obtain distance (r) from the radiation source to thedetector’s effective crystal center is illustrated in Table 5.2.1.
ax
Table 5.2.2 illustrates an example calculation of attenuation ( e ) by air.
Tables 5.2.3 and 5.2.4 illustrate the coefficients in the radiation source holder (in the ground)for gamma rays (μh) and the soil compositions used in this calculation, respectively.
Calculate N0/φ as a function of energy. An example is shown in Figure 5.1.

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1 When determining the distance between the source and the detector, it is useful to introduce the notion of

the “effective center” of the detector, as the detectors have finite dimensions. There are several ways toobtain the effective center. Here, the method from the IAEA’s TECDOC-1092 (1999) is introduced.

23

Table 5.2.1 Calculation to obtain distance (r) from the radiation source to the detector’s effective
crystal center
Unit: (cm)Gamma-ray
Ge crystal thickness d (cm)μenergy(cm−1)
4
5
6
7
8
(MeV)
0.1

2.94

100.8

100.8

100.8

100.8

100.8

0.15

1.32

101.2

101.3

101.3

101.3

101.3

0.2

0.883

101.5

101.6

101.6

101.6

101.6

0.3

0.601

101.8

101.9

102.0

102.1

102.1

0.4

0.496

101.9

102.1

102.2

102.3

102.4

0.5

0.437

101.9

102.2

102.3

102.4

102.5

0.6

0.397

102.0

102.2

102.4

102.6

102.7

0.8

0.342

102.1

102.3

102.5

102.7

102.9

1

0.305

102.1

102.4

102.6

102.8

103.0

Distance from the radiation source to the detector front cap =100 cmμ : Attenuation coefficient in Ge crystal for gamma rays
(cm−1)x:d0: Distance from detector cap to Ge crystal face (cm) = 0.5 cm

Table 5.2.2

Calculation of attenuation ( eax) by airGamma-ray energy(MeV)

Linear attenuation coefficient at
20 °C*(×10−4 cm−1)

Attenuation by air at a
distance of 1 m

0.06
0.08
0.1
0.15

2.159
1.945
1.820
1.616

0.979
0.981
0.982
0.984

0.2
0.3

1.476
1.281

0.985
0.987

0.4
0.5

1.148
1.048

0.989
0.990

0.6
0.8

0.96890.8513

0.990
0.992

1
1.5

0.76490.6227

0.992
0.994

2

0.5351

0.995*PHOTX database24

Table 5.2.3

Attenuation coefficient for gamma rays in radiation source holder (in the ground)(μh)
Gamma-ray energy
Mass attenuation
Linear attenuation*(keV)
coefficient of soil
coefficient of soil**(cm2/g)(cm−1)
20
2.78
4.45
25
1.52
2.43
30
0.938
1.50
35
0.644
1.03
40
0.471
0.754
45
0.381
0.610
50
0.314
0.502
55
0.277
0.443
60
0.248
0.397
65
0.230
0.368
70
0.214
0.342
75
0.202
0.323
80
0.190
0.304
85
0.185
0.296
90
0.178
0.285
95
0.173
0.277
100
0.167
0.267
150
0.139
0.222
200
0.125
0.200
250
0.115
0.184
300
0.108
0.173
350
0.101
0.162
400
0.09630.154
450
0.09190.147
500
0.08750.140
550
0.08440.135
600
0.08130.130
6500.07880.126
7000.07560.121
750
0.07310.117
800
0.07130.114
850
0.06940.111
900
0.06750.108
9500.06500.104
1000
0.06380.102
1500
0.05210.08342000
0.04490.07182500
0.0401
0.06423000
0.03640.0582*

Cited from HASL-258
The soil density is given at 1.6 g/cm3 for the calculation**

25

Table 5.2.4 Soil composition
Component

Composition

Al2O3

13.5 wt%

Fe2O3

4.5 wt%SiO2

67.5 wt%

CO2

4.5 wt%H2O

10 wt%ρ (density)

1.6 g/cm3(Cited from HASL-258)

Figure 5.1 Fluctuation by energy of count rate per fluence rate (N0/φ) in several detectors ofdifferent relative efficiencies
P1 to P5 and O1 to O3 are different types of detectors, and the values to the right indicate
the relative efficiencies.
(Cited from HASL-300)

26

(2) Nf/N0
Nf/N0 is a component of the directional dependence correction of a detector, and it iscalculated by the following equation, considering the direction of incidence ofenvironmental gamma rays and the detector’s directional dependence:NfN0
φ(θ)/φ:N(θ)/N0 :20( ) N( )
d
N0

(5.6)

The proportion in which an environmental gamma ray of energy E
enters the detector at angle θ in a given geometry. To illustrate, some
calculations of cesium-137 (662 keV) in several geometries are shownin Figure 5.2. Note that this value does not vary significantly dependingon the gamma-ray energy.
The relative sensitivity (to a baseline of 0°) of a gamma ray of energy E
with respect to the angle of incidence θ. The calculation is repeatedseveral times using different gamma-ray radiation sources of varying
energy and by changing the angle of incidence.

27

Gamma-ray fluence (relative value)

Gamma-ray energy: 662 keV

Angle (°)

Figure 5.2

Distribution of gamma-ray incidence direction 1 m above ground levelα/ρ in the figure is an inverse of β in Expression (5.1)(Cited from JAERI-M 6498)

For detectors with Ge crystals of approximately equal dimensions (length (L) and diameter(D)), their sensitivity to angular dependence is small; thus, the correction term Nf/N0 is close
to 1. In the in-situ measurement, many gamma rays enter the detector at almost horizontal
angles. For these reasons, it is possible to calibrate the equipment with the irradiation only inthis direction. In this case, Equation (5.4) takes a simple form as follows:

Nf

N90 φ
Aa = φ Aa

N90

(5.7)

: Peak count rate (s−1) of the gamma ray of energy E incident at rightangle (90°) to the detector axis of symmetry.
In the in-situ measurement, the detector is usually set in a
downward-facing position, and many environmental gamma rays are
injected horizontally. Therefore, N90 = Nf can be assumed.

28

(3) φ/Aa
φ/Aa indicates the relationship between the radioactivity concentration in the ground (Aa) andthe detection position (at a height of 1 m) in terms of the gamma-ray fluence rate, the value of
which varies depending on the radiation source distribution. Table-1 illustrates the φ/Aa valuesof major nuclides by the vertical profile in the ground of radioactive materials (β).
Ideally, β is obtained through measurements, but this is usually very difficult to do.
Alternatively, therefore, the values indicated in Table 5.3 can be in principle applied,according to the time elapsed since the radioactive deposition and the amount of precipitation
as a guide. If the ground surface is nonpermeable (rooftop, asphalt, and concrete), it isappropriate to assume β to be 0.1 g/cm2. The values in Table 5.3 can only be used if the soil is
not of a special kind (humic soil, etc.) and not disturbed by human activities. Moreover, theaccuracy of the analysis of in-situ measured results can be enhanced by sampling the soil at
the measurement site by soil layers and evaluating the radioactive distribution by depth.
Note that details are given in Explanation E, because the accurate understanding of thevertical profile of radioactive materials in the ground is the most effective way to improve thereliability of in-situ measurements.

Table 5.3 Recommended values for vertical profile of radioactive materials in the ground

Time elapsed since
deposition (years)

Precipitation
after deposition(mm)Vertical profile parameterβ (g/cm2)

0~1

<3

0.1

0~1

3

1.0

1~5

-

3.0

5~20

-

10
(Cited from ICRU Rep.53)

29

5.2.2 Analysis of radioactive materials evenly distributed in the groundThe same calculation method can be applied to the naturally occurring radioactive materials thatare considered to be distributed evenly in the ground, such as the uranium series, thorium series,
and potassium-40. However, radioactivity concentration is expressed by activity per unit mass Am,
where the unit is Bq/g. Table 5.4 lists the φ/Am calculated assuming a homogeneous distribution.
In most natural environments, it is valid to assume that naturally occurring radioactive materials are
distributed evenly in the ground. However, artificial structures standing nearby also containnaturally occurring radioactive materials; thus, their influence must be considered.
Note that, when analyzing nuclides of the uranium series, the radioactive equilibrium must be
considered. As radon-222 dissipates from the soil, its product nuclides such as lead-214 andbismuth-214 may indicate low values. While the dissipation rate of radon-222 is generally
approximately 15%, the extent of underestimation cannot be generalized owing to the influence of
airborne disintegration products. Similarly, precipitation collects radon-222 and its disintegration
products onto the ground surface. Therefore, measurements taken in rain or within several hours
after precipitation are not suitable for analyses of uranium-series nuclides in the ground.
Radium-226 of the uranium series releases a gamma-ray of energy 186 keV, which is identical tothat of radium-235 (186 keV), and thus is not usable for quantification.

30

Table 5.4 Relationship between radionuclide activity in the ground and gamma-ray fluence rate 1 m
above ground (φ/Am)(Assuming a homogeneous distribution of radioactive materials in the ground)
EnergyEmissionφ/Am
SeriesNuclides(MeV)
ratio*(cm−2s−1)/(Bq/g)
238
214
U
Pb
0.295
0.192
0.828
Same as above
0.352
0.369
1.71
214
Bi
0.609
0.469
2.75
Same as above
0.665
0.01580.0965Same as above
0.768
0.04970.325
Same as above
0.934
0.0319
0.229
234m
Pa
1.001
0.008450.0629
214
Bi
1.120
0.155
1.22
Same as above
1.238
0.06100.507
Same as above
1.378
0.04100.361
Same as above
1.408
0.02500.223
Same as above
1.509
0.02200.203
Same as above
1.730
0.03000.298
Same as above
1.765
0.162
1.62
Same as above
1.847
0.0216
0.222
Same as above
2.119
0.01250.138
Same as above
2.204
0.05250.592
Same as above
2.448
0.0162
0.193
232
212
Th
Pb
0.239
0.434
1.73
224
Ra
0.241
0.0397
0.158
228
Ac
0.338
0.120
0.547
Same as above
0.463
0.0464
0.241
208
Tl
0.511
0.08090.438
Same as above
0.583
0.306
1.76
212
Bi
0.727
0.0675
0.430
228
Ac
0.795
0.0484
0.322
208
Tl
0.861
0.0453
0.313
228
Ac
0.911
0.290
2.060
Same as above
0.965
0.05450.398
Same as above
0.969
0.175
1.282
Same as above
1.588
0.0371
0.352
212
Bi
1.621
0.0149
0.143
228
Ac
1.630
0.0195
0.187
208
Tl
2.615
0.359
4.418
40
40
K
K
1.461
0.107
0.971
*238U and 232Th series show the per-decay emission rates of the parent nuclides(Cited from ICRU Rep. 53)

31

5.3 Calculation of ambient dose rate
For the radioactive materials deposited on the ground surface or distributed evenly in the ground,the ambient dose rate measured 1 m above ground level (“dose rate”) is calculated by the following
equation:

I=
I
Nf

Nf /

Nf
I

(5.8)

: dose rate (ambient absorbed dose rate: μGy/h, or peripheral dose equivalent rate:
μSv/h): peak count rate of energy E (s−1) in in-situ measurement

Nf/I is calculated by the following equation:

N0 N f φ
I = φ N0 I

Nf

(5.9)

N0
: peak count rate (s−1) of gamma-ray incident from the detector axis of symmetry
direction (0°)φ: fluence rate (cm−2s−1)
Equation (5.9) is the same as Equation (5.4), except that the radioactivity concentration A is
replaced with the dose rate I. φ/I is a coefficient that expresses the relationship in a given geometry
between the gamma-ray fluence rate of the gamma ray of energy E emitted by a radioactive
material and the total dose rate of this radioactive material. The dose rate I must include not only
the primary gamma ray of energy E, but also contributions by scattered rays and all other gamma
rays of the radioactive material. For the scattered ray evaluation, it is necessary to use either theMonte Carlo method or Lattice Boltzmann transport equation. Values for φ/I of radioactive
materials exponentially distributed in the ground are listed in Table-2-1 (air absorbed dose rate:
μGy/h unit) and Table-2-2 (peripheral dose equivalent rate: μSv/h unit). Table 5.5 lists φ/I in thecase of homogeneous distribution in the ground (naturally occurring radioactive materials). φ/I for
uranium- and thorium-series nuclides indicates the relationship between the fluence rate of a given
gamma ray released by a given nuclide within the series and the dose rate of gamma rays from all
nuclides within the series. It follows that the dose rate of a given gamma ray of a given nuclide is
the same as the dose rate of all the nuclides within that series.*1 When analyzing either the gamma
rays emitted by several nuclides of the same series (e.g., 214Pb and 214Bi of the uranium series) orseveral gamma rays emitted by a single nuclide (214Bi: 609 keV, 1765 keV, etc.), the average dose
rate of those gamma rays*2 is used as the series dose rate.
The dose rate can also be calculated from radioactivity concentration. Conversion factors in thecase of exponential distribution are listed in Table-3-1 (air absorbed dose rate: μGy/h unit) and――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1 This enables obtaining the results equivalent to those by measuring up to 3 MeV using a NaI(Tl)scintillation detector, for the dose rate calculated through the analysis of gamma rays of less than 2 MeV
includes contributions by gamma rays of more than 2 MeV emitted by other nuclides of the same series,
even though the energy range for in-situ measurement is limited to a maximum of 2 MeV.
*2 Take weighted average using a counting error.

