Borated Polyethylene Research Articles

INVESTIGATION OF THE EXTENDED RANGE REM-COUNTER SMARTREM-LINUS IN REFERENCE AND WORKPLACE FIELDS EXPECTED AROUND HIGH-ENERGY ACCELERATORS.

Radiation survey instrumentation is adequate for the use around high-energy accelerators if capable to measure the dose arising from neutrons with energies ranging from thermal up to a few gigaelectronvolts. The SmartREM-LINUS is a commercial extended range rem-counter, consisting of a central (3)He-proportional counter surrounded by a spherical moderator made of borated polyethylene with an internal shield made of lead. The dose rate indicated by the SmartREM-LINUS was investigated for two different irradiation conditions. The linearity and the angular dependence of the indicated dose rate were investigated using reference neutron fields produced by (241)AmBe and (252)Cf. Additional measurements were performed in two different workplace fields with a component of neutrons with energies >20 MeV, namely the CERN-EU high-energy reference field and near the beam dump of the SwissFEL injector test facility. The measured dose rates were compared to a commercial rem-counter (WENDI2) and the results of Monte Carlo simulations.

SU‐F‐T‐655: Evaluation of Neutron Dose Equivalent Inside of Treatment Vault of Radiotherapy by Photoneutron Production in the Primary Barriers

Purpose:Photon beams with energy higher than 10 MV interact with metal material in the primary barriers, where lead or steel have been widely used, neutrons can be generated. Monte Carlo simulations were performed to simulate the production of photoneutrons and the neutron shielding effect.Methods:For two photon beam energies, 15 MV and 18 MV, we simulated to strike metal sheets (steel and lead), and the ambient dose equivalents were calculated at the isocenter (in the patient plane) while delivering 1 Gy to the patient. For these cases, the thickness of the neutron shielding materials (Borated polyethylene (BPE) and concrete) were simulated to reduce the patient exposure by neutron doses.Results:When 18 MV photons interact with the metal sheets in the primary barrier, the evaluated neutron doses at the isocenter inside the treatment vault were 48.7 µSv and 7.3 µSv for lead and steel, respectively. In case of 15 MV photons, the calculated neutron doses were 18.6 µSv and 0.6 µSv for lead and steel, respectively. The neutron dose delivered to the patient can be reduced to negligible levels by including a 10 cm thick sheet of BPE or 22 cm thick sheet of concrete.Conclusion:When bunker shielding is designed with a primary barrier including a metal sheet inside the wall for a high energy machine, proper neutron shielding should be constructed to avoid undesirable extra dose.

Reliability of Monte Carlo simulations in modeling neutron yields from a shielded fission source

Using the combination of a neutron-sensitive 6Li glass scintillator detector with a neutron-insensitive 7Li glass scintillator detector, we are able to make an accurate measurement of the capture rate of fission neutrons on 6Li. We used this detector with a 252Cf neutron source to measure the effects of both non-borated polyethylene and 5% borated polyethylene shielding on detection rates over a range of shielding thicknesses. Both of these measurements were compared with MCNP calculations to determine how well the calculations reproduced the measurements. When the source is highly shielded, the number of interactions experienced by each neutron prior to arriving at the detector is large, so it is important to compare Monte Carlo modeling with actual experimental measurements. MCNP reproduces the data fairly well, but it does generally underestimate detector efficiency both with and without polyethylene shielding. For non-borated polyethylene it underestimates the measured value by an average of 8%. This increases to an average of 11% for borated polyethylene.

Use of borated polyethylene to improve low energy response of a prompt gamma based neutron dosimeter

The feasibility of using a combined sample of borated polyethylene and normal polyethylene to estimate neutron ambient dose equivalent from measured prompt gamma emissions is investigated theoretically to demonstrate improvements in low energy neutron dose response compared to only polyethylene. Monte Carlo simulations have been carried out using the FLUKA code to calculate the response of boron, hydrogen and carbon prompt gamma emissions to mono energetic neutrons. The weighted least square method is employed to arrive at the best linear combination of these responses that approximates the ICRP fluence to dose conversion coefficients well in the energy range of 10−8MeV to 14MeV. The configuration of the combined system is optimized through FLUKA simulations. The proposed method is validated theoretically with five different workplace neutron spectra with satisfactory outcome.

