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Year : 2020  |  Volume : 43  |  Issue : 3  |  Page : 154-161  

Shielding designing of 241Am-Be neutron source housing experiment and Monte Carlo simulation

1 Radiological Physics and Advisory Division, Health, Safety and Environment Group, Bhabha Atomic Research Centre; Department of Atomic Energy, Homi Bhabha National Institute, Anushaktinagar, Mumbai, Maharashtra, India
2 Radiological Physics and Advisory Division, Health, Safety and Environment Group, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

Date of Submission09-Oct-2020
Date of Decision10-Nov-2020
Date of Acceptance12-Nov-2020
Date of Web Publication6-Jan-2021

Correspondence Address:
A K Bakshi
Radiological Physics and Advisory Division, Health, Safety and Environment Group, Bhabha Atomic Research Centre, Mumbai - 400 094, Maharashtra; Homi Bhabha National Institute, Anushaktinagar, Mumbai - 400 094, Maharashtra
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Source of Support: None, Conflict of Interest: None

DOI: 10.4103/rpe.rpe_54_20

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Shielding of a neutron source housing for 1 Ci 241Am-Be source has been designed and fabricated based on the simulation carried out using Monte Carlo code FLUKA. Neutron and photon dose equivalent rates at the surface and at 1 m distance from the surface of the housing were calculated using FLUKA simulation and measured using gamma and neutron dose rate meters. The calculated and measured gamma and neutron dose equivalent rates agree well. Neutron spectra outside the source housing were generated using FLUKA simulation and measured with the ROSPEC + simple scintillation spectrometer neutron spectrometry system and also agree reasonably well. Gamma spectra outside the source housing and residual activity due to activation products in stainless steel and lead of the housing were also generated through simulation.

Keywords: 241Am-Be, FLUKA simulation, neutron source, residual activity, shielding housing

How to cite this article:
Bakshi A K, Chatterjee S, Dawn S, Beck M, Selvam T P. Shielding designing of 241Am-Be neutron source housing experiment and Monte Carlo simulation. Radiat Prot Environ 2020;43:154-61

How to cite this URL:
Bakshi A K, Chatterjee S, Dawn S, Beck M, Selvam T P. Shielding designing of 241Am-Be neutron source housing experiment and Monte Carlo simulation. Radiat Prot Environ [serial online] 2020 [cited 2023 May 28];43:154-61. Available from: https://www.rpe.org.in/text.asp?2020/43/3/154/306283

  Introduction Top

Neutron personnel monitoring in India is carried out based on polyallyl diglycol carbonate (PADC) based nuclear track detector. The trade name of this detector is CR-39. Thermal neutron-sensitive thermo luminescence dosimeters (TLDs) have also been developed[1] earlier by us, and a field trial in nuclear fuel cycle facilities is in progress. The types of institutions covered under neutron monitoring are nuclear power plants, fuel-processing facilities, research reactors, research activities involving neutron sources, and oil-well logging industries. One of the recommended laboratory neutron sources for calibration of neutron dosimeters as per ISO[2] is 241Am-Be source. In this source, neutrons of energy in the range of 100 keV–10 MeV are produced through the reaction 9Be(α, n)12C.[2] The fluence-averaged energy of this neutron source is about 4.16 MeV. In addition to neutrons, gamma-ray of energy 59.5 keV from 241Am source and prompt gamma of energy 4.43 MeV from the excited state of 12C* are also produced.[3] 241Am-Be source of activity 1 Ci is used for routine calibration of neutron dosimeters. The calibration room housing this neutron source also houses a 137Cs-based gamma irradiation facility for the calibration of TLD personnel dosimeter. The presence of 241Am-Be source without proper shielding is considered to be a potential source of the gamma radiation background in the calibration room which is unwanted for TLD calibration as well as potential radiation hazards for personnel involving in calibration of TLDs.

In view of the above, the present study was undertaken to achieve the following objectives: (i) Optimization of shielding and fabrication of a movable source housing for 241Am-Be neutron source, (ii) to reduce the background radiation level due to 241Am-Be source in the room to well below 10 μSv/h at 1 m from the source housing, (iii) to explore on the use of the source housing as a lower energy neutron irradiation facility for the irradiation of thermal neutron detectors, and (iv) estimation of residual activity in the shielding material of the source housing in long term as the 241Am-Be source has a half-life of 432 years.