32

Table-3-2 (peripheral dose equivalent rate: μSv/h unit). Table 5.6 indicates the conversion factors(air absorbed dose rate: μGy/h unit) in the case of homogeneous distribution (naturally occurring
radioactive materials). The total values for the uranium series and thorium series in Table 5.6 are
calculated assuming that the series nuclides are in radioactive equilibrium.*3――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*3 Although the uranium series includes gaseous radon, which suggests a possibility that radioactive

equilibrium is not established, this does not usually pose a problem as the in-situ measurement measures
the post-radon-222 products of lead-214 and bismuth-214, and also because dose contributions are
attributed mainly to these decay products.

33

Table 5.5 Relationship between dose rate of radioactive materials evenly distributed in the groundmeasured 1 m above ground and gamma-ray fluence rate (φ/I)
Energyφ/I−2 −1SeriesNuclides(MeV)(cm s )/(μGy/h)
238
214
U
Pb
0.295
1.79
Same as above
0.352
3.70
214
Bi
0.609
5.95
Same as above
0.665
0.209
Same as above
0.768
0.703
Same as above
0.934
0.496
234m
Pa
1.001
0.136
214
Bi
1.120
2.64
Same as above
1.238
1.10
Same as above
1.378
0.781
Same as above
1.408
0.483
Same as above
1.509
0.439
Same as above
1.730
0.645
Same as above
1.765
3.51
Same as above
1.847
0.481
Same as above
2.119
0.299
Same as above
2.204
1.28
Same as above
2.448
0.418
232
212
Th
Pb
0.239
2.86
224
Ra
0.241
0.262
228
Ac
0.338
0.906
Same as above
0.463
0.399
208
Tl
0.511
0.725
Same as above
0.583
2.91
212
Bi
0.727
0.712
228
Ac
0.795
0.533
208
Tl
0.861
0.518
228
Ac
0.911
3.41
Same as above
0.965
0.659
Same as above
0.969
2.12
Same as above
1.588
0.583
212
Bi
1.621
0.237
228
Ac
1.630
0.310
208
Tl
2.615
7.31
40
40
K
K
1.461
23.3
Dose rate (I) is regarded to be of all the nuclides within the series,
assuming the establishment of radioactive equilibrium among them.

34

Table 5.6 Relationship between radioactivity concentration in the ground measured 1 m above groundand dose rate (I/Am)*(Assuming a homogeneous distribution of radioactive materials in the ground)

Nuclides238
U series
238
U
234
U
234
Th
234m
Pa
234
Pa
230
Th
226
Ra
222
Rn
214
Pb
214
Bi
210
Tl
210
Pb
Total
232
Th series
232
Th
228
Ra
228
Ac
228
Th
224
Ra
220
Rn
212
Pb
212
Bi
208
Tl
Total
40
K

Kerma rate per radioactivity concentration(μGy/h)/(Bq/g)
4.33/10-5
5.14/10-5
9.47/10-4
3.00/10-3
4.49/10-4
6.90/10-5
1.25/10-3
8.78/10-5
5.46/10-2
4.01/10-1
1.15/10-4
2.07/10-4
4.62/10-1
4.78/10-5
5.45/10-5
2.21/10-1
3.44/10-4
2.14/10-3
1.73/10-4
2.77/10-2
2.72/10-2
3.26/10-1
6.04/10-1
4.17/10-2*

The values assume the establishment of radioactive equilibrium.
(Cited from ICRU Rep.53)

35

Chapter 6 Interpretation of measured results
The in-situ measurement is based on certain assumptions in the analysis in terms of the topography ofthe surrounding areas, installation height of the detector, etc.; thus, it is necessary to understand theimpact on the analysis results if the actual conditions differ from these assumptions.
6.1 Implications of discrepancies between analytical and actual measurement conditions
6.1.1 Breadth of surrounding areas
The coefficients required for the calculation of radioactivity concentration from an in-situ
measurement (φ/A in Equation 5.4 and the values in Table-1 and Table 5.4) assume an infinitely open
and flat landscape (infinite plane) with no obstacles that attenuate gamma rays. In reality, however, it
is impossible to expect such a perfect infinite plane; therefore, the analysis of radioactivity
concentration based on this assumption of an infinite plane will result in underestimation. Figure 6.1
illustrates the peripheral contribution ratio in the gamma-ray fluence rate 1 m above the ground withan assumption that cesium-137 is exponentially distributed in the ground (β: 4.8 g/cm2). The
contributions from the area of 10 m radius from the detector are equal to 85% of the total contribution
in an infinite plane. The extent to which the fluence rate is underestimated because the measured area
is not an infinite plane depends on the vertical profile of radioactive materials in the ground. To be
more precise, it also depends on the gamma-ray energies. Details are given in Explanation F.1.
The underestimation may be corrected by the correction coefficients given in Table 6.1, provided thatthe area of radioactive deposition can be identified. However, in typical cases, the extent of theunderestimation would be within some tens of percent, even if the correction is not applied. However,note that if β is 0.1 g/cm2 and the deposition is confined within a small area of 10 m radius, correction
must be considered as the result would be significantly underestimated.
To do this, measure the radioactive deposition area (an average radius) centering on the detector, using
a tape measure, etc. and multiply the measured value of the concentration by the correction coefficientfrom Table 6.1. The correction coefficients included in Table 6.1 are applicable to gamma rays of 600
keV, and they can be used for gamma rays of other energies as the difference in the correction
coefficients with respect to energy variation is not considerable. In cases where artificial radioactive
materials as have been deposited as fallout for some time, and it is more likely to be on grasslands orbare grounds than tarmacked surfaces or buildings, then correction must be performed based on thearea over grasslands, bare grounds, etc. When a deposition area cannot be identified, such as during
the early stages of a nuclear accident, correction is difficult to achieve. Thus, it must be performed
after the deposition has taken place.
However, note that correction is not obligatory if the in-situ measurement is for the purpose of
fixed-point fluctuation surveillance.
Similarly, dose rate values need not be corrected as the differences in the breadth of surrounding areas
have a little impact on dose rates.
6.1.2 Height of detector installation
The detector is normally installed at a height of 1 m; thus, correction is not necessary. The impact of
the installation height of the detector on the measured value of radioactivity concentration is describedin Explanation F.2.

36

6.1.3 Moisture in the ground
Moisture in the ground does not need correction, as it is included in the effects of the vertical
distribution of radioactive materials in the ground (see Explanation E). The impact of the in-soil
moisture on the measured value of radioactivity concentration is described in Explanation F.3.

Figure 6.1 Peripheral contribution ratio in the gamma-ray fluence 1 m above ground with anassumption that cesium-137 is exponentially distributed in the ground (β: 4.8 g/cm2) (cited from
HASL-300)

Table 6.1 Correction coefficients for the breadth of surrounding areas
Vertical profile parameterβ (g/cm2)

Area of radioactive deposition (radius: m)
10

15

20

25

0.100

1.6

1.4

1.3

1.2

1.00

1.3

1.2

1.1

1.1

3.00

1.2

1.1

1.1

1.1

10.0

1.1

1.1

1.0

1.0

4.8

1.2

1.1

1.1

1.0

∞ (homogeneous distribution)

1.1

1.0

1.0

1.0

37

6.2 Management of accuracy in measured result
6.2.1 Routine and periodical inspections
To achieve reliable measurement readings from in-situ measurement, it is desired that the following
routine and periodical inspections are performed.
Daily inspection• If the detector is cooled at all times, check the device temperature (if monitoring of thetemperature is possible) and liquid nitrogen consumption (for liquid nitrogen cooling systemonly).
• Verify the energy resolution using a radiation source such as cobalt-60.
Periodical inspection• Verify the peak efficiency (see 3.4.2).
• Take the detector in for a manufacturer servicing, etc.
6.2.2 Intercomparison measurement
As part of the measurement accuracy management to improve the reliability of in-situ measurement
readings, the intercomparison measurement of measurement equipment is an effective method (seeExplanation I). Urgent in-situ measurement in a nuclear disaster is well anticipated, and it is important
to ascertain the equipment conditions regularly and routinely. If multiple devices of measuring
equipment are employed in surveying a wide area for radioactive deposition distribution, it isadvantageous to verify in advance that all pieces of equipment are equal in terms of measurement
accuracy.

38

Explanation

39

40

Explanation A
Explanation A.1

Calculation of peak efficiency by simulated calculation

OutlineIt is possible to obtain the peak efficiency by simulation computing without using a standard point
radiation source. For the computation, peak efficiency simulation software is often used, employing
the MCNP Monte Carlo code, etc. By modeling the detector in preparation in terms of the relationshipbetween the gamma-ray energy and geometrical conditions (distance, angle of incidence, etc.), it ispossible to quickly prepare a peak efficiency that is compatible with a given energy, shape, and size ofthe sample.
However, it is different from the peak efficiency calibration using a traceability-ensured standard point
radiation source, and it is desirable that the reliability of the simulation-computed peak efficiency beverified regularly by comparing it with the peak efficiency prepared based on a standard point
radiation source.
For its use, give thorough consideration to the computing requirements, including the detector
information, and verify the validity of the computing result thoroughly before using it with caution.
The simulation requires modeled data of the detector; thus, it is important to manage the detector suchthat its attributes are not altered. Specifically, it is desirable that a Ge detector be cooled at all times
and managed such that its dead layer is not altered.

41

Explanation A.2

Comparison with the peak efficiency from a standard point radiation source

Table A.1 shows a comparison between the peak efficiency prepared using a standard point radiation
source measured with a Ge detector of relative efficiency 30.1% (hereafter, “radiation source
efficiency”) and peak efficiency based on simulation*1 *2 (hereafter, “simulation efficiency”).

Table A.1

Comparison between radiation source efficiency and simulation efficiency(^1^)Nuclides(^2^)Energy(keV)(^3^)Radiation source efficiency(A)(^4^)Simulation efficiency(B)(^5^)Ratio(B) / (A)(^6^)Am-241

59.5

1.729E-04

1.613E-04

0.933(^9^)Ba-133

81.0

1.676E-04

1.666E-04

0.994(^10^)Ba-133

356.0

8.353E-05

8.201E-05

0.982(^13^)Cs-137

661.7

5.119E-05

4.925E-05

0.962(^14^)Co-60

1173.23.060E-05

3.108E-05

1.016(^15^)Co-60

1332.52.715E-05

2.832E-05

1.043

In the energy range approximately between 60 and 1300 keV, the difference between the radiation
source efficiency and simulation efficiency was approximately within 5%.
Using the spectrum from the in-situ measurement, the radioactivity concentration and dose rate wereanalyzed in terms of the uranium series, thorium series, 40K, 134Cs, and 137Cs using their respective
peak efficiencies. The comparison results are presented in Tables A.2 to A.6. Note that the nuclides
analyzed of the uranium and thorium series are 214Bi and 214Pb for the former, and 208Tl and 228Ac forthe latter.