A technique for searching for the 2K capture in 124Xe with a copper proportional counter

An experimental technique for searching for the 2K capture in 124Xe with a large low-background copper proportional counter is described. Such an experiment is conducted at the Baksan Neutrino Observatory of the Institute for Nuclear Research of the Russian Academy of Sciences. The experimental setup is located in the Low-Background Deep-Level Laboratory at a depth of 4900 m.w.e., where the flux of muons of cosmic rays is suppressed by a factor of 107 relative to that at the Earth’s surface. The setup incorporates a proportional counter and low-background shielding (18 cm of copper, 15 cm of lead, and 8 cm of borated polyethylene). The results of processing the data obtained in 5 months of live measurement time are presented. A new limit on the half-life of 124Xe with respect to the 2K capture is set at the level of 2.5 × 1021 years.

Pulsed Power Active Interrogation of Shielded Fissionable Material

We report on a collaborative test campaign conducted at the Naval Research Laboratory's Mercury pulsed power facility in December of 2012. The experiment sought to use Mercury in the Intense Pulsed Active Detection (IPAD) [1] mode to interrogate a fissionable material target (depleted uranium, DU) and benchmark the effects of shielding the target with either a low-Z (2% borated high-density polyethylene, BPE) or high-Z (steel) material. A large suite of instrumentation, including 3He, BF3, NaI(Tl), and liquid scintillation detectors were used to measure the delayed γ and neutron signatures from the DU. The test campaign consisted of a series of single IPAD pulses, i.e., “shots,” employing incremental shielding configurations of BPE (up to 50 g/cm2) and steel (up to 150 g/cm2) encapsulating the DU target. We show the results from each detector array, for varying amounts of shielding, in terms of the signal-to-noise vs. time.

SU‐E‐T‐239: Design and Evaluation of a Nozzle Shielding System in Passively Scattered Proton Therapy

Purpose:To design, implement and evaluate a shielding system that will reduce out‐of‐field dose experienced by the patient and associated electronic systems in passively scattered proton therapy treatment.Methods:A multi‐stage neutron shielding system was retrofitted to the Gantry 1 treatment nozzle at Loma Linda University Medical Center. The system uses multiple borated polyethylene plates staged after the primary beam modifying devices to attenuate and absorb neutrons produced by such devices. This arrangement locates increasing levels of shielding between the sources of secondary particles in the nozzle and the patient. Additionally, the design of this shielding structure allows it to be easily retrofitted to an existing proton nozzle system without impacting design or treatment beam characteristics. The effectiveness of the shielding was evaluated both through experimental measurements and Geant4 Monte Carlo simulations.Results:Measurements were completed with Landauer Luxel+ dosimeters that use optically stimulated luminescence and CR‐39 to detect fast neutrons, thermal neutrons, protons, photons and beta particles. Measurements of a 250 MeV proton beam indicated that the shielding system reduced out‐of‐field dose to the patient by almost half with dose equivalent values at 50 and 40 cm from the field edge decreasing from 0.965 and 1.262 mSv/Gy to 0.596 and 0.777 mSv/Gy respectively. The installation of the multi‐stage shielding system also reduced dose equivalent experienced by electronic systems installed in the treatment room by up to 80%. Geant4 simulations were also used to evaluate the neutron fluence at various positions in the treatment room as well as provide information on microdosimetry spectra within the patient and treatment room.Conclusion:The shielding system described above proved to be an effective an inexpensive method of reducing out‐of‐field doses to the patient and electronic systems and can be easily retrofitted to existing passive scattering nozzles.

An introduction to γ3 a new versatile ultralow background gamma spectrometer. Background description and analysis

The γ3 setup has been designed as a versatile, high sensitivity spectrometry platform. State-of-the art techniques have been implemented to reduce its background to minimum level even though the system is installed at ground level. The shield design and background performance of the setup are presented. The spectrometer is composed of three identical HPGe detectors for high detection efficiency or coincidence measurement and can accommodate several sample geometries. Its shield includes three layers of increasing purity lead, a cosmic veto, an inner borated polyethylene layer, and a radon-free gas injection system. The spectrometer normalized background count rate is 4.4 counts per minutekgGe−1 (in the 40–2500keV energy range). Its background characteristics, cosmic veto efficiency, and radon-free gas injection performances are discussed.

