Hasil untuk "Nuclear engineering. Atomic power"

Bin Wei, Yue Wang, Jincheng Wang et al.

Space lithium-cooled fast reactors have a high power density and are gradually becoming an ideal power source for deep space probes. However, in the case of launch failure, the reactor may fall into the water, leading to a supercritical accident with increased core reactivity. In this paper, a control ring design scheme and an axial gasket design scheme are proposed, and a spectral shift absorber design is added to the core to improve the submerged criticality safety of the core. The results show that both the control ring scheme and the axial gasket scheme can effectively absorb the excess thermal neutrons generated due to the dropout accident. When B4C is used as the neutron poison, the reactivity of the control ring scheme in a flooded accident is reduced by 9637 pcm compared to the traditional design. The proposed axial gasket design reduces the number of sliding reflector blocks to be slid out by 16.7 % in a submerged accident, which effectively improves the reactivity control efficiency. The simulation results of this paper provide a basis for the safety design and research of small space lithium-cooled fast reactors.

SUN Yiwen1, TIAN Lichao1, ZHANG Xiang2, YANG Xiaohu1, XU Weili2, HU Jiali2

In order to ensure the long-term, stable and safe operation of compact all-solid-state reactors, real-time monitoring of neutron fluxes within the reactor is essential. Neutron flux not only reflects the nuclear activity of the reactor, but is also directly related to the safety and efficiency of the reactor. Among the existing neutron detector technologies, there are some limitations in the application of 3He proportional counting tubes, fission gas ionization chambers, and traditional Si and Ge semiconductor detectors in high-temperature and strong radiation environments, such as the noise of gas detectors will increase significantly in high-temperature environments, which affects the detection accuracy. Although the traditional Si and Ge semiconductor detectors have good energy resolution in low temperature environments, the leakage current becomes larger in high temperature and strong irradiation environments, and the introduced noise will increase significantly, thus limiting their application in extreme environments. As a wide band-gap semiconductor material, SiC detector has the characteristics of large band gap, strong radiation resistance, high breakdown electric field strength and large saturated electron drift velocity, which makes it have excellent high temperature resistance and radiation resistance in extreme environments. The purpose of this study is to design a SiC neutron detector based on the 10B4C conversion layer and optimize its structural parameters to achieve high-precision real-time monitoring of neutron flux in the reactor. The energy spectrum of thermal neutrons in the sensitive volume of 10B4C coating with different thicknesses was simulated by Geant4 simulation software, and the relationship between the energy spectrum of γ rays of different energies in the sensitive volume of SiC and the variation of thermal neutron detection efficiency with the thickness of 10B4C coating under different thresholds were analyzed to determine the optimal thickness values of SiC epitaxial layer and 10B4C conversion layer. Combined with the dismantling of the conversion layer, the SiC detector was used to detect γ rays and thermal neutrons using 60Co, 137Cs, 22Na radioactive sources and neutron sources. By comparing the energy deposition of SiC for thermal neutron detection at different layer thicknesses, the accuracy of the simulation results was verified and experimental basis for subsequent applications was provided. The simulation and experimental results show that the optimal thickness of the conversion layer is 2 μm, 300 keV can be used as the screening γ ray threshold, the air gap has great influence on the energy deposition of nuclear reaction products, and the energy resolution of SiC can reach 2.08% when working at low pressure of 20 V. The overall results verify the feasibility of the design parameters of the SiC neutron detector, and realize the detection of neutron energy spectrum and flux.

Chansoo Choi, Bangho Shin, Yeon Soo Yeom et al.

The International Commission on Radiological Protection (ICRP) recently released the pediatric mesh-type reference computational phantoms (MRCPs) through ICRP Publication 156 to overcome the limitations of the pediatric voxel-type reference computational phantoms (VRCPs) of ICRP Publication 143. In the present study, these pediatric MRCPs were implemented into the Geant4 Monte Carlo code to produce a comprehensive dataset of dose coefficients for idealized external photon exposure geometries. This dataset includes 29 individual organ/tissue dose coefficients and effective dose coefficients, covering six irradiation geometries (i.e., AP, PA, LLAT, RLAT, ROT, and ISO) and 55 monoenergetic photon energies ranging from 10 keV to 10 GeV. The dataset comparison with the adult MRCPs of ICRP Publication 145 revealed substantial differences in effective dose coefficients throughout most energy ranges, primarily due to the age-related variations in organ/tissue depths, the differences reaching up to a factor ∼9. In addition, the comparisons with the pediatric VRCPs identified notable differences in effective dose coefficients, reaching up to a factor ∼4, for photon energies below 50 keV, due to the differences in phantom geometries and updates in simulation setups. For photon energies above 50 keV, deviations in effective dose coefficients between the pediatric MRCPs and VRCPs remained within 10 %.