Table A.2

Comparison of in-situ measurement spectrum analytical results (grassland 1 - β: 1.4
g/cm2)(^23^)Weight buffer depth

β 1.4(^24^)Radiation source efficiency(^25^)Simulation efficiency(^27^)(A)(^30^)Radioactivity
concentration(^26^)Ratio(^28^)(B)
rate(^32^)Proportionsof component
nuclides(^39^)nGy/h(^40^) ％(^31^) Dose(^33^)Radioactivity
concentration(^29^)(B) / (A)(^34^)
Dose rate(^35^)Proportionsof component
nuclides(^41^)
nGy/h(^42^) ％(^36^)Radio
concentration(^37^)
Dose rate(^38^)Proportionsof component
nuclidesactivity(^43^)U series(^44^)-

3.7

7.2(^45^)-

3.3

6.3(^46^)-

0.89

0.87

(^47^)Th series(^48^)-

5.8

11.3(^49^)-

6.2

11.6(^50^)-

1.05

1.03

(^51^)K-40

1.99E-01(^52^)Bq/g

8.3

16.1

1.85E-01(^53^)Bq/g

7.7

14.5

0.93

0.93

0.90

(^54^)Cs-134

2.98E+09(^55^)Bq/km2 12.4

24.0

3.18E+09(^56^)Bq/km2 13.2

24.9

1.07

1.07

1.04

(^57^)Cs-137

1.32E+10(^58^)Bq/km2 21.3

41.3

1.40E+10(^59^)Bq/km2 22.6

42.7

1.06

1.06

1.03

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1

For the simulation, peak efficiency simulation software (calculation code: MCNP Monte Carlo code*2)
was used.
*2
Briesmeister, J. F., “MCNP-A General Monte Carlo N-particle Transport Code Version 4C”, Los Alamos
National Laboratory Report LA-13709-M (2000)

42

Comparison of in-situ measurement spectrum analytical results (gravel - β: 1.4 g/cm2)

Table A.3(^85^)Radiation source efficiency(^86^)Simulation efficiency(^88^)(A)(^91^)Radioactivity
concentration(^87^)Ratio(^89^)(B)(^92^)Dose
rate(^93^)Proportionsof component
nuclides(^100^)
nGy/h(^101^) ％(^94^)Radioactivity
concentration(^90^)(B) / (A)(^95^)D
ose rate(^96^)Proportionsof component
nuclides(^102^)
nGy/h(^103^) ％(^97^)Radioact
ivity
concentration(^98^)D
ose rate(^99^)Proportionsof component
nuclides(^104^)U series(^105^)-

37.1

36.4

(^106^)-

36.6

34.8

(^107^)-

0.99

0.96

(^108^)Th series(^109^)-

34.8

34.1

(^110^)-

37.1

35.3

(^111^)-

1.07

1.03

(^112^)K-40

1.09E-01(^113^)Bq/g

4.5

4.4

1.01E-01(^114^)Bq/g

4.2

4.0

0.93

0.93

0.90

(^115^)Cs-134

2.22E+09(^116^)Bq/km2

9.2

9.0

2.37E+09(^117^)Bq/km2

9.8

9.4

1.07

1.07

1.03

(^118^)Cs-137

1.01E+10(^119^)Bq/km2

16.3

16.0

1.07E+10(^120^)Bq/km2

17.4

16.5

1.06

1.06

1.03

102.0

(^123^)-

105.1(^125^)-(^126^)-(^127^)-(^128^)-(^122^)-(^121^)Total

Table A.4(^124^)-

Comparison of in-situ measurement spectrum analytical results (tarmacked - β: 1.4 g/cm2)(^146^)Radiation source efficiency(^147^)Simulation efficiency(^149^)(A)(^150^)(B)(^152^)Radioactivity
concentration(^153^)
Dose rate(^154^) Proportionsof component
nuclides(^161^)nGy /h(^162^) ％(^148^)Ratio(^151^)(B) / (A)(^155^)Radioactivity concentration(^157^) Proportionsof component
nuclides(^163^)nGy /h(^164^) ％(^158^)
Radioactivity
concentration(^159^)Dose rate(^160^) Proportionsof component
nuclides(^165^)U series(^166^)-

10.1

20.7

(^167^)-

9.7

20.0

(^168^)-

0.96

0.96

(^169^)Th series(^170^)-

10.9

22.3

(^171^)-

11.6

23.8

(^172^)-

1.06

1.06

(^173^)K-40

4.00E-01(^174^)Bq/g

16.7

34.0

3.71E-01(^175^)Bq/g

15.5

31.7

0.93

0.93

0.93

(^176^)Cs-134

9.85E+08(^177^)Bq/km2

4.1

8.4

1.05E+09(^178^)Bq/km2

4.4

8.9

1.07

1.07

1.07

(^179^)Cs-137

4.43E+09(^180^)Bq/km2

7.2

14.6

4.72E+09(^181^)Bq/km2

7.6

15.6

1.06

1.06

1.07

49.0

(^184^)-

48.8(^186^)-(^187^)-(^188^)-(^189^)-(^183^)-(^182^)Total

Table A.5(^185^)-

Comparison of in-situ measurement spectrum analytical results (grassland 2 - β: 1.4
g/cm2)(^207^)Radiation source efficiency(^208^)Simulation efficiency(^209^)Ratio(^211^)(B)(^212^)(B) / (A)(^210^)(A)(^213^)Radioactivity
concentration(^214^)
Dose rate(^215^)Proportionsof component
nuclides(^216^)Radioactivity
concentration(^217^)
Dose rate(^218^)Proportionsof component
nuclides(^219^)Radioa
ctivityconcentration(^220^)
Dose rate(^221^)Proportionsof component
nuclides(^222^)nGy /h(^223^) ％(^224^)nGy /h(^225^) ％(^226^)U series(^227^)-

8.3

17.9

(^228^)-

7.8

16.6

(^229^)-

0.94

0.93

(^230^)Th series(^231^)-

8.2

17.7

(^232^)-

8.7

18.4

(^233^)-

1.06

1.04

(^234^)K-40

1.99E-01(^235^)Bq/g

8.3

17.8

1.84E-01(^236^)Bq/g

7.7

16.3

0.93

0.93

0.91

(^237^)Cs-134

1.88E+09(^238^)Bq/km2

7.8

16.8

2.00E+09(^239^)Bq/km2

8.3

17.6

1.07

1.07

1.05

(^240^)Cs-137

8.53E+09(^241^)Bq/km2

13.8

29.7

9.07E+09(^242^)Bq/km2

14.7

31.1

1.06

1.06

1.05

46.4

(^245^)-

47.2(^247^)-(^248^)-(^249^)-(^250^)-(^243^)Total(^244^)-(^246^)-

43

Table A.6

Comparison of in-situ measurement spectrum analytical results (concrete (basement
inside a building) - β: 1.4 g/cm2)(^268^)Radiation source efficiency(^269^)Simulation efficiency(^271^)(A)(^274^)Radioactivity
concentration(^270^)Ratio(^273^)(B) / (A)(^272^)(B)(^275^)
Dose rate(^276^)Proportionsof component
nuclides(^277^)Radioactivity
concentration(^278^)
Dose rate(^279^)Proportionsof component
nuclides(^280^)Radioa
ctivityconcentration(^281^)
Dose rate(^282^)Proportionsof component
nuclides(^283^)nGy /h(^284^) ％(^285^)nGy /h(^286^) ％(^287^)U series(^288^)-

19.8

25.7

(^289^)-

19.8

25.9

(^290^)-

1.00

1.01

(^291^)Th series(^292^)-

25.6

33.3

(^293^)-

27.3

35.7

(^294^)-

1.06

1.07

(^295^)K-40

7.57E-01(^296^)Bq/g

31.6

41.0

7.03E-01(^297^)Bq/g

29.3

38.4

0.93

0.93

0.94

(^298^)Cs-134(^299^)-(^300^)Bq/km2(^301^)-(^302^)-(^303^)-(^304^)Bq/km2(^305^)-(^306^)-(^307^)-(^308^)-(^309^)-(^310^)Cs-137(^311^)-(^312^)Bq/km2(^313^)-(^314^)-(^315^)-(^316^)Bq/km2(^317^)-(^318^)-(^319^)-(^320^)-(^321^)-

76.3(^326^)-(^327^)-(^328^)-(^329^)-(^322^)Total(^323^)-

76.9(^325^)-(^324^)-

In each analysis, the discrepancies in the radioactivity concentration and dose rate are approximately
within 10%. It is important that the employment of simulation efficiency is regularly accompanied by
the verification of analytical results based on radiation source efficiency.

44

Explanation B
Explanation B.1

Measurable ranges and measuring times for in-situ measurement

Detectable levels (minimum) and measuring time

Assuming the detectable level to three times the counting error, the detectable levels under thefollowing conditions are listed in Table B.1.
Cesium-137 is distributed on the ground (infinite plane)

Ge detector with 25% relative efficiency is employed

Background reading is equal to the average level in Japan (dose rate: 50 nGy/h)Table B.1

Examples of detectable levels
Detectable levels for 137Cs

Measuring time(min)

Radioactivity
concentration(kBq/m2)

Dose rate(nGy/h)

1

0.34

0.87

5

0.13

0.32

10

0.09

0.22

20

0.06

0.15

30

0.05

0.12

60

0.03

0.08

The detectable levels of cesium-137 fluctuate depending on the contributions of other radionuclides.
The values indicated here are for reference purposes only.
Note that the detectable level (radioactivity) for measuring in-situ for 60 min is almost equivalent to
the level of measuring for 10 h using a Marinelli beaker in a laboratory. The detectable level (doserate) is approximately 1/1000 of 1 mSv/year (equivalent to approximately 140 nGy/h).

45Explanation B.2

Upper limit in measurement

Measuring in places of high dose rates is expected to have the risk of underestimation of radioactivity
concentration because of increased photon injections, resulting in increased dead time and leading tocounting losses and/or pulse pile-ups. Additionally, from the viewpoint of personnel exposure toradiation, it is necessary to set an upper limit to determine whether to pursue the in-situ measurement.
Based on general MCA performance, when the number of incident photons (Input Count Rate)
increases to a certain level, the number of photons detected (Throughput Count Rate) decreases(Figure B.1), the resolution increases (Figure B.2), and the dead time also increases (Figure B.3). If thedead time grows in proportion (100 − % Live time), the error in measuring the peak area also increases(Figure B.4).
*1

Figure B.1 Relationship between Input Count Rate andFigure B.2 Relationship between Input

Throughput Count Rate*1

Count Rate and FWHM*1*2

Figure B.3 Relationship between Input Rate

Figure B.4 Relationship between % live time and peakand dead time*2

area*2 (dead time = 100 − % live time)

It is difficult to determine the feasibility of in-situ measurement by the input count rate on site; it isbetter in practice if the judgment reference is provided before conducting the in-situ measurements.
Given this, we considered setting an upper limit using the dose rate previously measured on the same――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1“Performance of Digital Signal Processors for Gamma Spectrometry,” Canberra Industries, Inc.,
Application Note (2008)*2“Comparisons of the Portable Digital Spectrometer Systems,” Duc T. Vo, Phyllis A. Russo,LA-13895-MS, Los Alamos NATIONAL LABORATORY (2002)46

site with a survey meter. Figure B.5 illustrates the relationship between the dose rate measured withthe survey meter and the in-situ Ge dead time based on readings taken in Fukushima prefecture (dateDead time (%)

of measurement: August to September 2012; taken in-situ by one Ge detector of the same model).

Dose rate measured using survey meter (μGy/h)Figure B.5 Relationship between dose rate by survey meter and in-situ Ge detector dead time
Based on the results shown in Figure B.5 and the following conditions, the relationships between thedose rates measured with a survey meter, dead time, and measuring time are organized in Table B.2.


Measuring time: 30 minGe detector with 20.6% relative efficiency is employed

Table B.2 Relationships between dose rate by survey meter, dead time, and measuring time

Dose rate measured using
survey meter

Dead time

Measuring time

(μGy/h)

(%)

(min)

1

3

31

5

14

35

10

27

41

15
20

40
53

50
64

47

According to Table B.2, the dead time is expected to reach approximately 50% if the survey-meter
dose rate is 20 μGy/h. A dead time of 50% means that the required measurement time doubles. Fromthe viewpoints of work efficiency and personnel exposure, it is not realistic to pursue a measurementin places where the dead time exceeds this level; thus, the upper limit was set at 20 μGy/h. The
processing time varies between detector models. Thus, the upper limit indicated here is only provided
as a guide.
As stated above, measurements at locations of high dose rates are likely to suffer from increased deadtime due to counting losses and pulse pile-ups, resulting in fewer photons detected (Throughput Count
Rate) with respect to the number of incident photons (Input Count Rate).
Meanwhile, the peak counts of target artificial radionuclides will increase, which makes it possiblethat a relatively short measuring time may ensure sufficiently accurate readings if they are the soletarget for measuring. Therefore, based on the comprehensive consideration of the nuclides to measure,measurement accuracy, and personnel safety, it is important to set limits to the maximum dose rate andthe measuring time. As for the relationship between the measuring time and measurement accuracy inlocations of high dose rates, it is described in Explanation B.3.

48

Explanation B.3

Relationship between measuring time and measurement accuracy in locations ofhigh dose rates
Measurements taken at locations of high dose rates are expected to have sufficient accuracy with shortmeasuring time, owing to increased numbers of peak counts for target artificial radionuclides.
We considered the following data measured in Fukushima prefecture. Figure B.6 shows an in-situ
measurement spectrum, and Figure B.7 illustrates the relative counting errors of the detected artificial
radionuclides.
Data measured in Fukushima prefecture (measured on December 27, 2011)
Real time: 4478.5 s.