TANGRA-Setup for the Investigation of Nuclear Fission Induced by 14.1 MeV Neutrons

The new experimental setup TANGRA (Tagged Neutrons & Gamma Rays), for the investigation of neutron induced nuclear reactions, e.g. (n,xn’), (n,xn’γ), (n,γ), (n,f), on a number of important isotopes for nuclear science and engineering (235,238U, 237Np, 239Pu, 244,245,248Cm) is under construction and being tested at the Frank Laboratory of Neutron Physics (FLNP) of the Joint Institute for Nuclear Research (JINR) in Dubna.The TANGRA setup consists of: a portable neutron generator ING-27, with a 64-pixel Si charge-particle detector incorporated into its vacuum chamber for registering of α-particles formed in the T(d, n)4He reaction, as a source of 14.1 MeV steady-state neutrons radiation with an intensity of ∼5x107n/s; a combined iron (Fe), borated polyethylene (BPE) and lead (Pb) compact shielding-collimator; a reconfigurable multi-detector (neutron plus gamma ray detecting system); a fast computer with 2 (x16 channels) PCI-E 100 MHz ADC cards for data acquisition and hard disk storage; Linux ROOT data acquisition, visualization and analysis software. The signals from the α-particle detector are used to ‘tag’ the neutrons with the coincident α-particles. Counting the coincidences between the α-particle and the reaction-product detectors in a 20ns time-interval improves the effect/background-ratio by a factor of ∼200 as well as the accuracy in the neutron flux determination, which decreases noticeably the overall experimental data uncertainty.

Preliminary production of 211At at the Texas A&M University Cyclotron Institute.

A feasibility study for the production of the alpha particle-emitting radionuclide At was performed at the Texas A&M University Cyclotron Institute as part of the Interdisciplinary Radioisotope Production and Radiochemistry Program. The mission of this program centers upon the production of radionuclides for use in diagnostic and therapeutic nuclear medicine with the primary focus on development of novel therapeutic strategies. As a first step in establishing this program, two goals were outlined: (i) verify production of At and compare results to published data, and (ii) evaluate shielding and radiological safety issues for large-scale implementation using an external target. The radionuclide At was produced via the Bi (α, 2n) At reaction using the K500 cyclotron. Two experiments were conducted, using beam energies of 27.8 MeV and 25.3 MeV, respectively. The resulting yields for At were found to be 36.0 MBq μA h and 12.4 MBq μA h, respectively, which fall within the range of published yield data. Strategies for increasing absolute yield and production efficiency were also evaluated, which focused chiefly on using a new target designed for use with the K150 cyclotron, which will enable the use of a higher beam current. Finally, neutron and gamma dose rates during production were evaluated by using the Monte Carlo code MCNPX. It was determined that a simple structure consisting of 4-in thick borated polyethylene will reduce the neutron dose rate within the cyclotron production vault by approximately a factor of 2, thereby decreasing activation of equipment.

SU‐E‐T‐495: Neutron Induced Electronics Failure Rate Analysis for a Single Room Proton Accelerator

Purpose:To determine the failure rate as a function of neutron dose of the range modulator's servo motor controller system (SMCS) while shielded with Borated Polyethylene (BPE) and unshielded in a single room proton accelerator.Methods:Two experimental setups were constructed using two servo motor controllers and two motors. Each SMCS was then placed 30 cm from the end of the plugged proton accelerator applicator. The motor was then turned on and observed from outside of the vault while being irradiated to known neutron doses determined from bubble detector measurements. Anytime the motor deviated from the programmed motion a failure was recorded along with the delivered dose. The experiment was repeated using 9 cm of BPE shielding surrounding the SMCS.Results:Ten SMCS failures were recorded in each experiment. The dose per monitor unit for the unshielded SMCS was 0.0211 mSv/MU and 0.0144 mSv/MU for the shielded SMCS. The mean dose to produce a failure for the unshielded SMCS was 63.5 ± 58.3 mSv versus 17.0 ±12.2 mSv for the shielded. The mean number of MUs between failures were 2297 ± 1891 MU for the unshielded SMCS and 2122 ± 1523 MU for the shielded. A Wilcoxon Signed Ranked test showed the dose between failures were significantly different (P value = 0.044) while the number of MUs between failures were not (P value = 1.000). Statistical analysis determined a SMCS neutron dose of 5.3 mSv produces a 5% chance of failure. Depending on the workload and location of the SMCS, this failure rate could impede clinical workflow.Conclusion:BPE shielding was shown to not reduce the average failure of the SMCS and relocation of the system outside of the accelerator vault was required to lower the failure rate enough to avoid impeding clinical work flow.