In a similar study, Mazrou et al.[4] have reported Monte Carlo simulation and experimental measurement for characterizing the beam qualities for 241Am-Be source irradiator. Monte Carlo simulation was carried out using Monte Carlo N-particle and experimental measurement using BF3- and 3He-based neutron area dosimeters. Neutron fluence rate and ambient dose equivalent rate were estimated using both methods and compared in their study. In this study, source housing was designed by the manufacturer and the influence of biological shielding was carried out in the metrology and dosimetric parameters of the neutron beam, whereas in the present study, the source housing design was optimized based on the simulation and fabricated locally and measurement was carried out subsequently with neutron dose rate meter and gamma survey meter to validate the simulation.

  Materials and Methods Top

Monte Carlo simulation

The shielding material considered for the 1 Ci 241Am-Be neutrons is high-density polyethylene (HDPE) of density 0.95 g/cm3 as it acts as a good moderator for neutrons. In addition to HDPE, lead (Pb) and stainless steel (SS) were considered as the shielding materials for gamma. Optimization of the thickness of the HDPE, Pb, and SS was carried out using Monte Carlo code FLUKA.[5],[6] FLUKA can simulate the interaction and propagation in matter of about 60 different particles, including photons, electrons, neutrons, and heavy ions. FLUKA can handle complex geometries using an improved version of the well-known Combinatorial Geometry package. The geometry of the simulation was made using FLAIR (version 2.02).[7]

The neutron source spectra needed for the simulation were taken from ISO.[2] The source was assumed to be point isotropic. In the FLUKA calculations, the transport cutoff energies for photons, neutrons, and electrons were taken as 1 keV and 10-8 keV and 10 keV, respectively. When the energy of these particles falls below the cutoff energy, they are not transported further. A total of 8 × 107 neutron histories were followed for 5 cycles. It can be noted that 59.5 keV gammas from 241Am were not considered in the simulations as the thickness of shielding material of HDPE and SS is considered to be sufficient to attenuate it to negligible level.

For the shielding of the neutrons and the photons, different shielding designs were considered in the FLUKA simulations [Figure 1]a,[Figure 1]b,[Figure 1]c,[Figure 1]d. [Figure 1]a shows the housing having the thickness of 15 cm HDPE on all sides except the bottom and the top of the housing, where the thickness of 13 cm HDPE was considered. The inner side of the housing contains a cylindrical cavity meant for storing the source with wall made from 2 mm thick Pb sandwiched between 1.2 mm thick SS. The inner diameter of the cylinder is 4 cm and height is 6 cm. [Figure 1]b shows the same size of the housing made from HDPE with combination of SS and Pb lining as mentioned above, except that SS lining with Pb has been moved at the exterior part of the housing. In [Figure 1]c, the HDPE thickness of the sidewall has been increased to 25 cm, keeping the thicknesses at the top and bottom (13 cm) same as mentioned above [Figure 1]b. Here, the SS lining has been kept at the exterior part of the housing. [Figure 1]d shows the housing of similar size as that used in [Figure 1]c, but the SS + Pb lining has been placed inside the housing near to the source in order to reduce the weight of the housing.
Figure 1: Different designs considered in the simulation for designing of 241Am-Be source housing involving high-density polyethylene, stainless steel, and lead as shielding materials (a) 15 cm HDPE on all sides except bottom and top of housing and cylindrical cavity of wall made from SS lining of thickness 1.2 mm sandwiching 2 mm thick Pb (b) housing dimension and material same as (a) except SS lining sandwiching Pb at outer side of housing, (c) 25 cm of HDPE thickness on all sides except top and bottom which are 13 cm thick wall is 25 cm and ss lining same as in (b) and (d) Housing dimension and material same as (c) except the cylindrical hole meant for source storage is made of SS lining sandwiching Pb including the inner side of the top part of thickness same as (a)

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Calculation of ambient dose equivalent rate H* (10) was carried out with USRBIN card of FLUKA based on the fluence to ambient dose equivalent conversion coefficients of photons and neutrons. The conversion of fluence to dose equivalent for photons and neutron in FLUKA is carried out using a user routine based on ICRP-74.[8] Ambient dose equivalents due to neutron and gamma were calculated at five positions, namely four sides, and over the top of the housing. The detectors were placed very near to the housing, so that the surface dose can be calculated. Simulation was also carried out to determine neutron and gamma dose equivalent at 1 m from the surface of the housing.