Live time: 3600 s. DT: 32.7%

Dose rate: 6 μSV/h

Bi-214(1764.5keV)

K - 40(1460.8keV)
Ag-110m(1475.8keV)
Ag-110m(1505.0keV)

Cs-134(1365.2keV)
Ag-110m(1384.3keV)

Cs-134(1168.0keV)

Cs-134(1038.6keV)

Ag-110m(884.7keV)
Ac-228(911.2keV)
Ag-110m(937.5keV)

Cs-134(795.9keV)
Cs-134(802.0keV)

Cs-137(661.7keV)

Cs-134(563.2keV)
Cs-134(569.3keV)
Cs-134(604.7keV)

Cs: 8.0 × 105 Bq/m2, 137Cs: 9.3 × 105 Bq/m2, 110mAg: 3.5 × 103 Bq/m2
Cs-134(475.4keV)

134

1000000

100000

Count

10000

1000

100

10

1
0

500

Figure B.6

1000

1500

2000
Channel

2500

3000

3500

Example of spectrum acquired by in-situ measurement

49

4000

○134Cs (605 keV) (measured for 1 h)
Total count: 2023934

Net count: 1906609

Base count: 117325

Counting error: 1463.3

Relative counting error: 0.077%○137Cs (662 keV) (measured for 1 h)
Total count: 1971520

Net count: 1906912

Base count: 64608

Counting error: 1426.9

Relative counting error: 0.075%○110mAg (885 keV) (measured for 1 h)
Total count: 9098 Net count: 4714
Counting error: 116.1

Figure B.7

Base count: 4384

Relative counting error: 2.5%

Relative counting errors of detected artificial radionuclides

50

As seen in Figure B.7, the relative counting error of 137Cs measured at a location with a dose rate of 6
μGy/h for 1 h is 0.075%, which is more than sufficient in terms of statistical accuracy. Based on thismeasurement data, we evaluated the fluctuation of the measurement accuracy by varying theRelative counting error (%)

measuring time, which is illustrated in Figure B.8.

Measuring time (min)Figure B.8

Relationship between measuring time and relative counting error of 137Cs

Based on Figure B.8, in locations where the dose rate is as high as 6 μGy/h, a measurement of a few
minutes yields a reading with a relative counting error of 1% or lower, which is considered sufficient
to ensure statistical accuracy. As longer measuring time to increase the accuracy of the measurement
necessarily increases the personnel exposure, it is important to switch to short-period measurements,
in contrast to nonemergency situations.

51

Count

0
500
1000
1500
2000
Channel

52
2500
3000

Bi-214(1764.5keV)

K - 40(1460.8keV)

Bi-214(1120.3keV)

Ac-228(969.0keV)

Ac-228(911.2keV)

Cs-137(661.7keV)

Tl-208(583.2keV)
Bi-214(609.3keV)

Be-7(477.6keV)

Pb-214(351.9keV)

Pb-214(295.2keV)

Pb-212(238.6keV)

Explanation C
Example of spectrum by in-situ measurement

Spectrum measured in the vicinity of Mt. Fuji (March 7, 2001)

1000

100

10

1
3500
4000

Count

0

500

1000

1500

53

2000
Channel

2500

3000

3000
Bi-214(1764.5keV)

2500

K - 40(1460.8keV)

2000

Cs-134(1365.2keV)

Cs-134(1168.0keV)

Bi-214(1120.3keV)

Cs-134(1038.6keV)

1500

Ac-228(969.0keV)

Ac-228(911.2keV)

Cs-134(795.9keV)
Cs-134(802.0keV)

1000

Bi-212(727.3keV)

Cs-137(661.7keV)

Cs-134(563.2keV)
Cs-134(569.3keV)
Tl-208(583.2keV)
Cs-134(604.7keV)

500

Cs-134(475.4keV)

0

Pb-214(351.9keV)

Pb-212(238.6keV)

Count

Bi-214(1764.5keV)

K - 40(1460.8keV)

Bi-214(1120.3keV)

Ac-228(969.0keV)

Ac-228(911.2keV)

Bi-212(727.3keV)

Cs-137(661.7keV)

Tl-208(583.2keV)
Bi-214(609.3keV)

Be-7(477.6keV)

Pb-214(351.9keV)

Pb-214(295.2keV)

Pb-212(238.6keV)

Spectrum measured in Chiba prefecture before the Fukushima Dai-ichi Nuclear Accident (December 2,
2010)

10000

1000

100

10

1
Channel
3500

3500

Spectrum measured in Fukushima prefecture nine months after the Fukushima Dai-ichi Nuclear
4000

Spectrum measured in Chiba prefecture nine months after the Fukushima Dai-ichi Nuclear Accident(December 7, 2011)

10000

1000

100

10

1

4000

Count

0

500

1000

1500

54

2000

2500

3000

3000
Bi-214(1764.5keV)

2500

K-40(1460.8keV)

2000
ChannelCs-134(1365.2keV)

Cs-134(1168.0keV)

Bi-214(1120.3keV)

Cs-134(1038.6keV)

Ac-228(969.0keV)

1500

Ac-228(911.2keV)

1000

Cs-134(795.9keV)
Cs-134(802.0keV)

500

Cs-137(661.7keV)

0

Cs-134(563.2keV)
Cs-134(569.3keV)
Cs-134(604.7keV)

Cs-134(475.4keV)

Count

Bi-214(1764.5keV)

K - 40(1460.8keV)
Ag-110m(1475.8keV)
Ag-110m(1505.0keV)

Cs-134(1365.2keV)
Ag-110m(1384.3keV)

Cs-134(1168.0keV)

Cs-134(1038.6keV)

Ag-110m(884.7keV)
Ac-228(911.2keV)
Ag-110m(937.5keV)

Cs-134(795.9keV)
Cs-134(802.0keV)

Cs-137(661.7keV)

Cs-134(563.2keV)
Cs-134(569.3keV)
Cs-134(604.7keV)

Cs-134(475.4keV)

Accident (December 27, 2011)

1000000

100000

10000

1000

100

10

1
3500

3500

4000

Spectrum measured in Fukushima prefecture three years and eight months after the Fukushima
Dai-ichi Nuclear Accident (November 27, 2011)

1000000

100000

10000

1000

100

10

1

Channel4000• Spectrum measured in Chiba prefecture in the traversing plume caused by the Fukushima Dai-ichi

K - 40(1460.8keV)

Ac-228(911.2keV)
I-132(954.6keV)

I-132(772.6keV)

Tl-208(583.2keV)
Bi-214(609.3keV)
I-131(637.0keV)
Cs-137(661.7keV)
I-132(667.7keV)

I-132(522.7keV)
I-133(529.9keV)

Pb-214(351.9keV)
I-131(364.5keV)

Te-132(228.2keV)
Xe-133m(233.2keV)
Xe-135(249.8keV)
I-131(284.3keV)

Xe-133(81.0keV)

Nuclear Accident (March 15, 2011) (measurement taken by a fixed in-situ Ge detector installed
outdoor in an upright position)

1000000

100000

Count

10000

1000

100

10

1

0

500

1000

1500

2000

2500

3000

3500

Channel

X-ray

133

132

Xe(81 keV)

Te(228 keV)
133m

Xe(233 keV)
135

133

Xe(250 keV)

Xe(160 keV)131

I(284 keV)For measuring a plume, it is possible to calculate the radioactivity concentration and dose rate,
assuming that radionuclides are evenly distributed in the air.

55

4000

Explanation D

Energy calibration in a nuclear disasterIn a nuclear accident, a large quantity of artificial radionuclides released during the accident is
expected to increase the Compton continuum count. The radionuclides listed in Table D.1, especiallythose in the low-energy region, cannot be confirmed by their peaks; therefore, these may not besuitable for energy calibration (Figure D.1). If this is the case, it may be necessary to consider using
the peaks of the artificial radionuclides emitted through the accident, as seen in Table D.2.

Table D.1

Naturally occurring radionuclides usable for energy calibration and their gamma-ray
energies
Nuclides
212
Pb
214
Pb
208
Tl
214
Bi
228
Ac
40
K
208
Tl

Emission ratio*
0.434
0.369
0.306
0.469
0.290
0.107
0.359

Gamma-ray energy (keV)
239
352
583
609
911
1461
2615

*

For 238U and 232Th series, per-decay emission rates of the parent nuclides are shown(cited from ICRU Rep. 53, Table 3.4)

Table D.2

Artificial radionuclides usable for energy calibration in a nuclear accident and their
gamma-ray energies
Nuclides
134
Cs
137
Cs
134
Cs
110m
Ag
110m
Ag
110m
Ag

Gamma-ray energy (keV)
605
662
796
885
1384
1505

Emission ratio
0.975
0.849
0.851
0.729
0.243
0.131
(cited from ICRU Rep. 53, Table A.1)

56

Data measured in Fukushima prefecture (measured on December 27, 2011)
Real time: 4478.5 s.

Live time: 3600 s. DT: 32.7%

Dose rate: 6 μSV/h

Bi-214(1764.5keV)

K - 40(1460.8keV)
Ag-110m(1475.8keV)
Ag-110m(1505.0keV)

Cs-134(1365.2keV)
Ag-110m(1384.3keV)

Cs-134(1168.0keV)

Cs-134(1038.6keV)

Ag-110m(884.7keV)
Ac-228(911.2keV)
Ag-110m(937.5keV)

Cs-134(795.9keV)
Cs-134(802.0keV)

Cs-137(661.7keV)

Cs-134(563.2keV)
Cs-134(569.3keV)
Cs-134(604.7keV)

Cs-134(475.4keV)

Cs: 8.0 × 105 Bq/m2, 137Cs: 9.3 × 105 Bq/m2, 110mAg: 3.5 × 103 Bq/m2

1000000

100000

10000

Count

134

1000

100

10

1
0

500

1000

1500

2000
Channel

208

212

2500

3000

3500

4000

Tl

Pb
214

Pb
214

Bi

Figure D.1 Natural radionuclides not usable for energy calibration

57

228

Ac

Explanation E

Vertical profile of radioactive materials in the ground

Explanation E.1 Evaluation of vertical profile parameter βAlthough an accurate understanding of β, a parameter to indicate a vertical profile of radioactive
materials in the ground, is the most effective way to improve the reliability of in-situ measurements,
it is not easy to attain an accurate evaluation of β as it significantly fluctuates depending on the soil
quality, weather conditions, and the elapsed time. It is also necessary to consider the land erosionand human activities on the ground following the deposition. For reference, Table E.1 lists various
reports by different research groups on cesium and its β values.

58Table E.1 Evaluations for vertical profile of radioactive materials (cesium) in the ground
Time elapsed
after deposition

Fallout
category

3–4 weeks

Chernobyl

5–6 weeks

Chernobyl

Germany

0.5–1.0b<1 yChernobyl

Sweden2.2<1 y

Chernobyl

Germany

1.4<1 y

Chernobyl

Germany

0.5–4

Jacob et al., 1994a

1–3 yChernobyl

Germany

1.0–10

Jacob et al., 1994a

1–3 yChernobyl

1.4 ± 0.2

Golikov et al.,1993

4yChernobyl

1–7

Miller et al., 1991

4y

Chernobyl

1.4–5.6

IAEA, 1991a

1–5 yWeapons test

4.2

Beck, 1966

5yChernobyl

2–4.5

Jacob et al., 1994a

3–6 yChernobyl

Germany

2.5–15

Jacob et al., 1994a

3–6 yChernobyl

Ukraine

1–4

Jacob et al., 1994a

3–6yChernobyl

3.3 ± 0.7b

Golikov et al., 1993

>15 yWeapons test

Western U.S.

14 ± 4c

Beck and Krey, 1980

>15 y

Weapons test

Western U.S.

2.9 ± 1.6d

Miller and Helfer, 1985

>15 y

Weapons test

Southern U.S.

14–20

Faller, 1992

>15 y

Weapons test

Eastern U.S.

a
b
c
d

Location
Western
RussiaWestern
RussiaWestern
RussiaBelarus,
UkraineEastern U.S.
Western
RussiaWestern
Russiaβa g/cm2

Reference

0.1–2.0b

Golikov et al.,1993

2–7 (forests)
8–19 (fields)

Jacob and Meckbach,1992
Karlberg, 1987
Winkelmann etal.,1988

Miller et al., 1990

Symbol ± indicates a standard deviation.
Deposition through precipitation.
Dry area, irrigated turfs.
Dry area, uncultivated land.
(Cited from ICRU Rep.53)

59

E.1.1
Simplified evaluation method
The values given in Table E.2 (Table 5.3 in the main text) can be used according to the time elapsed
and amount of precipitation after the radioactive deposition, provided that the land is not of a special
type*1 and not disturbed by human activities. The values for β in Table E.2 take into account thecoarseness of the ground surface. Figure E.1 illustrates the chronological fluctuation of β following the
Fukushima Dai-ichi Nuclear Accident.
Using a greater value of β in radioactivity concentration analysis yields higher results, leads to asafe-side evaluation. Therefore, if uncertainties are involved in terms of the post-deposition time flow
and precipitation amounts, it is necessary to use a greater β within an assumed possible degree.
To express the radioactive vertical profile in the ground, the soil density of that location considered;thus, either the parameter β(g/cm2) or α/ρ(cm2/g) should be used. However, some units of analyticalsoftware currently available on the market use α(cm−1) as a parameter for the in-soil vertical
distribution of radioactive materials in the ground. Therefore, Table E.2 presents conversions for β, α,etc.
Table E.2

Conversion table of parameters for vertical profile of radioactive materials in the ground(The soil density (ρ) is assumed to be 1.6 g/cm3)

Time elapsed after Precipitation
deposition (years)(mm)

β(g/cm2)

RL
(cm)

α/ρ
(cm2/g)α(cm−1)

0–1

<3

0.1

0.063

10

16

0–1

3

1.0

0.63

1.0

1.6

1–5

-

3.0

1.9

0.33

0.53

5–20

-

10

6.3

0.10

0.16

4.8

3.0

0.21

0.33*2Workgroup Report

Values are cited from ICRU Rep. 53, except for a few
RL is obtained by dividing β by the soil density (ρ), α/ρ is an inverse of β, and α is an inverse of RL.