Simulation and Design of a Simple and Easy-to-use Small-scale Neutron Source at Kyoto University

A simple and easy-to-use compact neutron source based on a low power level proton accelerator (proton energy 3.5 MeV and 0.35kW beam power) at Kyoto University was designed with the conception of low cost, compact size, high safety and intensive thermal neutron flux via Monte Carlo method with PHITS code. By utilizing (p, n) reactions in a beryllium target coupled to a polyethylene moderator and graphite reflector with a wing configuration, this facility is expected to produce time-averaged thermal neutron fluxes suitable for neutron scattering and development of instrumentation, and play a role in educating students in neutron science and performing research with neutrons. Borated polyethylene (BPE) and ordinary concrete were combined to shield the neutron and photon. By using niobium as target backing and water as cooler, it is promising to cope with the problem of thermal damage and hydrogen embrittlement damage. The sizes of moderator and reflector are optimized to have thermal neutron flux as high as possible, while keeping the low ratio of fast neutron flux to thermal neutron flux. The neutron and gamma dose equivalent rates were evaluated and the current shielding configuration is acceptable.

Radiological shielding design for the TOF polarisation analysis spectrometer Pelican at the OPAL reactor

The shielding for the time-of-flight Polarisation Analysis Spectrometer (Pelican) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 4B. The proposed shielding design has produced compact shielding assemblies, such as the Pelican monochromator drum, lead snout interfaced to guide shielding, and shielding around the Optical System, Sample Chamber and Detector Chamber Systems. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. The neutron dose rates around the Pelican instrument were reduced to < 0.5 μSv/h and the gamma dose rates were lowered to a safe working level of ≤ 3 μSv/h. The Optical System, with its components and the Detector Chamber were housed in borated polyethylene to shield from background signal noise.

Finding a suitable shield for mixed neutron and photon fields based on an Am–Be source

This study aimed to investigate a shielding design against neutron and photon rays from neutron irradiators based on Am–Be sources, using the Monte Carlo simulation. Different shielding materials were studied, including borated polyethylene, DaGa concrete, and epoxy resin with colemanite. The Monte Carlo N-particle code (MCNP) was used to design shielding. A new type of neutron and photon shielding material based on 40 % galena, 55 % polyethylene, and 5 % boric acid is proposed. The results show that the total effective dose of radiation is significantly reduced by the optimum radius of this shielding system.

SU-E-T-267: Construction and Evaluation of a Neutron Wall to Shield a 15 MV Linac in a Low-Energy Vault.

To design and quantify the shielding efficacy of an inner Borated Polyethylene (BPE)wall for a 15 MV linac in a low energy vault. A Varian TrueBeam linac with a maximum photon energy of 15 MV was installed in asmaller, preexisting vault. This vault originally housed a low-energy machine and did not havesufficient maze length recommended for neutron attenuation. Effective dose rate calculationswere performed using the Modified Kersey's Method as detailed in NCRP Report No. 151 andfound to be unacceptably high. An initial survey following the machine installation confirmedthese calculations. Rather than restrict the linac beam energy to 10 MV, BPE was investigatedas a neutron moderating addition. An inner wall and door were planned and constructed using4'×8'×1″ thick 5% BPE sheets. The resulting door and wall had 2″ of BPE; conduits and ductwork were also redesigned and shielded. A survey was conducted following construction of thewall. The vault modification reduced the expected effective dose at the vault door from 36.23to 0.010 mSv/week. As specific guidelines for vault modification are lacking, this project quantitativelydemonstrates the potential use of BPE for vault modification. Such modifications may provide alow-cost shielding solution to allow for the use of high energy modes in smaller treatment vaults.

The neutron dose equivalent evaluation and shielding at the maze entrance of a Varian Clinac 23EX treatment room