HAO Sijia, DING Xijia, HU Zhen, XU Qian, HAN Zhijie

The performance of zirconium alloy cladding has always been a concern during reactor accidents. Ensuring the safe operation of high fuel consumption under reactivity initiated accident condition is also the core issue that scholars from various countries pay common attention to. The United States, France, Japan, Russia and other countries have carried out a large number of tests under reactivity initiated accident conditions. Some countries have also carried out the most advanced biaxial loading tests outside reactors, which has laid a solid foundation for these countries to occupy a leading position in the field of nuclear fuel safety. At present, there was no experimental study on pellet cladding mechanical interaction (PCMI) failure mode in Reactivity initiated accident (RIA) research in China. In order to establish our own safety criteria for reactivity initiated accident operating conditions, it is urgent to carry out domestic independent reactivity insertion accident test research, so as to provide data support for the commercial application of advanced fuel components. PCMI failure of cladding under reactivity insertion accident condition is a typical failure mode of high burnup fuel. Exploring effective test methods and obtaining test data to analyze the mechanism of pellet-cladding mechanical interaction is necessary. In this paper, out-pile simulated cladding failure was studied. Out-pile test device simulated PCMI was designed and domestic Zr-4 cladding was chosen as test material. High speed camera technology was used to assist. Hydrogen content absorbed in test material was used to simulate different high burnup cladding mechanical properties. The data of load, strain and fracture during the failure process of biaxial loading fuel element cladding were obtained. The macroscopic morphology and residual strain of test materials were analyzed after tests and stress state and maximum circumferential strain data of cladding tube under out-pile simulated reactivity insertion accident condition were obtained. The comparison with foreign research data verifies the feasibility and reliability of the PCMI failure test technology under out-pile simulated reactivity insertion accident condition. The results show that the circumferential strain of cladding tube decreases slightly with the increase of hydrogen content, and the damage degree of cladding is intensified. The circumferential strain of cladding tube increases obviously with the increase of temperature, and the lower limit of residual strain ratio tends to increase. The results of residual strain ratio are between -0.1 and 0.5, which accords with the range of residual strain ratio in reactor. The research results provide technical reserve and experience accumulation for future tests. It also provide reference data for the safety standards.

K. Krieger, M. Balden, A. Bortolon et al.

The efficacy of boron powder injection as a wall conditioning tool method in terms of its ability to create a sufficiently uniform boron layer on plasma-facing surfaces has been studied in ASDEX Upgrade. Boron powder was injected in two series of dedicated plasma discharges at varied injection rates and total amount injected. The resulting boron deposition was determined quantitatively by exposure of witness samples followed by ex-situ surface analysis of the retrieved samples. In both experiments, isotopically enriched boron was used to distinguish the deposition of the newly injected material from the residual boron fraction in the machine originating from previous glow discharge boronisations. It could be confirmed that the injected boron is migrating and re-deposited across plasma-facing wall surfaces already within one discharge. At erosion dominated divertor areas, the boron influx from the main chamber results in formation of a mixed tungsten–boron surface layer with a boron area density of O(1nm) whereas at deposition dominated areas closed boron layers grow with ongoing boron injection to a thickness of up to O(1μm). Extrapolating the radial boron deposition profile on the samples exposed in the main chamber to the limiter front yields a similar boron coverage of O(1μm), which is about ten times higher than typical values for glow discharge boronisation. Together with the observed reduction of oxygen level and improved wall pumping, the surface analysis results demonstrate that boron powder injection provides a suitable means to refresh the wall conditioning effect of a preceding glow discharge boronisation.