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1*2

In forested areas, β tends to be low owing to the humic soil.
“Evaluation of Public Radiation Dose in Safety Inspection Conducted at Light Water Reactor Power
Stations” by the Nuclear Safety Commission (2001)
60

Effective weight buffer depth βeff (g/cm2)

Geometric mean
(Geometric standard
deviation)

Days elapsed since the Fukushima Dai-ichi Nuclear Accident (days)

Figure E.1 Chronological fluctuation of β following the Fukushima Dai-ichi Nuclear Accident*3――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*3 Data source: Norihiro Matsuda, Japan Atomic Energy Agency

61

E.1.2
Evaluation by taking soil samples
To ensure higher accuracy of analysis based on in-situ measurement readings, it is necessary to sample
the soil at the measurement site by soil layers to evaluate the distribution of artificial radioactive
materials in the ground with respect to depth. The HASL manual suggests soil sampling at the layer
depths of 0–2.5, 2.5–5, and 5–30 cm, or 0–5, 5–10, and 10–30 cm if the radioactive materials are
considered to have penetrated the ground deeply. It has been reported*4 that the 90% penetration depthof radioactive cesium (the depth from the ground surface within which 90% of radioactive cesium
deposition amount is contained) was 4.1 cm on average in Fukushima prefecture, measured 4.5 years
after the Fukushima Dai-ichi Nuclear Accident (Figure E.2). Therefore, it is considered effective totake detailed samples near ground surface (within a depth of 10 cm) for several years following
deposition.
The distribution of radioactive materials may be heterogeneous in a horizontal direction, but in a
vertical direction it can be assumed to be reasonably homogeneous, provided that the soil quality isequally homogeneous. Therefore, it is not necessary to take many soil samples from the same location.
For surveying radioactivity distribution over a wide area, it is more efficient to select several points inthe area so that these points are not concentrated in one area, considering the environment of adjacentareas and the soil qualities, to obtain average depth-distribution evaluations. It is important to ensurethat the mitigation of personnel exposure is considered to radiation when taking soil samples following
a nuclear accident.
As for the methods of sampling, core samples may be taken with a soil auger, etc. and a scraper plate,etc. can be used to take layered samples. Taking core samples with an auger, etc. is more quickly
executed than sampling with a scraper plate, which is useful for a simplified depth distribution
analysis. For the method of sampling with an auger, see the Radioactivity Measurement Series No. 16“Methods of environmental sampling.” In this section, the method of sampling with a scraper plate is
explained, as it can adjust the sampling depths minutely and is useful for a detailed depth distribution
evaluation.
Days elapsed since the Fukushima Dai-ichi Nuclear Accident (days)

L90% penetration depth (cm)Geometric mean

December 2011 (n = 84)
August 2012 (n = 82)
August 2012 (n = 82)
July 2013 (n = 82)
November 2013 (n = 82)
August 2014 (n = 74)
November 2014 (n = 78)
August 2015 (n = 76)

Figure E.2

Relationship between the days elapsed since the Fukushima Dai-ichi Nuclear Accident
and 90% penetration depth of radioactive cesium*4――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*4“FY 2015 Commissioned Investigation of Radioactive Materials (collection of data on the distributionof radioactive materials in relation to the accident at Tokyo Electric Power Company’s Fukushima
Dai-ichi Nuclear Power Plant) Outcome Report: Research on Vertical Profile of Radioactive Cesium in
the Ground” by Norihiro Matsuda and Kimiaki Saito (2016)62E.1.2.1 Soil sampling using a scraper plate
A scraper plate is a device designed to scrape a layer of soil at any given vertical interval (minimum 5mm) (Figure E.3), comprising a metal frame to install the device on the ground and a metal plate forscraping the soil inside the frame. By attaching a metal bar on the metal plate, the depth is adjusted forsampling. Cross-contamination can be mitigated by taking the layered samples carefully.

Figure E.3 Scraper plate(1) Selection of sampling points
The sampling point is usually selected near the in-situ measurement location (see 4.1), where thetopographic conditions are very close to those of the in-situ measurement site. Sampling normallytakes several hours at a time, and for this reason it is unrealistic to sample from each in-situ
measurement location if there are many. In this case, select several major sampling points so that they
are spread evenly, in consideration of the environment of adjacent areas and the soil qualities to obtainaverage depth-distribution evaluations. It is important to ensure that the mitigation of personnel
exposure to radiation is considered when taking soil samples following a nuclear accident.
(2) Determination of sampling layers
Determine the layer from which to take samples by considering the time elapsed since thedeposition of radioactive materials. If it is immediately after the deposition has taken place, theradioactive materials are expected to be distributed close to the ground surface. Thus, samples areintensively taken from the near-surface segment at a small interval (every 5 mm). If the transfer ofradioactive materials is evident over time, deeper sampling is necessary. Similarly, if
decontamination has been evidently conducted, this must accordingly inform the decision regarding
the depth of the sample.

63

(3) Preparation of sampling tools(1) Scraper plate(2) Plastic bags to hold sampled soil (per sampled layer)(3) Vat(4) Hammer, a pair of shears, gardening shovel, spatula, forceps(5) Disposable rubber gloves(6) Pure water for cleansing measurement equipment (wet tissue papers are acceptable)(7) Tarpaulin (this can be substituted with disposable paper sheets if sampling takes place in several
locations to prevent contamination.)(8) Scale (only if soil weighing takes place on site)(9) U8 container (per sample layer; only if sample preparation takes place on site)(10) Tent or parasol for rain cover and creating a shade (if necessary)(4) Sampling procedures(1) Fix the scraper frame on the ground, using a hammer, etc. (see Figure E.4). Ensure that the frame
is directly on the soil and stable.

Figure E.4 Fixing the frame(2) If vegetation grows in the ground, trim it with shears above ground and remove it carefully (seeFigure E.5). If there are stones, remove those that are not buried in the soil. If these removed grassor stones should be measured for radioactivity concentration, keep them separately from the soil
samples.

64Figure E.5 Removing vegetation (Left: before, Right: after)(3) Attach and fix the metal bar on the plate so that the sample will be of desired depth (see Figure
E.6).

Figure E.6 Adjusting the plate(4) Slide the plate horizontally in the frame to scrape the soil, then place the soil in a plastic bag (seeFigure E.7). If the plate is held at an angle, the sample will be taken from a deeper section thanintended. Keep the plate vertical at a right angle while scraping the soil. Take as much soil as
possible for the sampled layer (see Figure E.8). Pay especially careful attention to the edges andcorners inside the frame, and use a gardening shovel or spatula to ensure all the soil is taken for the
sample. Any residual soil may contaminate the sample of subsequent layers, and may also
contribute to errors when calculating the soil density.

65Figure E.7 Soil sampling using a scraper plate

Figure E.8 After the sampling is complete(5) Once a layer of soil sample is taken, clean the plate, etc. with pure water to wash off the remaining
soil, and then wipe it dry (see Figure E.9). If pure water is not available on site, use wet tissue
paper, etc. to wipe the soil off the plate. Change the rubber gloves used when sampling to a newpair, and discard the old pair.

Figure E.9 Cleaning sampling tools66

(6) Repeat procedures (3) to (5) for the number of layers to be sampled.
(7) When all the layers are sampled, fill the hole created by sampling with the soil from thesurrounding areas, etc. (it is recommended to agree on how to refill the hole in advance).
Procedures (8) to (10) below may be performed subsequently in the laboratory.
(8) Weigh each sampled layer. Calculate the soil volume by the area inside the scraper frame and thethickness of the layer, and then obtain the density using these values.*5(9) Mix each layer of soil well in a container, such as a bag, while it is still moist.
(10) Fill the U8 container with the sampled soil per layer while it is moist, and then measure the heightand net weight. This will serve as a measurement sample for radioactivity concentration.
(5) Precautions for sampling exercises*6(1) As the sampling progresses, the internal wall of the dug hole may crumble and contaminate thelower layer samples. If this happens, remove the fallen segments of soil carefully to prevent
contamination of the lower layers by the upper-layer soil. If the hole easily crumbles, spray water
onto the walls to keep the soil in place.
(2) If there are roots in the soil, nip them with scissors carefully and include in the soil of theapplicable layer.
(3) If a stone occupies several layers of soil, include it in the layer to which the major part of thestone belongs. If it is difficult to determine, include the stone in the upper layer (for it is usually
the upper soil that has higher radioactivity concentration).
(4) If the ground is frozen, either wait until it thaws or defreeze it with a gas burner, etc. before
starting sampling.
(5) If snow is on the ground, remove the snow carefully, so as not to disturb the ground surface before
sampling.
(6) Recording(1) Sampling location(2) Sampling date(3) Weather condition at sampling(4) Land use category(5) Depth and weight of sampled soil
(6) Photograph
Overall landscape and close up of the sampling point, sampling progress, sampled ground, etc.
(7) Other remarks――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*5 Alternatively, where the sample is taken with an auger, the density can be obtained from the weight ofthe soil sampled and the capacity of the equipment.
*6 “Handbook for the Assessment of Soil Erosion and Sedimentation Using Environmental Radionuclides”

F. Zapata (2010)67E.1.2.2 Calculation of β(1) When vertical distribution of radioactive materials in the ground can be exponentiallyapproximatedAlthough the exponential model of vertical distribution of radioactive materials in the ground issimply an approximation, it suggests relatively realistic values after a certain time has elapsed since
the radioactive fallout. If the exponential approximation is possible, a β value based on themeasurement of sampled soil can be obtained by approximating the radioactivity concentration (Bq/g)against mass depth by an exponential function and obtaining β(g/cm2) from Equation (E.1). The
conversion of depth (cm) to mass depth (g/cm2) requires the soil density*7 of that location.

Z

A( Z )

A0 exp(A(Z) :

Radioactivity concentration (Bq/g) at mass depth Z

Z

Mass depth (g/cm2)

:

)

(E.1)

The depth measured from the ground surface by per-unit-areamass of the soil.
A0:Radioactivity concentration (Bq/g) at ground surfaceβ:Weight buffer depth (g/cm2)
A parameter to indicate vertical distribution of radioactive
materials in the ground. It expresses the degree of penetration,
where greater values indicate deeper penetration.
As parameters, in some cases, β divided by the soil density32
ρ(g/cm ) is RL(cm), the inverse of β is α/ρ(cm /g), and theinverse of RL is α(cm−1).

Some of the analytical software packages presently available on the market are designed to useα(cm−1) as a parameter for the vertical distribution of radioactive materials in the ground, but thisneeds to be treated carefully.*8
Example calculation of weight buffer depth β(1) Create a table such as Table E.3 based on the records and readings of the sampled soil.
Obtaining mass depth
The depth of each layer of sampled soil is between the ground surface (0 cm) to the middle of thelayer. The mass of the soil is a sum of the soil mass from the ground surface to the layer
immediately above the said layer and half of the mass of the said layer. For the area of thesampled soil, apply the area inside the auger or scraper plate used for sampling. The mass depth iscalculated by dividing the soil mass by the soil sampling area.
――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*7

It is often the case that 1.6 g/cm3 is used for soil density in general, but obtaining β after sampling soil
requires the use of the actual soil density of that location.
*8 To express the radioactive vertical profile in the ground, the soil density of that location must
considered; thus,s either the parameter β(g/cm2) or α/ρ(cm2/g) should be used. However, some of theanalytical software presently available on the market uses α(cm−1). If this is the case, it is necessary toobtain α specifically by multiplying α/ρ(cm2/g) calculated from sampled soil by the soil density
programmed in the software (e.g., 1.6 g/cm3) instead of using the actual (apparent) α.
68

Table E.3 Example of a β calculation table
Sampled layer
(cm)

Depth
(cm)

Soil weight ofsampled layer (g)Soil weight to thedepth (g)Sampling
area (cm2)Mass depth(g/cm2)

0.0–0.5
0.5–1.0
1.0–1.5
1.5–2.0
2.0–3.0
3.0–4.0
4.0–5.0
5.0–8.0

0.25
0.75
1.25
1.75
2.50
3.50
4.50
6.50

47.4
154.2
131.5
259.2
538.5
479.1
560.9
1718.2

23.7
124.5
267.4
462.7
861.6
1370.4
1890.43029.9

450
450
450
450
450
450
450
450

0.053
0.277
0.594
1.028
1.915
3.045
4.201
6.733

Cesium-137
radioactivity
concentration (Bq/g)
1.003
0.856
0.711
0.523
0.195
0.065
0.028
0.009

How to obtain mass density
Calculate the soil density of a segment from the ground surface (0 cm) to the depth by dividing themass depth by the depth in Table E.3.
The soil density of the sampled layer is obtained by dividing the soil weight by the soil volume(thickness of sampled layer × sampling area) of the layer.
(2) Calculation of βCreate a chart with the mass depth Z as the X-axis and cesium-137 radioactivity concentration A(Z) as
the Y-axis. Based on the plot yielded, apply an exponential function to approximate and obtain an
approximation formula (Figure E. 10).