[Table 1] shows the values of H* (10) for neutrons and gamma photons for the above mentioned designs [Figure 1]d at the surface of the housing. The neutron yield from 241Am-Be source was taken as 2.50 × 106 neutrons/s for 1 Ci activity.[9]
Table 1: Simulated neutron and photon ambient dose equivalent rates [Ḣ*(10)] (for 1 Ci of 241Am-Be neutron source) at the surface of the housing along the horizontal and vertical direction

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Induced activity calculations were also carried out by FLUKA using RESNUCLEI card for optimized shielding of the housing, as shown in [Figure 1]d. Different isotopes generated in different shielding materials were identified. The total induced activity produced in different parts of the housing having different materials such as SS, Pb, and HDPE was also evaluated for different storage times using the USRBIN card of FLUKA. The irradiation profile consisted of an isotropic source (1 Ci 241Am-Be) for a total irradiation time of up to 10 years. The total induced activity was evaluated at different positions (SS near the source, Pb sandwiched between the SS lining, SS after lead in the cylindrical hole of the source housing meant for the storage of source) due to the residual nuclides generated because of the storage of the source.


For the measurement of gamma dose equivalent rate, RAMION portable 500 cc ion chamber-based survey meter manufactured by M/s Rotem Industries Ltd., Israel, was used. It is calibrated against 137Cs gamma-rays having known output traceable to National Standards, and the calibration factor is fed into the system through software. This survey meter can measure dose equivalent rate in the range of 1 μSv/h to 500 mSv/h in the dose rate mode and 0.01 μSv to 10 Sv in the integrated mode for gamma-, beta-, and X-rays. The energy response of this survey meter is better than ±20% at 20 keV to 1.3 MeV. The dose equivalents are measured with an accuracy of ± 10%. In the higher energy range beyond 1.3 MeV, it has been reported that RAMION over-responds to the extent of 5% at 6 MeV produced by 16N decay.[10]

For the measurement of neutron dose equivalent rate, REM counter model LB6411 manufactured by Berthold Technologies GmbH and Co, Germany, was used. It has a cylindrical 3He tube of active length 40 mm, at its center, with a 250 mm diameter spherical polyethylene moderator.[11] It is calibrated against 241Am-Be neutron source traceable to National Standards in terms of ambient dose equivalent rate H* (10). It can measure the neutron dose equivalent rate from thermal energies up to 20 MeV. The energy dependence is ±30% between 50 keV and 10 MeV. It can measure the neutron dose rate from 0.03 μSv/h to 100 mSv/h with a gamma discrimination factor of 3 × 103 (https://www.berthold.com/en/rp/lb-6411-neutron-probe). The optimized spherical geometry of the tube and moderator results in an angular response of ±10% in the energy range of 1–20 MeV of the detector over the entire angular range of 0°–180°. Along with acquisition and control unit model 5700/RTM-WM and operational software manufactured by M/s ELSE, Italy, the ambient dose equivalent can be integrated for predecided durations such as 1 min, 10 min, 1 h, and 24 h. Each data point of neutron dose equivalent rates is derived after taking the average of 5 dose equivalents each integrated over 1 min.

For the measurements of neutron spectrum and ambient dose equivalent rate outside the source housing, ROSPEC and simple scintillation spectrometer (SSS) neutron spectrometer systems, procured from M/s BTI, Canada, were used. ROSPEC is a rotational spectrometer having 6 gas-based detectors for different energy ranges from thermal to 4.5 MeV, whereas in SSS, there is an array of small plastic scintillators coupled to a photomultiplier tube which can generate a spectrum in the energy range of 4–18 MeV. With the unfolding software provided with the system, it is possible to unfold the neutron spectrum. More details on these spectrometer systems can be found in the literature.[12],[13] The measurement of neutron spectra was carried out for 12 h considering the lower dose equivalent rate (~10–12 μSv/h) outside the housing. It is noted that these spectrometer systems are provided with a calibration certificate against bare 252Cf and D2O-moderated 252Cf neutron sources based on the measurement carried out at the National Institute of Standards and Technology, USA, and the uncertainty in the estimation dose equivalent for the above sources was in the range of ± 2%.