A. Fisli, E. Yulianti, B. Hanurajie et al.

This work studied protective material consisting of several metal oxide composites (Pb3O4, WO3, SnO2, and Bi2O3)-natural rubber (NR) for X-ray and gamma-ray shielding. The composites were prepared through open milling and vulcanization processes and further characterized by scanning electron microscope (SEM), X-ray diffraction (XRD), rheometry analysis, and density gauge. The attenuation coefficient of the sample was investigated using X-ray generators with voltages ranging from 50 to 140 kV and gamma-ray energies ranging and 356 to 1250 keV, respectively. The experimental results show that the linear attenuation coefficient of NR filled with metal oxides was significantly improved compared to pure NR. For gamma-ray 661 keV, the HVL of NR decreased from 9.0 cm to between 4.4 - 6.2 cm after it was filled with metal oxides. The Bi2O3-NR is the best suitable material for gamma-ray attenuation, followed by Pb3O4-NR, WO3-NR, and SnO2-NR. Meanwhile, for x-rays, the HVL of NR decreased from 2.0 cm to between 0.17 -0.31 cm after it was filled with metal oxides. The proposed metal oxide-NR composites can be appropriate as a flexible protective material for manufacturing wearable radiation shielding products such as gloves, aprons, rubber underwear, and other wearable materials.

SUN Yifan;BAI Liang;SHUANG Xiaochuan;TIAN Wenbo;YOU Hongjun

With the development of aerospace avionics, there is an increasing demand for high performance chips to handle data processing and transmission requirements. In recent years, the SRAM�based FPGAs have been widely used in on�board data processing systems for their abundant performance and high flexibility. However, compared to the anti�fuse FPGAs, the SRAM�based FPGAs are more vulnerable to the single error events (SEE), especially to the single error upsets (SEU) in space. This disadvantage restricts the SRAM�based FPGAs to be applied in on�board systems with high reliability requirements. There are quite a lot of studies in this area, trying to improve the anti�SEU performance for the SRAM�based FPGAs. Most of the researches focus on the anti�SEU design for the configuration RAM inside the FPGAs, for soft errors occurred in the configuration RAM can lead to more serious functional problems than any other parts in the FPGAs. With the technical advances of on�chip error detection and correction methods, Xilinx has introduced a new type of soft error mitigation (SEM) method on its SRAM�based FPGA. The SEM uses internal scrubber architecture, which relies on the on�chip error correction code (ECC) and the cyclic redundancy check (CRC) calculation circuits to detect and repair soft errors occurred in the configuration RAM. Compared to the traditional external scrubber architecture, internal scrubber is much more efficient due to its faster on�chip datapath and calculation circuits, which leads to better soft error rate performance in orbit. However, the SEM core is still built up by the SEU�vulnerable logic and memory resources in the FPGA, which may lead to certain risk of soft error problems. As a result, it is necessary to develop a systematical method to deal with the possible failure of the SEM core with as low cost as possible. In this paper, an analysis on the necessity and possible problems of applying the SEM technique was provided and a solution to handle the SEM core’s failure by introducing an external monitor was proposed. The monitor was implemented in an anti�fuse FPGA to ensure reliability and would only reprogram the FPGA when SEM failure was detected, leaving most ordinary soft errors to be handled by SEM core efficiently. Thus, both goals of high efficiency and high reliability were achieved. The solution is applied on a XC7K410T FPGA and verified by the SEE tests with heavy ions.

M.B. Ajani, P.P. Maleka, S. Penabei et al.