69

Cesium-137 concentration A(Z) at mass depth
Z (Bq/g humidity)

A(Z) = 0.957 exp(−0.753 Z)

Mass depth Z (g/cm2)

Figure E.10

Relationship between mass depth and cesium-137 concentration(when exponential approximation is possible)

Obtain β from the coefficients in the approximation formula. In Figure E.10, Equation (E.1) gives−Z/β = −0.753 Z, and therefore, β = 1.33.

Cesium-137 concentration A(Z) at mass depth
Z (Bq/g humidity)(2) When vertical distribution of radioactive materials in the ground can be approximated by a
hyperbolic secant function
As time elapses, the migration and diffusion of radioactive materials result in distributions that involvea peak at certain depth (Figure E.11).

Mass depth Z (g/cm2)

Figure E.11
Relationship between mass depth and cesium-137 concentration(when approximation by hyperbolic secant function is possible)

70

For this type of distribution, β can be calculated by an approximation formula (E.2) based on a
hyperbolic secant function.*9 *10A( Z )

A0 cosh( Z 0 / ) sec h(ZZ0 ) /＊ sec h( x) 1 / cosh( x)、 cosh( x)
Z0

:

(E.2)

exp( x) exp( x) / 2

Mass depth (g/cm2) at which radioactivity concentration is maximum

A0, Z0, and β are calculated based on the depth profile and by the least square using an iterativealgorithm. Equation (E.2) characteristically becomes the same function for deeper areas in soil as
Equation (E.1).
In the case of a depth distribution approximated by an exponential function, there are two parameters(A0 and β) in Equation (E.1), of which the weight buffer depth β, when given, will determine the depth
profile pattern. Thus, this enables the quantification of the radioactive cesium deposition amount(Bq/cm2) from an in-situ measurement. However, if there are peaks of radioactivity concentrations at a
certain depth over the profile, Equation (E.2), which has three parameters (A0, Z0, and β), cannot
determine the depth distribution pattern with only β given. Therefore, in this case, it is practical for the
purpose of analysis to evaluate the buffer depth that is equivalent to the exponential function (hereafter,“effective weight buffer depth βeff”). Given this, we propose Equation (E.3) for calculating dose rateand Equation (E.4) for calculating deposition amount, both using the conversion factor by Saito etal.,*11 such that the relationship between the deposition amount and dose rate in the depth distribution
analyzed through Equation (E.2) is equal to that in the exponential depth distribution. Then, theeffective weight buffer depth βeff can be obtained to satisfy both equations.*10

A0,eff exp( Z /eff) I C ( Z ) dZ0eff

A0 cosh(Z 0 / ) sec h(ZZ0 ) /I C ( Z ) dZ (E.3)

0A0, eff

AZ 0 ( / 2) tan 1sinh(Z 0 / )(E.4)

A0,eff : Effective (tentative) radioactivity concentration (Bq/g) at ground surface
Iγ: Gamma-ray emission rate at radionuclide decay
C(Z) : Conversion factor proposed by Saito et al.*11
AZ0 : Radioactivity concentration (Bq/g) at the mass depth where the radioactivity
concentration is at maximum
βeff : Effective weight buffer depth (g/cm2)――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*9“Depth profiles of radioactive cesium in soil using a scraper plate over a wide area surrounding the
Fukushima Dai-ichi Nuclear Power Plant” by N. Matsuda, S. Mikami, S. Shimoura, J. Takahashi, M.
Nakano, K. Shimada, K. Uno, S. Hagiwara and K. Saito, Journal of Environmental Radioactivity 139,
427-434 (2015)*10 “FY 2015 Commissioned Investigation of Radioactive Materials (collection of data on the distributionof radioactive materials in relation to the accident at Tokyo Electric Power Company’s Fukushima
Dai-ichi Nuclear Power Plant) Outcome Report: Research on Vertical Profile of Radioactive Cesium in
the Ground” by Norihiro Matsuda and Kimiaki Saito (2016)*11 K. Saito, P. Jacob, Fundamental data on environmental gamma-ray fields in the air due to source in the
ground, JAERI-Data/Code 98-001 (1998)
71

In the case where the radioactivity against mass depth decreases exponentially as in Equation (E.1),
the right side of Equation (E.3) becomes identical with the left side. Therefore, the weight buffer depth
obtained by Equation (E.1) is the effective weight buffer depth as is.
By applying effective weight buffer depth βeff in the exponential function model, the readings ofground surface measurements can be associated with the radiation source in the ground. Similarly, thecoarseness of the ground surface can also be treated with the effective weight buffer depth βeff.
Obtaining the average of β calculated from soil samples of different sampling spots can be done either
by using the arithmetic mean or the geometric mean. Preferably, the average β is assumed to be thearithmetic mean when the created frequency distribution is similar to a normal distribution, and it isassumed to be the geometric mean when it is more similar to a log-normal distribution. However, βvalues taken from the locations where human activities such as decontamination has disturbed thenatural vertical distribution must be excluded from the average calculation.

72

Ratio of fluence rate

E.1.3 Use of gamma rays of different energies emitted from the same nuclide
If a nuclide emits gamma rays of different energies, the ratio of their fluence rates can be used toestimate β. The ratio of the fluence rates between the cesium-137 gamma ray of 662 keV and X ray of
32 keV (32 keV/662 keV) fluctuates by β, which is illustrated in Figure E.12. The difference in the
decay of the 32 keV (X ray) and 662 keV (gamma ray) in soil is evident. This is used to estimate β. To
apply this method, it is necessary to use an n-type Ge detector so that the low-energy area can bemeasured to detect 32 keV.
This method is also effective as it reduces the influence resulting from the coarseness of the ground.
However, note that this method is not useful during the first month following a nuclear accident, asradiation from nuclides with short half-lives interferes with reading the energy region around 32 keV.
It is equally difficult to apply this method when a long time has passed after an accident, for the X rayof 32 keV becomes difficult to detect as cesium penetrates deep into the ground.
The method may be applied to lanthanum-140 or cesium-134, but it is not as effective as when it isapplied to cesium-137.

β (g/cm2)
Figure E.12 Relationship between fluence rate ratio of gamma rays of different energies
emitted by the same nuclide and vertical profile (β) of radioactive materials in
the ground(Cited from ICRU Rep.53)

73

Explanation E.2 Influence on analytical results
Figure E.13 illustrates the results from analyzing a spectrum taken at a certain location to obtain thedose rate and radioactivity concentration (per-unit-area radioactivity: Bq/cm2) by altering theparameter β for the vertical profile of radioactive materials in the ground. Although the dose rate does
not significantly depend on β, the radioactivity concentration is considerably dependent on β.
Therefore, when calculating radioactivity concentration, understanding β correctly is most crucial for
ensuring the credibility of the measurement. Currently, β is assumed to be 4.8 g/cm2 (α: 0.33 cm−1)*12
for evaluating cesium-137 derived from atmospheric nuclear tests, but this value cannot be used forevaluating the post-accident ground surface distribution of radioactivity concentration, for it would
overestimate the results by approximately three times.
Similarly, in the case where gamma rays of different energies are emitted by the same nuclide, if
discrepancies are evident in the analytical results of radioactivity concentration between energies, it ispossible that the assumptions of the vertical profile of radioactive materials in the ground, etc. areerroneous.

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*12“Evaluation of Public Radiation Dose in Safety Inspection Conducted at Light Water Reactor Power
Stations” by the Nuclear Safety Commission (2001)
74

Relative value

Radioactivity concentration (137Cs)
analytical results

Dose rate (137Cs) analytical results

Figure E.13

Effects on the analytical results of vertical profile parameter βThis shows the results from analyzing cesium-137 (137Cs) in agiven spectrum from a measurement with β varied from 0 to 20,
expressed as relative values by comparing with the groundsurface distribution (β = 0).
The analytical results for radioactivity concentration increase aslarger values are given for β. A greater value of β means that theradioactive material penetrates deep in the soil. Thus, dividing
the measured peak count rate (constant in this case) by adiminished efficiency (Nf /Aa) yields a higher radioactivity
concentration.
The dose rate is not considerably affected by β because theefficiency for the dose rate (Nf /I) does not change significantly,
even when the radioactive materials penetrate deep in the soil.

75

Explanation F Implications of discrepancies between analytical and actual
measurement conditions
Explanation F.1

Breadth of surrounding areas

The coefficients required for the calculation of radioactivity concentration from an in-situ
measurement (φ/A in Equation 5.4 and values in Table-1 and Table 5.4) assume an infinitely open
and flat landscape (infinite plane) with no obstacles that attenuate gamma rays, such as buildings
and vehicles. However, in reality, it is impossible to expect such a perfect infinite plane; therefore,
an analysis of radioactivity concentration based on this assumption of infinite plane will result inunderestimation.
The impact of the breadths of surrounding areas on measurement readings is illustrated in Figures
F.1 to F.6, by different vertical distributions (β) of radioactive materials in the ground. The values
given in the charts indicate the efficiencies*1 *2 (peak count rate/radioactivity concentration)
simulated by varying the breadth of the surrounding areas, which are expressed in relative values
with respect to the values for a flat area of 150 m radius (considered as an infinite plane). These
charts serve as a reference point to evaluate the extent of underestimation.
Given the acceptable margin of underestimation as −20%, a measurement of cesium-137(approximately 600 keV) distributed on and near the ground surface (β = 0.1 g/cm2, Figure F.1)
requires a breadth of at least 25 m radius. If β = 4.8 g/cm2 (Figure F.5), the radius becomes 10 m,with a radius of 5 m for an in-soil homogeneous distribution.
If it is difficult to secure a sufficiently open area, correction can be applied to the measurement
readings using Figures F.1 to F.6. However, to minimize the uncertainties due to correction, it isdesirable to secure at least a 10-m radius of open area if the distribution is found on and near the
ground surface.
These results are based on a simulation*1 *2 using a p-type detector, which is a Ge detector withrelatively less directional dependence. Figures F.7 and F.8 illustrate the results of calculating anear-surface distribution (β = 0.1 g/cm2) using a general p-type detector and an n-type detector,
which can measure low energies. There are no major differences between the detectors in terms of
the impact of the area breadths on their readings. As for the directional dependence of thesedetectors, see Explanation G.
Figures F.9 to F.11 show the extent of gamma ray contributions to in-situ measurements withrespect to distance from the detector. Note that Figure 9 is based on the vertical distribution (β) ofradioactive materials in the ground and Figures F.10 and F.11 are based on the difference ingamma-ray energies.

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1

For the simulation, peak efficiency simulation software (calculation code: MCNP Monte Carlo code*2)
was used.
*2
Briesmeister, J.F., “MCNP-A general Monte Carlo N particle Transport Code Version 4C”, Los Alamos
National Laboratory Report LA-13709-M (2000)

76

−1

2

(s-1 /(γ/cm/(γ/cm ))2))ｓ(cps
Nf /ANf/A(Relative value against infinite(無限平面の値に対する相対値)
plane value)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

60

Breadth of surrounding
周囲の広がり(半径(m))
areas (radius in m)
図B.1
周囲の広がりによる計測値への影響
Figure F.1
Impact of breadth of surrounding areason2（検出器：p型
β: (detector:
0.1 g/cm) β:measurement readings
p-type,0.1 g/cm2)−1

2(s-1 /(γ/cm/(γ/cm ))2))ｓ(cps
Nf /ANf/A
value against infinite(Relative(無限平面に対する相対値)
plane)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

Breadth of surrounding
周囲の広がり(半径(m))
areas (radius in m)
図B.2
周囲の広がりによる計測値への影響
Figure F.2
Impact of breadth of surrounding areason2（検出器：p型
β: (detector:
1.0g/cmp-type,) β:
measurement readings
1.0 g/cm2)

77

60

−1

2

(s-1 /(γ/cm/(γ/cm ))2))ｓ(cps
Nf /ANf/A
value against infinite(Relative(無限平面に対する相対値)
plane)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

60

Breadthof surrounding
周囲の広がり(半径(m))
areas (radius in m)
図B.3
周囲の広がりによる計測値への影響
Figure F.3
Impact of breadth of surrounding areason2（検出器：p型
β: (detector:
3.0g/cmp-type,)measurement readingsβ:
3.0 g/cm2)(s-1−1/(γ/cm/(γ/cm2))2))ｓ(cps
Nf /ANf/A
value against infinite(Relative(無限平面に対する相対値)
plane)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

Breadthof surrounding
周囲の広がり(半径(m))
areas (radius in m)
図B.4
FigureF.4 周囲の広がりによる計測値への影響
Impact of breadth of surrounding areas on2（検出器：p型
β: (detector:
10.0g/cm) β:measurement readings
p-type,10 g/cm2)