  Results and Discussions Top

Gamma and neutron ambient dose equivalent rates

Calculated neutron and photon ambient dose equivalents using Monte Carlo code FLUKA are presented in [Table 1] for different trial designs of the housing. It can be seen that in [Table 1], the neutron and photon dose equivalent rates (H* [10]) are lower for the designs shown in [Figure 1]c and [Figure 1]d as compared to the designs shown in [Figure 1]a and [Figure 1]b along the horizontal direction. It can also be seen that the gamma dose equivalent rate is slightly higher for [Figure 1]d along horizontal direction with respect to the design shown in [Figure 1]c which may be attributed to more contribution of prompt gamma of energy 2.22 MeV from (n, γ) reaction with hydrogen of the additional thickness of HDPE and 4.43 MeV gamma from the excited state of 12C* of 241Am-Be source, whereas in vertical direction, there is no change in the dose equivalent rates and small change observed is within statistical error. It is noted that the gamma dose outside the source housing both horizontal and vertical direction is attributed to unattenuated gamma of energy 4.43 MeV and 2.2 MeV gamma produced in HDPE by thermal neutron. The reduction in the neutron dose equivalent rate along horizontal direction is attributed to the higher thickness of HDPE used in the designs shown in [Figure 1]c and [Figure 1]d. Out of the two design considerations shown in [Figure 1]c and [Figure 1]d, [Figure 1]d is finalized as this will reduce the overall weight of the housing without compromising on the radiation levels at the surface and at 1 m from the surface. [Figure 2] shows the metal lining details of the source housing for the finalized design [Figure 1]d. The photograph of the fabricated source housing is shown in [Figure 3]a and [Figure 3]b. After the fabrication of the housing, gamma dose rate and neutron dose rate calculations were verified at the same positions. [Table 2] shows the experimental measurement and FLUKA-calculated neutron and photon dose equivalent rates at different positions outside the housing. It is seen that the dose equivalent rates are higher at the top position of the housing with respect to side positions. This is attributed to the higher attenuation of gamma and neutron along the horizontal direction due to higher thickness (25 cm) of HDPE than the vertical (13 cm) direction. The neutron and photon dose equivalents calculated by FLUKA agree well with the measured values [Table 2]. However, it is observed that calculated dose equivalents are slightly lower than the measured values, which may be attributed to the scattered radiation due to the presence of room which was not considered in the simulation. Neutron dose equivalent rate at the front side of the housing was also measured by ROSPEC + SSS spectrometer and is found to be 11.66 μSv/h which matches well with the measured values of Berthold REM counter (13.06 μSv/h). The slight difference in dose equivalent measured by ROSPEC with respect to REM counter is attributed uncertainty in the measurement by REM counter. The measured and simulated dose equivalent rates at 1 m from the surface of the housing are in the range of 2–3 μSv/h for both photon and neutron and also agree well with each other. As the source housing will be kept at one corner of the calibration room when the source not in use, the dose rate at far away distance (>3 m) will be much <1 μSv/h. Hence it can be considered that optimized design of neutron source housing reduced the background gamma and neutron radiation level in the calibration room to an acceptable level. The gamma background level in the calibration room at a distance of 3 m is insignificant and hence will not influence the irradiation of TLD badge to Cs-137 source.
Table 2: Comparison of measured and simulated Photon and Neutron Dose equivalent rates at the surface and at 1 m distance from surface of the source housing

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Figure 2: Metal lining (lead and stainless steel) in enlarged view used in the geometry of Figure 1d

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Figure 3: Photographs of neutron source housing (a) in closed condition (b) in open condition

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Neutron fluence spectra

[Figure 4] shows the neutron fluence spectra calculated using FLUKA (i) inside the source housing which is without any shielding material and (ii) outside the housing at vertical and horizontal directions along with the corresponding error bar. It can be seen that the spectrum with shielding (outside the housing) has a prominent thermal peak. The thermal peak intensity has been increased due to the moderation of fast neutron in shielding medium by an order of magnitude with respect to that of without shielding. The thermal component contributes about 37% at horizontal position and 41% at vertical position with respect to the total fluence due to the presence of shielding. The neutron fluence and the photon fluence are higher on the vertical side than at the horizontal side. The lower thickness of HDPE in the vertical side can be attributed to this. [Figure 5] shows the comparison of FLUKA-calculated spectrum and measured fluence spectrum by ROSPEC + SSS spectrometer system. The spectrum matches reasonably well for high- and low-energy peaks. Due to poor resolution in the lower energy range (up to 1 eV) of the spectrometer, the measured thermal neutron peak appeared flat. Besides the design of the source housing, an effort was also made to use the housing as a facility for irradiation of thermal neutron detector at the surface of the housing, in view of the presence of lower energy neutrons [Figure 5]. For this purpose, fluence at the position of the source (at 1 cm) inside the housing and outside (at the surface) due to thermal neutron up to the energy of 0.4 eV was calculated. It is about 14% inside the housing, whereas it is about 37% outside (horizontal) the housing. Therefore, thermal neutron detector can be irradiated outside this housing for the purpose of testing, subject to the condition that the sensitivity of the detector to fast neutron and gamma should be negligible.
Figure 4: Calculated neutron fluence spectrum of 241Am-Be inside the housing, along vertical axis and along horizontal axis positions outside the source housing for design shown in Figure 1d