The study evaluates the presence of heavy metals concentrations in the soil samples collected, which are important to minimise health risks resulting from exposure to such toxic substances. Ten samples collected in Yapala, located in Mayo-Dallah, Chad were analysed using Inductively Coupled Plasma-Mass Spectrometer (ICP-MS). ICP-MS analysis revealed the presence of various toxic elements, Co (1–35 mg/kg), Cu (3–225 mg/kg), As (0.41–6.09 mg/kg), Se (0.04–0.44 mg/kg), Mo (0.15–15.50 mg/kg), Cd (0.004–1.896 mg/kg), Sn (0.29–8.89 mg/kg), Sb (below detectable limit-1.01 mg/kg), Hg (0.07–5.57 mg/kg), Pb (4–59 mg/kg), respectively. Elemental concentrations for Cu, Mo and Hg were higher than the maximum allowable limit in some of the analysed samples, while others were within the safe limit. The potential health risk showed the influence of the three exposure pathways were in the order of dermal contact > ingestion > inhalation. The Hazard Quotients (HQ) and Hazard index (HI) values for the heavy metal samples used were higher than 1 in dermal contact, which indicates that the individual may be subjected to potentially higher health risk. In accordance with the Food and Agriculture Organization (FAO)/World Health Organization (WHO) guidelines, our results show that some of the samples analysed possessed increase in the risk of exposure to heavy metals, and therefore the possibility of developing diseases by people inhabiting in this environment. As a result, regular monitoring of this area is needed, including further investigations to surrounding regions to provide a comprehensive assessment and decision.

Amira M. Gamal-Eldeen, Amani A. Alrehaili, Afaf Alharthi et al.

Perfluorodecalin emulsion (Perftoran®) is an oxygen carrier. Verteporfin, a photosensitizer, is used for cancer photodynamic therapy (PDT) in its liposomal form (Visudyne®). Hypoxia affects PDT efficacy. The study aimed to investigate the effect of Perftoran® on the efficacy of Visudyne®/PDT and on the hypoxia pathway in lung cancer. Lung cancer was induced in mice by diethylnitrosamine and carbon tetrachloride. Lung tissues have been analyzed for apoptosis and necrosis detection by dual DNA staining; mitochondrial transmembrane potential (ΔΨm) staining by Red CMX-Ros, hypoxia estimation by pimonidazole, miRNA-21-5p and miRNA-210-3p evaluation by qPCR, and HIF-1α and VEGF determination by ELISA. Compared to Visudyne®/PDT, Perftoran®/Visudyne®/PDT led to severe apoptosis and necrosis regions in the bronchiolar epithelium, induced late apoptosis and necrosis and suppressed ΔΨm and total hypoxia level more than Visudyne®/PDT. The expression of miR-21 and miR-210 and the concentration of HIF-1α and VEGF were significantly induced in tumor-bearing mice. Compared to Visudyne®/PDT, Perftoran®/Visudyne®/PDT dramatically suppressed the expression of miR-21 and miR-210 and the concentration of HIF-1α and VEGF. In conclusion, these findings indicate the potential role of Perftoran® in suppressing hypoxia through the inhibition of the hypoxamiRs expression and HIF-1α, which in turn led to the inhibition of VEGF, during Visudyne®/PDT.

XIAO Yue;WU Haicheng;WU Xiaofei;ZHANG Huanyu

The development of high precision burnup analysis requires high accuracy burnup database. The benchmarking method of burnup database is very important for developing high accuracy burnup database. Based on the burnup database composed of the multitemperature continuous point section database CENACE13 made by China Nuclear Data Center and Cinder90 burnup database, the modeling of SF95 samples in the post irradiation experiment of TAKAHAMA3 power water reactor was taken as an example. The spent fuel composition database SFCOMPO20 gives the detailed information of postirradiation experiment of TAKAHAMA3 power water reactor, including component information, power history, boron concentration history and nuclides inventory value. The influence of modeling elements on burnup calculation was studied, and the modeling method of burnup benchmark experiment was determined. The influence of modeling elements such as material temperature and density, boron concentration and burnup step on modeling calculation was determined. As an important part of burnup credit, burnup analysis aims to obtain the nuclide composition and inventory of spent fuel. Using the burnup benchmark experimental modeling method, the burnup credit was studied and the comparison of calculated values and experimental values of interested actinide nuclides and fission products nuclides was carried out. The preliminary results showed that the deviation between the calculated values and the experimental values of main actinide nuclides is less than 2%, the relative deviation between the calculated values and the experimental values of most minor actinides nuclides is less than 10%, and the deviation between the calculated values and the experimental values of most important fission products nuclides is less than 5%. In this paper, based on the adjacent burnup chain of 125Sb, the variation law of 125Sb inventory with burnup depth was theoretically analyzed, and it is confirmed that there are defects in the measurement results of destructive radiochemical experiment. In this study, the destructive experimental value of 125Sb was modified by nondestructive experimental measurement value and experimental value modification formula. And the corrected experimental values of 125Sb inventory were obtained, which reduced the calculation deviation between calculated value and experimental value from nearly 170% to less than 20%. This study shows that the benchmarking of burnup database using the spent fuel composition database requires not only the appropriate burnup benchmark experimental modeling method, but also the appropriate evaluation and verification of inventory measurement data, so as to be able to be used for benchmarking of burnup database and burnup credit research.