78

60

2

1.0

0.8

−1

(s-1 /(γ/cm/(γ/cm ))2))(cps
Nf /ANf/Aｓvalue against infinite(Relative(無限平面に対する相対値)
plane)

1.2

0.6

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.4

0.2

0.0
0

10

20

30

40

50

60

Breadthof surrounding
周囲の広がり(半径(m))
areas (radius in m)
図B.5
Figure F.5周囲の広がりによる計測値への影響
Impact of breadth of surrounding areason2（検出器：p型
β: 4.85g/cm)
measurement readings(detector: p-type,β:
4.8 g/cm2)−1

2(s -1/(γ/cm/(γ/cm))2))(cps
Nf /ANf/Aｓ(cps/(γ/g))
value against infinite(Relative(無限平面に対する相対値)
plane)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

Breadthof surrounding
周囲の広がり(半径(m))
areas (radius in m)
図B.6F.6 周囲の広がりによる計測値への影響
FigureImpact of breadth of surrounding areas on（検出器：p型
均一分布)
measurement readings(detector: p-type,homogeneous distribution)

79

60

−1

2

(s /(γ/cm/(γ/cm ))2))(cps
Nf /A Nf/A
value against infinite(Relative(無限平面に対する相対値)
plane)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

60

Breadth of surrounding areas
周囲の広がり(半径(m))(radius in m)2
図B.7(β:on0.1g/cm） readings
Figure 周囲の広がりによる計測値への影響
F.7 Impact of breadth of surrounding areas
measurement2(β: 0.1（検出器：p型、L/D；0.9、不感層；1mm、相対効率；40％)g/cm ) (detector: p-type, L/D; 0.9, dead layer; 1 mm, relative efficiency;
40%)

−1

2

(s-1/(γ/cm/(γ/cm ))2))ｓ(cps
Nf /ANf/A
value against infinite(Relative(無限平面に対する相対値)
plane)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

Breadthof surrounding
周囲の広がり(半径(m))
areas (radius in m)

50

60

図B.8 周囲の広がりによる計測値への影響 (β: 0.1g/cm 2）
FigureF.8 Impact of breadth of surrounding areas on measurement readings (β:（検出器：n型、L/D；1.0、不感層；0.1μm、相対効率；25％)
0.1 g/cm2) (detector: n-type, L/D; 1.0, dead layer; 0.1 μm, relative efficiency;
25%)

80

of contribution
Ratio寄与割合

100%
90%
80%
70%
60%
50%
40%
30%
20%
10%
0%

β： 0.10 g/cm2
β： 4.85 g/cm2
Homogeneous
均一

0

5

図B.9

10
15
20
Distancefrom the detector (m)検出器からの距離（ｍ）

25

30

in-situ測定における周囲からの寄与割合 (600keV)

Figure F.9 Ratio of contribution from surrounding areas in in-situ measurement(gamma-ray energy: 600 keV)of contribution
Ratio
寄与割合

100%
90%
80%
70%
60%
50%
40%
30%
20%
10%
0%

50keV
600keV
2500keV

0

5

10
15
20
Distancefrom the detector (m)検出器からの距離（ｍ）

25

30

図B.11
Figure F.10
Ratioin-situ測定における周囲からの寄与割合
of contribution from surrounding (β:areas0.1g/cm2)
in in-situ measurement

寄与割合
of contribution
Ratio(β: 0.1 g/cm2)
100%
90%
80%
70%
60%
50%
40%
30%
20%
10%
0%

50keV
600keV
2500keV

0

5

10
15
20
Distancefrom the detector (m)検出器からの距離（ｍ）

25

図B.11 in-situ測定における周囲からの寄与割合 (β: 4.85 g/cm2)

Figure F.11

30

Ratio of contribution from surrounding areas in in-situ measurement(β: 4.8 g/cm2)

81

Explanation F.2

Height of detector installation

The analysis for in-situ measurements assumes that the detector is installed at 1 m
above the ground to take the measurement. Normally, there is no need for correction,
as the detector is installed at 1 m above the ground. However, it is possible that,
owing to topographical difficulties, etc., the detector may be installed at a different
height. Therefore, the manner in which the height of the detector installation affects
the measurement readings is described below.
Figures F.12 to F.14 illustrate the results of simulations*3 *4 of in-situ measurement
efficiencies (peak count rate/radioactivity concentration) at varied installation heights
with respect to different vertical distributions (β) of radioactive materials in the
ground. The values shown on the charts are relative to those of a 1-m installation.
The higher the installation, the lower the efficiency. Therefore, if the detector isinstalled higher than 1 m, or if the surrounding areas are lower because, for example,
there is a cliff, caution must be taken for possible underestimation.
A particular caution is required when radioactive materials are distributed near the
ground surface (Figure F.12, β = 0.1 g/cm2), as the impact of installation height isgreater.

2

1.0

0.8

−1

(s /(γ/cm))2))/(γ/cm(cps
Nf/ANf/A
value against 1-m(Relative(1mに対する相対値）
installation)

1.2

0.6

50keV
100keV
300keV
600keV
1000keV
2500keV

0.4

0.2

0.0
0

2

4

6

8

10

12

Detector installation height
検出器設置高さ(m) (m)2 2)
FigureF.12 Impact
of detector installation height on measurement readings(β: 0.1 g/cm
図B.12
検出器設置高さによる計測値への影響（β：0.1g/cm)――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*3

For the simulation, peak efficiency simulation software (calculation code: MCNP Monte Carlo code*4)
was used.
*4
Briesmeister, J.F., “MCNP-A general Monte Carlo N particle Transport Code Version 4C”, Los Alamos
National Laboratory Report LA-13709-M (2000)

82

2

1.0

0.8

−1

/A (s /(γ/cm/(γ/cm ))2))
Nf/ANf(cpsvalue against 1-m(Relative(1mに対する相対値）
installation)

1.2

0.6

50keV
100keV
300keV
600keV
1000keV
2500keV

0.4

0.2

0.0
0

2

4

6

8

10

12

Detector
installation(m)height (m)検出器設置高さ

図B.13 検出器設置高さによる計測値への影響 （β：4.85g/cm 2)
Figure F.13 Impact of detector installation height on measurement readings(β: 4.8 g/cm2)

2

1.0

0.8

−1

(s /(γ/cm
Nf/A(cps))/(γ/g))
Nf/A
value against 1-m(Relative(1mに対する相対値）
installation)

1.2

0.6

50keV
100keV
300keV
600keV
1000keV
2500keV

0.4

0.2

0.0
0

2

4

6

8

10

12

Detector installation height
検出器設置高さ(m) (m)

図B.14 検出器設置高さによる計測値への影響 （均一分布）
Figure F.14 Impact of detector installation height on measurement
readings (homogeneous distribution)

83

Explanation F.3

Moisture in the ground

Moisture in the ground does not need correction, except in certain special cases, as it
is included in the effects of vertical distribution of radioactive materials in the ground(Explanation E). However, in the case of a longitudinal measurement at the same
point, it is conceivable that comparisons are made between in-situ measurements
taken at different times, which may involve measurements taken under fine weather
and others immediately after rain. Therefore, the manner in which the moisture in
the ground can affect measurement readings is described below.
Figures F.15 to F.18 illustrate the results of simulations*5 *6 of in-situ measurement
efficiencies (peak count rate/radioactivity concentration) at varied moisture levels in
the ground with respect to different vertical distributions (β) of radioactive materials
in the ground. The values shown on the charts are relative to those with a moisturelevel of 10%.
Naturally, the impact of moisture in the ground is small if radioactive materials are
distributed near the ground surface (β = 0.1 g/cm2) (Figure F.15). If radioactive
materials penetrate the ground to some extent (β = 4.8 g/cm2), a greater amount ofmoisture in the ground will lower the efficiency more, owing to its shielding effect(Figure F.16). This means that the measurement reading will yield lower values.
Immediately after the deposition of radioactive materials, the impact of moisture in
the ground is negligible, but after some time has passed and radioactive materials
have penetrated the ground deeply, caution must be taken as the measurement
readings will be affected by the moisture in the ground at the time of themeasurement. Therefore, it is recommended that moisture levels in the ground at thetime of measurement be recorded if possible.
In the analysis of radioactive nuclides distributed homogeneously, there is hardly aneffect of moisture in the ground (Figure F.17). This suggests that the efficiency(s−1/(Bq/g moist soil)) does not fluctuate significantly as the reduced count rate due tothe shielding by water balances out the effect of diluting the radioactivity
concentration.
However, the efficiency for radioactivity per dry soil (s−1/(Bq/g dry soil)) is not affectedby dilution by water, whereas moisture shields radioactivity, which reduces thecounting rate. Therefore, the efficiency is affected by the in-soil moisture (Figure F.18).
The value obtained from the in-situ measurement is the radioactivity per the actual
soil (moist). Normally, the in-soil radioactivity level measured in a laboratory is per
dry-soil measurement; thus, caution must be taken when comparing such values.

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*5

For the simulation, peak efficiency simulation software (calculation code: MCNP Monte Carlo code*6)
was used.
*6
Briesmeister, J.F., “MCNP-A general Monte Carlo N particle Transport Code Version 4C”, Los Alamos
National Laboratory Report LA-13709-M (2000)

84

−1

2

(s /(γ/cm/(γ/cm ))2))(cps
Nf /ANf/A
value against 10%(Relative(10%の値に対する相対値)
moisture)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

Moisture in the ground (%)
土壌中水分（％）

図B.15 土壌中水分による計測値への影響
2*2
Figure F.15 Impact of moisture in the ground on measurement(when
β: 0.1g/cmreadings（放射性物質が地表近くに分布している場合、RL:0.0625cm)
radioactive materials are distributed near the ground surface, β: 0.1
g/cm2)

1.2−12(s /(γ/cm))2))/(γ/cm(cps
Nf /ANf/A
value against 10%(Relative(10%の値に対する相対値)
moisture)

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

Moisture in the ground (%)
土壌中水分（％）

図B.16 土壌中水分による計測値への影響
Impact of moisture in the ground on measurement
readings (when*2
β: 4.85g/cm2（放射性物質が土壌中に浸透した場合、RL:3.03cm)
2

Figure F.16

radioactive materials have penetrated into the ground, β: 4.8 g/cm )

85

1.2

−1
moist soil))/(γ/g/(γ/g湿土))(cps/Af/A (sNf Nagainst 10%
value(Relative(10%の値に対する相対値)
moisture)

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20
30
40
Moisture in the ground (%)
土壌中水分（％）

50

図B.17 土壌中水分による計測値への影響
Impact of moisture in the ground on measurement（湿土あたりの放射能を計測する場合,
readings (when measuring radioactivity均一分布)
per moist soil,

Figure F.17

homogeneous distribution)−1
moist soil))
Nf/A(s/(γ/g乾土))(cps/(γ/g
Nf /A(Relative value against 10%(10%の値に対する相対値)
moisture)

1.2

1.0

0.8

50 keV
100 keV
300 keV
600 keV1000 keV
2500 keV

0.6

0.4

0.2

0.0
0

10

20

30

40

50

土壌中水分（％）
Moisture
in the ground (%)
図B.18

土壌中水分による計測値への影響

Figure F.18
Impact of soil moisture on measurement readings(when（乾土あたりの放射能を計測する場合,
均一分布)
measuring radioactivity per dry soil, homogeneous
distribution)

86

Explanation G Directional characteristics of detector (angular dependence of peakefficiency)
Examples of directional characteristics in terms of the p-type and n-type Ge detectors are shown in
Figures G.1 and G.2, respectively. In in-situ measurements, the detector is normally installed facing
downward. Given the right angle at the ground is zero degrees, the relative efficiencies are described
below between the efficiency of a given degree against that of 0°. Note that these figures are
simulation*1 *2 results, assuming the distance between the radiation source and the detector to be 1 m.
For the p-type detector, the parameters are as follows: relative efficiency 40%, diameter 61 mm, length
56 mm, and dead layer 1 mm. For the n-type: relative efficiency 25%, diameter 53 mm, length 53 mm,
and dead layer 0.1 μm.
The p-type detector shows the tendency to yield low values in the low-energy region (50–60 keV) for
gamma rays incident at an angle. This is considered to be because the angled incidence results in alonger passage through the dead layer, and therefore a greater attenuation. There is almost no
directional dependence recognized at energies of 100 keV or more.
As to the n-type detector, the dead layer is extremely thin and angled incidence does not cause thelowering of peak efficiency in the low-energy region. On the contrary, there was a tendency that higher
values were yielded. This is considered to be because the peak efficiency in the low-energy region
depends more on the cross-sectional area than the volume. Overall, no significant directional
dependence was recognized.
The detectors treated in this example those with the length–diameter ratio (L/D) of approximately 1.0.
Detectors that fall outside the L/D range of 0.9 to 1.1, and special-purpose detectors for measuring low
energy background, may have strong directional dependence. Therefore, it is necessary to understandthe directional characteristics when using such detectors.