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Figure 5: Calculated and measured neutron spectrum by ROSPEC + simple scintillation spectrometer outside the 241Am-Be source housing at 20 cm from the outside surface. Both fluence spectrums were normalized with respect to the highest peak position for comparison

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Prompt gamma spectra, residual activity, and associated gamma radiation level

Gamma spectrum, due to prompt gamma from activation products and (n, γ) reactions in HDPE, outside the source housing was calculated using FLUKA code and is shown in [Figure 6]. It has several peaks which are attributed to the prompt gamma-rays due to the following nuclear reactions:
Figure 6: Photon spectra at horizontal and vertical positions outside the final design of the 241Am-Be source housing

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H1 + n = H2+ γ (2.223 MeV)

Fe56 + n = Fe57+ γ (7.631 MeV)

Pb204 + n = Pb205+ γ (4.245 MeV).

Being the neutron source housing and considering the production of radioactivity in the materials such as SS and Pb of the source housing, residual activity calculation was also carried out for SS and Pb for a time of 6 months, 1 year, and 10 years for continuous storage of 241Am-Be source inside the housing. Some of the major isotopes produced in SS are Ni63, Ni59, Co58, Fe55, and Cr51. Some of the major isotopes produced in Pb are Pb209 and Pb205. It is noted that these isotopes such as Pb209 (T1/2: 3.25 years), Fe55 (T1/2 = 2.7 years), Ni63 (T1/2 = 96 years), Ni59 (T1/2 = 7.5 × 104 years), and Pb209(T1/2 = 3.25 years) have half-lives more than few years and are the potential source of excess radiation level in the future. The details of the radioisotopes produced in SS and Pb are shown in [Figure 7]a,[Figure 7]b,[Figure 7]c. The amount of residual activity produced after a period of 6 months, 1 year, and 10 years of continuous storage of 1 Ci 241Am-Be source is given in [Table 3]. The total residual activity is about 2 mCi in SS and 0.2 mCi in Pb for a storage period of 10 years. After a period of 4 years of the fabrication and source storage in the housing, gamma radiation level was measured directly above the cylindrical hole without source in it. The gamma radiation level was found to be 0.71 ± 0.06 μSv/h. It may be noted that as the RAMION has over-response in the higher energy range of gamma,[10] the gamma radiation level of 0.71 μSv/h is considered to a conservative estimate. The gamma radiation level just above the cylindrical hole with the 241Am-Be source in it was found to be 475 ± 26 μSv/h. From the above measurement, it can be inferred that 0.71 ± 0.06 μSv/h which is about 0.15% of the total dose is only due to prompt gamma from the induced radioactivity in SS and Pb of the housing. When the measurement is made in another portion of the housing (on the steel part, 10 cm outside the periphery of the cylindrical hole), it is 0.06 ± 0.01 μSv/h. In view of the activation products in SS and Pb, after a period of 10 years, it is expected that the gamma radiation level may increase slightly outside the housing but that will be negligible with respect to that of the actual source and 4.43 MeV prompt gamma produced in HDPE.
Figure 7: Details of the residual nuclei produced due to activation products (a) in stainless steel near the source (b) in lead which is sandwiched between stainless steel plates (c) in SS lining away from the source

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Table 3: Details of the residual activity (Bq) produced due to activation products in SS and Pb of the source housing

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  Conclusions Top

Shielding of a neutron source housing for storing 241Am-Be source has been designed using Monte Carlo code FLUKA and fabricated as per the design. The shielding material and thickness of the source housing were optimized based on calculation carried out using Monte Carlo simulation code FLUKA. Subsequent to this, the source housing was fabricated and the neutron and gamma dose equivalent rates were measured at the surface and at 1 m from the housing and compared with the calculated values. The calculated dose equivalent rates agree well with those of the measured values. Neutron spectrum outside the source housing was measured using ROSPEC + SSS spectrometer system. The presence of prominent lower energy neutron peak in the neutron fluence spectrum outside the housing can be exploited for the irradiation of thermal neutron-sensitive detector without disturbing the source housing for a long time. The residual activity due to neutron-induced activation products in SS and Pb of the housing was estimated for different storage periods up to 10 years and is found to be 0.2 mCi in Pb and about 2 mCi in SS for which there will be a negligible increase in gamma radiation level outside the housing. The source housing will help in bringing down the radiation level in the calibration room to an acceptable level of 1 μSv/h at a distance of more than 2 m from the housing when the source is in storage condition.