Myeong-Gie Kang

This study investigates the effect of an included angle and heat flux on heat transfer of V-shape tube array having a horizontal upper tube. The test uses two stainless steel tubes with a smooth surface submerged under the water at atmospheric pressure. The angle varies from 2° to 24°. The heat transfer coefficient gets decreasing in consequence as the angle increases. The enhancement due to the lower tube is distinct as the heat flux is lower than 60 kW/m2, where the effect of the convective flow is dominant. The present study and the published results show a similar tendency. Although the heat transfer coefficient for the present study is smaller than the symmetry case, enhanced heat transfer is observed compared to the tube array having a lower horizontal tube as the included angle is less than 10°. Keywords: Pool boiling, Tube array, V-shape tube, Advanced reactors, Included angle

Toshinori Yamada

Masao Tomonaga

Seventy-four years have passed since the atomic bombings of Hiroshima and Nagasaki. Approximately 210,000 victims died, and another 210,000 people survived. The damage to their health has continued, consisting of three phases of late effects: the appearance of leukemia, the first malignant disease, in 1949; an intermediate phase entailing the development of many types of cancer; and a final phase of lifelong cancers for hibakusha who experienced the bombing as a child, as well as a second wave of leukemia for elderly hibakusha and psychological damage such as depression and post-traumatic stress disorder. Thus, the human consequences of the atomic bombings have not ceased; many people are still dying of radiation-induced malignant diseases. Therefore, it is too early to finalize the total death toll. Hibakusha have faced a never-ending struggle to regenerate their lives and families under the fear of disease. As the only group of Homo sapiens experiencing real nuclear attacks, hibakusha have continued to engage in a lifelong movement to eliminate nuclear weapons. Political leaders, especially of nuclear-weapon states, must learn the wisdom of the hibakusha to save Homo sapiens from possible global extinction by nuclear war.

M. Cavallaro, C. Agodi, F. Cappuzzello et al.

Heavy-ion induced two-neutron transfer reactions (18O,16O) at 84 MeV were studied on several targets up to high excitation energy of the residual nucleus thanks to the use of the MAGNEX magnetic spectrometer to detect the ejectiles. The obtained results indicate of the important role played by the nuclear paring.

Zhaoming Meng, Wei Zhang, Jiazhi Liu et al.

Steam jet condensation is of great importance to pressure suppression containment and automatic depressurization system in nuclear power plant. In this paper, the condensation processes of sonic steam jet in a quiescent subcooled pool are recorded and analyzed, more precise understanding are got in direct contact condensation. Experiments are conducted at atmospheric pressure, and the steam is injected into the subcooled water pool through a vertical nozzle with the inner diameter of 10 mm, water temperature in the range of 25–60 °C and mass velocity in the range of 320–1080 kg/m2s. Richardson number is calculated based on the conservation of momentum for single water jet and its values are in the range of 0.16–2.67. There is no thermal stratification observed in the water pool. Four condensation regimes are observed, including condensation oscillation, contraction, expansion-contraction and double expansion-contraction shapes. A condensation regime map is present based on steam mass velocity and water temperature. The dimensionless steam plume length increase with the increase of steam mass velocity and water temperature, and its values are in the range of 1.4–9.0. Condensation heat transfer coefficient decreases with the increase of steam mass velocity and water temperature, and its values are in the range of 1.44–3.65 MW/m2°C. New more accurate semi-empirical correlations for prediction of the dimensionless steam plume length and condensation heat transfer coefficient are proposed respectively. The discrepancy of predicted plume length is within ± 10% for present experimental results and ± 25% for previous researchers. The discrepancy of predicted condensation heat transfer coefficient is with ± 12%. Keywords: Direct contact condensation, Sonic jet, Condensation regimes, Condensation heat transfer coefficients