To evaluate the neutron and photon dose equivalent rate (H(n,D) and H(G)) at the outer maze entrance and the adjacent treatment console area after the installation of a Varian Clinac 23EX accelerator with a higher beam energy than its predecessor. The evaluation was based on measurements and comparison with several empirical calculations. The effectiveness of borated polyethylene (BPE) boards, as a maze wall lining material, on neutron dose and photon dose reduction is also reported. A single energy Varian 6 MV photon linear accelerator (linac) was replaced with a Varian Clinac 23EX accelerator capable of producing 18 MV photons in a vault originally designed for the former accelerator. In order to evaluate and redesign the shielding of the vault, the neutron dose equivalent H(n,D) was measured using an Andersson-Braun neutron Rem meter and the photon dose equivalent HG was measured using a Geiger Müller and an ion chamber gamma-ray survey meter at the outer maze entrance. The measurement data were compared to semiempirical calculations such as the Kersey method, the modified Kersey method, and a newly proposed method by Falcão et al. Additional measurements were taken after BPE boards were installed on the maze walls as a neutron absorption lining material. With the gantry head tilted close to the inner maze entrance and with the jaws closed, both neutron dose equivalent and photon dose equivalent reached their maximum. Compared to the measurement results, the Kersey method overestimates the neutron dose equivalent H(n,D) by about two to four times (calculation/measurement ratio approximately 2.4-3.8). Falcão's method largely overestimates the H(n,D) (calculation/measurement ratio approximately 3.9-5.5). The modified Kersey method has a calculation to measurement ratio about 0.6-0.9. The photon dose equivalent calculation including McGinley's capture gamma dose equivalent equation estimates about 77%-98% of the measurement. After applying BPE boards as a lining material on the inner corner of the maze wall, the H(n,D) and the H(G) at maze entrance were decreased by 41% and 59%, respectively. This work indicates that the Kersey method overestimates the neutron dose equivalent H(n,D) for a Varian Clinac 23EX accelerator. The Falcão method overestimates the H(n,D) partially due to the discrepancy in the International Commission on Radiological Protection (ICRP) conversion factors caused by the uncertainties of the estimated average neutron energy. The modified Kersey method gives the closest estimation of a Varian Clinac 23EX accelerator operated at 18 MV photon mode in a maze with a similar design as in the authors' study. However, it should be used with caution because of its tendency to underestimate the H(n,D). A borated polyethylene lining can provide a cost effective method to reduce neutron and photon dose equivalent at the maze door for an existing linac vault, following the installation of a higher energy linac.

Neutron‐induced electronic failures around a high‐energy linear accelerator

After a new in-vault CT-on-rails system repeatedly malfunctioned following use of a high-energy radiotherapy beam, we investigated the presence and impact of neutron radiation on this electronic system, as well as neutron shielding options. We first determined the CT scanner's failure rate as a function of the number of 18 MV monitor units (MUs) delivered. We then re-examined the failure rate with both 2.7-cm-thick and 7.6-cm-thick borated polyethylene (BPE) covering the linac head for neutron shielding. To further examine shielding options, as well as to explore which neutrons were relevant to the scanner failure, Monte Carlo simulations were used to calculate the neutron fluence and spectrum in the bore of the CT scanner. Simulations included BPE covering the CT scanner itself as well as covering the linac head. We found that the CT scanner had a 57% chance of failure after the delivery of 200 MUs. While the addition of neutron shielding to the accelerator head reduced this risk of failure, the benefit was minimal and even 7.6 cm of BPE was still associated with a 29% chance of failure after the delivery of 200 MU. This shielding benefit was achieved regardless of whether the linac head or CT scanner was shielded. Additionally, it was determined that fast neutrons were primarily responsible for the electronic failures. As illustrated by the CT-on-rails system in the current study, physicists should be aware that electronic systems may be highly sensitive to neutron radiation. Medical physicists should therefore monitor electronic systems that have not been evaluated for potential neutron sensitivity. This is particularly relevant as electronics are increasingly common in the therapy vault and newer electronic systems may exhibit increased sensitivity.

SU‐GG‐J‐75: Neutron‐Induced Electronic Failures around a High‐Energy Linear Accelerator

Purpose: After a new in‐vault CT‐on‐rails system repeatedly malfunctioned following use of a high‐energy radiotherapy beam, we investigated the presence and impact of neutron radiation on this electronic system, as well as potential neutron shielding and neutron evaluation options. Methods: We first determined the CT scanner's failure rate as a function of the number of 18 MV monitor units (MUs) delivered. We then reexamined the failure rate with both 2.7‐cm‐thick and 7.6‐cm‐thick borated polyethylene (BPE) covering the linac head for neutron shielding. To further examine shielding options, as well as to explore which neutrons were relevant to the scanner failure, Monte Carlo was used to calculate the neutron fluence and spectrum in the bore of the CT scanner. Simulations included BPE covering the CT scanner itself as well as covering the linac head. Results: We found that the CT scanner had a 57% chance of failure after the delivery of 200 MUs. While the addition of neutron shielding to the accelerator head reduced this risk of failure, the benefit was minimal and even 7.6 cm of BPE was still associated with a 29% chance of failure after the delivery of 200 MU. This shielding benefit was achieved regardless of whether the linac head or CT scanner was shielded. Additionally, it was determined that fast neutrons were primarily responsible for the electronic failures. Conclusions: As illustrated by the CT‐on‐rails system in the current study, physicists should be aware that electronic systems may be highly sensitive to neutron radiation. Physicists should therefore monitor electronic systems that have not been evaluated for potential neutron sensitivity. This is particularly relevant as electronics are increasingly common in the therapy vault and may exhibit increased sensitivity in newer electronic systems.