The authors are grateful to Shri R M Suresh Babu, Director, Health Safety and Environment Group, and Dr B K Sapra, Head, RP&AD, BARC, for their encouragement in the work.

Financial support and sponsorship


Conflicts of interest

There are no conflicts of interest.

  References Top

Bakshi AK, Pradhan AS, Kher RK, Srivastava K, Varadharajan G, Chatterjee S, et al. Study on the response of indigenously developed CaSO4:Dy phosphor-based neutron dosemeter. Radiat Prot Dosimetry 2009;133:73-80.  Back to cited text no. 1
International Organization for Standardization (ISO), Reference neutron radiations: Part1 Characteristics and Methods of Production.Report No ISO-8529, (2001), Swizerland.  Back to cited text no. 2
Kamboj B, Shahani M. Precise measurement of the gamma to neutron ratio of an Am-α-Be neutron source using an improved manganese bath technique. Nucl Instrum Methods Phys Res A 1986;244:513-5.  Back to cited text no. 3
Mazrou H, Sidahmed T, Allab M. Monte-Carlo investigation of radiation beam quality of the CRNA neutron irradiator for calibration purposes. Appl Radiat Isot 2010;68:1915-21.  Back to cited text no. 4
Battistoni, Giuseppe, F. Cerutti, A. Fasso, A. Ferrari, S. Muraro, J. Ranft, S. Roesler, and P. R. Sala. “The FLUKA code: Description and benchmarking.” In AIP Conference proceedings, American Institute of Physics,(USA), 2007;896:31-49.  Back to cited text no. 5
Ferrari, Alfredo, Paola R. Sala, Alberto Fasso, Johannes Ranft, and U. Siegen. FLUKA: a multi-particle transport code. No. SLAC-R-773. Stanford Linear Accelerator Center (SLAC)(USA), 2005.  Back to cited text no. 6
Vlachoudis V. FLAIR: A Powerful but user Friendly Graphical Interface for FLUKA, Processing International Conference on Mathematics, Computational Methods & Reactor Physics. (M&C 2009). New York: Saratoga Springs; 2009.  Back to cited text no. 7
Roesler, Stefan, and Graham R. Stevenson. “deq99. f-A FLUKA user-routine converting fluence into effective dose and ambient dose equivalent.” CERN Technical Note CERNSC2006070RPTN, European Organization for Nuclear Research, Geneva, Switzerland 2006.  Back to cited text no. 8
Mowlavi AA, Koohi-Fayegh R. Determination of 4.438 MeV γ-ray to neutron emission ratio from a 241Am-Be neutron source. Appl Radiat Isot 2004;60:959-62.  Back to cited text no. 9
Chabot G E, Report on the responses of different radiation dosimeter using reactor produced N-16 gamma source, Internal report,Department of Physics and Applied Physics University of Massachusetts Lowell, MA, USA, 1997, private communication.  Back to cited text no. 10
Klett A, Burgkhardt B. The new remcounter LB6411: Measurement of Neutron Ambient dose Equivalent H*(10) According to ICRP60 with high Sensitivity, Nuclear Science Symposium, 1996. Conference Record, 1996 IEEE. IEEE; 1996. p. 132-4.  Back to cited text no. 11
Devine RT, Romero LL, Gray DW, Seagraves DT, Olsher RH, Johnson JP. Evaluation of spectrum measurement devices for operational use. Nucl Instrum Methods Phys Res A 2002;476:416-22.  Back to cited text no. 12
Vanhavere F, Vermeersch F, Chartier J, Itié C, Rosenstock W, Köble T, et al. A comparison of different neutron spectroscopy systems at the reactor facility VENUS. Nucl Instrum Methods Phys Res A 2002;476:395-9.  Back to cited text no. 13


  [Figure 1], [Figure 2], [Figure 3], [Figure 4], [Figure 5], [Figure 6], [Figure 7]

  [Table 1], [Table 2], [Table 3]


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