A.V. Antonov, V.A. Chepurko

There are three stages in the process of operation of technical equipment, each with a specific trend of the failure flow parameter (FFP) behavior. During normal operation, the FFP value is approximately constant. In this case, the equipment operation process is presumably time-homogeneous, and the reliability indicators are calculated by classical methods. The FFP decreases with time in the burn-in period and it increases at the stage of aging. This means that the operating times between two successive failures at the burn-in and aging stages are not identically distributed random quantities, and the flow of failures cannot be looked upon as recurrent. Calculation of reliability characteristics shall take into account that the failure flow is time-non-homogeneous. The paper describes a method to estimate the nuclear power plant (NPP) equipment reliability indicators that makes it possible to take into account the potential non-homogeneity of the failure flow. Peculiarities of obtained statistical data on failures are identified. Application of a normalizing flow function model to calculate the required reliability indicators is described. A practical example of an analysis of data on the CPS KNK-56 component failures at Bilibino NPP is provided. Presentation of the statistical data analysis procedure described in [1] is continued.

D'Auria Francesco, Glaeser Horst, Debrecin Nenad

A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On the one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs). Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry). The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty approach.

A.S. Bochkarev, P.N. Alekseev, A.S. Korsun et al.

The paper discusses a set of developed integrated one-dimensional models of thermal-hydraulic processes that contribute to the removal of decay heat in a BN-type reactor. The assumptions and constraints involved in one-dimensional equations of unsteady natural convection in closed circuits have been analyzed. It has been shown that the calculated values of the primary circuit sodium temperature and flow rate in conditions with a loss of heat sink and with a forced circulation of the primary coolant are in a reasonable agreement with the results of a benchmark experiment in the PHENIX reactor. The model makes it possible to assess the effects general thermophysical and geometrical parameters and the selected technology have on the efficiency of passive heat removal by the natural coolant convection in the reactor tank and in the emergency heat removal system's intermediate circuit and by the heat transfer through the reactor vessel. The model is a part of an integrated algorithm used to assess the inherent safety level of advanced fast neutron reactors and is intended primarily to develop, at the early conceptual design stage, the recommendations and requirements with respect to the reactor equipment parameters leading to an increase in the reactor inherent safety. The model will be used to identify the set of quantitative thermal-hydraulic criteria that have an effect on the dynamics of emergency transients leading to a potential loss of integrity by the reactor safety barriers, and to formulate such limits for the defined criteria as would cause, if observed, the requirement for the safety barrier integrity to be met under any combination of the accident initiating events.

JONG WOON KIM, SER GI HONG, YOUNG-OUK LEE

Scattering source calculations using conventional spherical harmonic expansion may require lots of computation time to treat full-coupled three-dimensional photon-electron transport in a highly anisotropic scattering medium where their scattering cross sections should be expanded with very high order (e.g., P7 or higher) Legendre expansions. In this paper, we introduce a modified scattering kernel approach to avoid the unnecessarily repeated calculations involved with the scattering source calculation, and used it with parallel computing to effectively reduce the computation time. Its computational efficiency was tested for three-dimensional full-coupled photon-electron transport problems using our computer program which solves the multi-group discrete ordinates transport equation by using the discontinuous finite element method with unstructured tetrahedral meshes for complicated geometrical problems. The numerical tests show that we can improve speed up to 17∼42 times for the elapsed time per iteration using the modified scattering kernel, not only in the single CPU calculation but also in the parallel computing with several CPUs.

HYEOK-JUNG KWON, HAN-SUNG KIM, KYUNG-TAE SEOL et al.

The beam position monitor (BPM) is an essential component for the PEFP 100-MeV linac's commissioning. A prototype stripline-type linac BPM was designed for this purpose. The electrode aperture is 20 mm in diameter, and the electrode is 25 mm long, so it can be installed between Drift Tube Linac (DTL)101 and DTL102, which is the shortest distance. One end of the electrode is connected to the Sub Miniature Type A (SMA) feed through for signal measurement, and the other end is terminated as a short. The signal amplitude of the fundamental component was calculated and compared with that of the second harmonic component. The designed BPM was fabricated and a low-power RF test was conducted. In this paper, the design, fabrication and low power test of the BPM for the PEFP linac are presented.