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1

For the simulation, peak efficiency simulation software (calculation code: MCNP Monte Carlo code) wasused.
*2
Briesmeister, J.F., “MCNP-A general Monte Carlo N particle Transport Code Version 4C”, Los Alamos
National Laboratory Report LA-13709-M (2000)

87

Relative sensitivity

Angle
Figure G.1

Example of angular dependence of peak efficiency

Relative sensitivity(Detector: p-type, L/D: 0.9, dead layer: 1 mm, relative efficiency: 40%)

Figure G.2

Angle
Example of angular dependence of peak efficiency(Detector: n-type, L/D: 1.0, dead layer: 0.1 μm, relative efficiency: 25%)88

Explanation H
Explanation H.1

Example of actual measurement

Artificial radioactive materials deposited on the ground

1. Purpose
To verify the validity of in-situ measurement for artificial radioactive materials deposited on
the ground.
2. Methods considered
Both in-situ measurement and measurement of a gamma-ray spectrum for soil samples wereconducted at the actual environmental site, and the results were compared. The measurement
location was chosen near Mt. Fuji because it had a low concentration of naturally occurring
radioactive materials, and it was possible to find unspoiled and uncultivated lands.
(1)

In-situ measurement
The measurement was conducted at an open area of 10 m or more, devoid of trees. The Ge
detector was installed at a height of 1 m, and the measurement was taken for 1 h. The analysis
was performed using the HASL method to calculate the in-soil cesium-137 radioactivity
concentration and gamma-ray dose rate at the location. Table H.1 lists the specifications of
the detector used.

(2)

Soil sampling and measuring
Soil samples were taken near the measurement point, using an auger of 5 cm diameter, andthe 30-cm depth was divided into six layers, each of thickness 5 cm. The soil samples were
not allowed to dry, and they were placed in measurement containers (U-8) after large stones,
etc. were removed. In the laboratory, they were then measured for approximately 70000 s
using a Ge detector, and the gamma-ray spectrum obtained was used to calculate thecesium-137 radioactivity concentration.

Table H.1

Specifications of the Ge detector used in the in-situ measurement

Relative
CrystalCrystal length
Detector efficiency diameter (D)(L/D)(L) (mm)(%)(mm)

End cap(mm)

Dead layer

Distance between
end cap andcrystal (mm)

A

25

53

53

1.00

0.5 (Al)0.1 μm

5.0

B

25

54.6

54.7

1.00

1.27(Al)

0.7 mm

3.0

89

3. Results and discussions
Table H.2 lists the cesium-137 radioactivity concentrations based on the in-situ measurement
and sampling.
Calculating the radioactivity concentration from in-situ readings requires vertical distribution ofradioactive materials in the ground; however, in this case, the analysis was conducted using
both the general assumption of β = 4.8 g/cm2 and the actual values taken from themeasurement.

Table H.2 radioactivity concentration measured at Mt. Fuji

Measurement
location

Parameter of vertical
distribution used in in-situ
analysis

Point A
(Hitoana,
Fujinomiya City)

Point B
(Fuji Skyline
Drive)

137

Cs radioactivity concentration (kBq/m2)

β (g/cm )in-situ(Detector A)

in-situ(Detector B)

4.8

1.6

1.6

2

Remarks
Sampled soil

1.7
8.15

1)

4.8

-

2.1

2

2

Humic soil

4.0
15.4

2)

-

4.2

1)2) Values taken from the sampled soil

Concerning the analytical results using β = 4.8 g/cm2, the in-situ and sampling measurements
yielded almost identical results at Point A, but at Point B there was a difference of almost double
the value. A possible explanation for this is that the assumed vertical radioactive profile in the
ground for analysis was discrepant from the actual profile.
We calculated β based on the vertical profiles of cesium-137 in the ground taken from eachsampling point (see Figures H.1 and H.2) and the soil density of these points.*1 The analysis of
the in-situ reading using the β (Tables H.2 1) 2)) yielded a value close to the 137Cs radioactivity ofthe sampled soil.
Figures H.3 and H.4 illustrate the in-situ spectra measured at the respective locations.

――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1 The calculation was done by adapting the soil sampling depth Z(cm) and 137Cs radioactivity

concentration A(Z) at depth Z (Bq/kg moist soil) to the exponential function A(Z) = exp(−αZ) to obtainα(cm−1), and then obtain β by β = ρ/α (ρ: density of sampled soil).

90β = ρ/αβ = 8.15 g/cm2
Soil density ρ = 1.06 g/cm3

137

Cs concentration A(Z) at depth Z(Bq/kg moist soil)

Point A
A(Z) = A0/exp(−αZ)

Depth Z (cm)Point B
A(Z) = A0/exp(−αZ)β = ρ/αβ = 15.4 g/cm2
Soil density ρ = 1.22 g/cm3

137

Cs concentration A(Z) at depth Z(Bq/kg moist soil)

Figure H.1 Relationship between in-soil cesium-137
concentration and depth(Point A)

Depth Z (cm)Figure H.2 Relationship between in-soil cesium-137
concentration and depth(Point B)91

Cs-137←

Count/channel

Channel(Measuring time: 3600 s, 0.5 keV/channel)
Figure H.3

In-situ measurement spectrum using Ge detector←

Count/channel

Cs-137(Point A)

Channel(Measuring time: 3600 s, 0.5 keV/channel)
Figure H.4

In-situ measurement spectrum using Ge detector(Point B)

92

Explanation H.2

Naturally occurring radioactive materials

1. Purpose
To verify the validity of in-situ measurement for naturally occurring radioactive materials in the
ground.
2. Methods considered
Both an in-situ measurement and measurement of a gamma-ray spectrum for a soil sample wereconducted at the actual environmental site, and the results were compared.
(1)

in-situ measurement
A location with an open area of at least 10 m around was selected. The Ge detector was
installed at a height of 1 m from the ground, and the measurement was taken for 1 h. The
resulting spectrum was analyzed using the HASL method to calculate the radioactivity
concentration of naturally occurring radioactive materials in the ground.

(2)

Soil sampling and measurement
Samples were taken near the measurement location using an auger. The soil samples were notallowed to dry, and they were placed in measurement containers (U-8) after large stones, etc.
were removed. Then, the samples were measured in the laboratory for approximately 70000 s
using a Ge detector, and the gamma-ray spectrum obtained was used to calculate radioactivity
concentration.

3. Results and discussions
Table H.3 summarizes the results of the in-situ measurement and measured results of the soil
sampled at every measurement location. The in-situ measurement and sampled soil
measurement yielded very similar readings of naturally occurring radioactive nuclides. Notethat the analysis of in-situ measurement was based on the assumption that the radioactive
materials were distributed homogeneously in the ground.
Table H.5 indicates the relationship between the results of the in-situ measurement andmeasured results of the soil sampled at every measurement location. The results of the in-situ
measurement and sampled soil measurement were well correlated.
Figure H.6 shows a view of the in-situ measurement, and Figure H.7 illustrates an example of ameasured spectrum.

93

Table H.3 Comparison between in-situ measurement results and gamma-ray spectrometry of sampled soil
Unit: Bq/kg moist soil
NuclidesU series
214

Sampling/measurement point

Point 1
Composite sample
Core sample
0~ 5cm5~10cm10~15cm
15~20cm
20~25cm
25~30cmin-situ
Point 2Composite sample
Core sample
0~ 5cm5~10cm10~15cm
15~20cm
20~25cm
25~30cmin-situ
Point 3Composite sample
in-situPb (352.0keV)

214

Th series

Bi (609.3keV)

208

Tl (583.1keV)

212

K-40

Pb (238.6keV)

228

Ac (911.1keV)

40

K (1460.8keV)

25 ±

1.0

23 ±

1.0

7.2 ±

0.47

24 ±

0.7

17 ±

1.6

380 ±

10

25 ±
27 ±
28 ±
27 ±
24 ±
27 ±
19

0.8
0.8
1.2
0.8
1.2
1.1

22 ±
23 ±
25 ±
23 ±
20 ±
24 ±
21

0.8
0.7
1.3
0.9
1.1
1.2

7.4 ±
6.8 ±
7.9 ±
7.6 ±
6.1 ±
7.0 ±
7.6

0.32
0.29
0.58
0.34
0.44
0.53

23 ±
23 ±
25 ±
22 ±
24 ±
23 ±
17

0.6
0.6
0.9
0.6
0.9
0.8

19 ±
20 ±
20 ±
20 ±
21 ±
23 ±
22

1.1
0.9
2.0
1.2
1.4
2.0

370 ±
350 ±
360 ±
370 ±
320 ±
360 ±
430

8
6
12
8
8
11

19 ±

0.6

16 ±

0.5

7.5 ±

0.26

26 ±

0.5

22 ±

0.8

470 ±

6

15 ±
21 ±
20 ±
21 ±
23 ±
24 ±
16

0.9
0.6
0.6
1.0
0.7
0.8

16 ±
17 ±
16 ±
20 ±
20 ±
21 ±
18

1.0
0.6
0.6
1.0
0.7
0.7

6.4 ±
9.0 ±
7.6 ±
9.1 ±
9.8 ±
11 ±
9.3

0.48
0.31
0.26
0.50
0.31
0.3

24 ±
28 ±
26 ±
31 ±
29 ±
37 ±
23

0.7
0.5
0.5
0.8
0.5
0.7

19 ±
25 ±
24 ±
31 ±
25 ±
32 ±
24

1.7
1.0
0.8
1.8
1.1
1.1

410 ±
470 ±
470 ±
510 ±
500 ±
530 ±
490

11
7
6
12
8
8

20 ±
17

0.7

20 ±
18

0.7

9.4 ±
11

0.33

29 ±
26

0.5

28 ±
29

1.1

500 ±
530

8

Composite sample: Mixed soil sampled from four corners of a 2-m square centering on the measurement point
Core sample:
Depth-layered soil sampled just under the measurement point
The in-situ measurement results were analyzed assuming homogeneous distribution in the ground.

94

1000
採取した土壌の測定結果
Sampled soil measurement reading（Bq/kg湿土）(Bq/kg moist soil)

Useries
U系列
Th
series
Th系列

K-40
100

10

1
1

10
100
In-situmeasurement reading(Bq/kg moist soil)
in-situ測定結果（Bq/kg湿土）

1000

Figure H.5 Comparison between in-situ measured results and gamma-ray spectrometry of soil
sampled at each location

95

K-40
←

Ac-228Tl-208Bi-214←←←

Pb-214←←

Count/channel

Pb-212

Figure H.6 Performing an in-situ measurement

Channel(Measuring time: 3600 s, 0.5 keV/channel)
Figure H.7

Example of in-situ measurement spectrum obtained by Ge detector

96Explanation I

Intercomparison measurement

The intercomparison of measurement equipment is an effective method to improve the reliability of
in-situ measurement readings. Urgent in-situ measurement in a nuclear disaster is highly anticipated,
and it is important to ascertain the equipment conditions regularly and routinely. If multiple pieces ofmeasuring equipment are employed in surveying a wide area for the radioactivity distribution, it isadvantageous to verify in advance that they are all equal in terms of measurement accuracy.

(1)

Selection of location

See “4.1 Selection of the measurement site.”(2)

Verification of surrounding areas

Verify the areas near the location for the intercomparison measurement (ideally, an area of radiusapproximately 30 m), using a survey meter, etc., to ensure that there is a little difference in dose rate
distribution in the areas.

(3)

Measurement methods

The main measurement methods are the following two:

(1) Simultaneous measurement using several detectors in a possibly small area (for example, on
concentric circles), arranged such that they do not disrupt the measurements of one another(Figures I.1 and I.2)
(2) Sequential measurement on the same point with alternating detectors

Owing to the possibility that circumstances may change with time at the measurement point, method(2) requires prompt execution of the measurement procedures.
To secure the validity of in-situ measurement readings, it is also recommended that the dose rate bemeasured at the same point simultaneously, using a survey meter.

(4)

Evaluation of results

Compare the artificial and natural radionuclides detected separately. If correction is necessary owing
to significant discrepancies between measurement points or changes in the environment, use the dose
rates measured using a survey meter, which may help to mitigate the impact of such discrepancies on
the comparisons (Figure I.3).

97

Figure I.1

Figure I.2

Performing the intercomparison measurement (simultaneous measurement)*1

Layout example of intercomparison measurement (simultaneous measurement)*1(The smallest of the three concentric circles indicates the measurement points for dose rate using asurvey meter. The figures indicated on the side are the dose rates (μSv/h) at 1 m above the ground interms of the mean value and standard deviation (n = 5). Measurements were taken simultaneously with
the Ge detectors installed at Points A to F for the intercomparison measurement.)――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――――*1 “In Situ Gamma Spectrometry Intercomparison in Fukushima, Japan” S. Mikami, S. Sato, Y.

Hoshide, R. Sakamoto, N. Okuda, K. Saito, Jpn. J. Health Phys., 50 (3), 182–188 (2015)98

Figure I.3

Example of intercomparison measurement reading for cesium-137*1(The vertical axis represents the dose rate (μSv/h) measured using a survey meter and cesium-137
deposition amount (Bq/m2) measured using a Ge detector.)99

100