SU‐FF‐T‐271: Measurement Validation of Attenuation Values of Proton Facility Shielding Materials

Purpose: Strategies for design and shielding methods vary considerably with proton facilities. Material chosen for shielding is balance of effectiveness and cost. For space‐restricted environments, high‐density materials are desirable. Shielding calculations, whether analytical or statistical rely, on the neutron attenuations properties of chosen materials. The purpose of this study was to measure and calculate attenuation of various materials to be used for shielding. Methods and Materials: We acquired samples of the actual materials to be used to shield neutrons for our new facility. The original calculations relied on analytical methods and classically accepted values for neutron attenuation. Planned materials include: standard concrete (SC)∼136lb/ft3, high‐density concrete (HDC)∼250lb/ft3, Hematite mix (HM), steel (S), and borated polyethylene (BP). For the measurements, we used a 5Ci plutonium‐beryllium (239PuBe) neutron source (1″ in diameter and 5″ in length) with an energy spectrum ranging up to 11MeV. A Bonner‐sphere spectrometer was used for neutron measurements. The material samples were configured in a manner to allow measurements through various paths. A Fortran code was used to reconstruct the neutron spectrum and calculate the neutron dose equivalent. Results: Transmission curves were generated to yield TVL values, which were compared to those used in our analytical calculations. Measured TVL (cm) for the SC, HDC, HM, S, and BP were 55.9, 55.1, 63.1, 35.6, 60.6, respectively. The TVL (cm) used for the calculations were 85.6, 52.4, 32.9 respectively for SC, HDC, and S. Conclusions: Measured TVL values matched analytical calculations very closely, with the exception of standard concrete. The measurements performed in this study assisted in better understanding of attenuation properties of each material and increased our comfort level pre‐construction. As expenditures in the range of millions of dollars were about to be spent, the question of material choices, and the validity of the actual material mix were alleviated.

SU‐EE‐A1‐03: The Neutron Dose Evaluation and Shielding at the Maze Entrance of the Varian Clinac 2300EX

Purpose: After the installation of a Varian Clinac 2300EX accelerator, neutron and photon doses at the outer maze entrance were measured and compared with several empirical calculations. The measurements were taken before and after borated polyethylene (BPE) boards were installed on the maze wall as neutron absorption material. Method and Materials: The accelerator delivered an 18 MV photon beam at 600 cGy/min dose rate. An NRC NP‐2 type neutron REM‐meter and a Ludlum 14C survey meter with a 44‐6 type of probe were used to measure the neutron dose and photon dose, respectively. The measurements were taken at the center axis of the maze, 0.8 m above the floor, 0.3 m away from the maze door. Results: With the gantry head tilted close to the inner maze entrance with closed jaws, both neutron dose and photon dose reach their maximum. The measurement data were compared with empirical calculation results obtained by Kersey's method, modified Kersey's method and a newly proposed method by R.C. Falcão el al. The estimation from Kersey's method is about 2 – 4 times of the measurement (Ratio ≈ 2.4–3.8). Falcão's method has estimation about 22 times of the measurement (Ratio ≈ 21.9). The modified Kersey's method has the best prediction of the dose (Ratio ≈ 1.0). The McGinley's photon dose equation gives estimation about 80% of the measurement. After applying borated polyethylene boards as lining on maze wall, the neutron dose and the photon dose at maze entrance were decreased by 41% and 58%, respectively. Conclusion: This work indicates that Kersey's method overestimates the neutron dose by about 2 to 4 times for this study. Falcão's method largely overestimates the neutron dose. The modified Kersey's method is recommended to be used in neutron shielding calculation for the 2300EX accelerator.