Fission Neutron Irradiation Research Articles

Challenge to Performance Evaluation of Fusion Reactor Materials by Fission Neutron and Charged Particle Irradiation

Since there are no fusion reactors generating high-flux 14 MeV neutrons, it is necessary to evaluate materials’ performance in fusion reactors based on a correlation of fission neutron and charged particle irradiations. However, the irradiation tests involve various issues which prevent simple correlation and evaluation. In this paper, the issues related to irradiation temperature control and dose rate effects are pointed out and analyzed, and proposals regarding future irradiation tests are given.

Potassium-39-derived 36Ar production during fission-neutron irradiation and its effect on 40Ar/39Ar ages

Various interference reactions producing unwanted Ar isotopes from K, Ca, Cl and Ar require correction to satisfy the 40Ar/39Ar age equation. Using GEANT4, we design and build a model Cadmium Lined In Core Irradiation Tube (CLICIT) irradiation facility, as used in the Oregon State TRIGA Reactor (OSTR). We illustrate the complexity of the irradiation of geologic samples within this framework and determine an overlooked production channel of 36Ar. The production of 36Ar is fed from the 39K(n,α)36Cl nuclear channel, 36Cl subsequently decays to 36Ar (39K(n,α,β) 36Ar). Simulations in this work using a 235U fission neutron energy spectrum and modelled CLICIT facility, determine a production ratio for this reaction (36Cl/39Ar)K = 0.40 ± 0.01 (1σ); greater than an order of magnitude larger than any other K interference. The magnitude of the resulting age bias for an unknown sample will be a function of the integrated neutron flux, the length of irradiation (fluence), the time elapsed since irradiation, and the age relationship between the unknown and neutron fluence monitor. We show using the raw data of (Niespolo et al., 2017) that the age of Alder Creek sanidine can be modified to be ca. 0.1% older (1σ), at the 2σ level of current analytical precision for the Alder Creek age for this study. The 39K(n,α,β)36Ar inference should be incorporated into routine data analysis and may be especially important in the intercalibration of the 40Ar/39Ar system with other chronometers (e.g., 206Pb/238U).

The neglected activation of tantalum in reduced activation materials

Due to the expression of “reduced activation”, the severe radioactivity of tantalum (Ta) after neutron irradiation in early decay years is easily neglected. In the presented study, after an irradiation of only 0.3 dpa by fission neutron and followed by a cooling stage of 305 days, the RAFM steel specimen with a weight of 0.8 g possessed drastic radioactivity of 4 mSv/hour at a measuring distance of 20 cm. It was ascertained that the high radioactivity was generated by 182Ta. The severe radioactivity seriously increased the danger of material experiments, even though the original Ta addition of the steel was only 0.13 wt%. Similar deteriorative radioactivity of Ta can be generated in both fission and fusion neutron irradiations. We appeal to seriously deliberate the radioactivity of Ta and control the Ta content in the RAFM steel.

Probabilistic failure assessment of an irradiated DEMO breeding blanket component under ferromagnetic loads: Impacts of material data scattering and uncertainty in local stress intensities

The most significant technical challenge in developing structural materials for in-vessel components in the Demonstration Power Plant (DEMO) fusion reactor is to qualify the lifetime of material to guarantee the lifetime of components. The qualification needs to be conducted theoretically rather than empirically because it is conducted based on the knowledge and data acquired in fission neutron irradiation experiments and various simulation irradiation experiments rather than in experiments performed in the actual fusion environment. Such experiments will not be achievable until the DEMO fusion reactor becomes operational. Consequently, various uncertainties are expected in irradiation data and loading conditions.The probabilistic approach, where the probability of failure is calculated based on the probability density function of postulated load distribution and material property distribution, is one path to mitigate such uncertainties. In this approach, those probabilities constitute the subjective probability based on the theoretical understanding of fusion neutron irradiation effects on materials to maximize the confidence in the estimated value.A case study of the Japanese DEMO breeding blanket is the basis for the discussion of the technical issues involved in the probabilistic approach. The breeding blanket design was simplified, the failure probability was estimated, and the impact of electromagnetic force on ferromagnetic material caused by a strong magnetic field induced by toroidal field coils was considered. The discussion of the impact of the irradiation damage is based on the irradiation data for F82H reduced-activation ferritic/martensitic steel. A direction for the development of fusion structural materials is suggested in terms of design technology and the statistical quality of the irradiation database.

A fusion relevant data-driven engineering void swelling model for 9Cr tempered martensitic steels

The UCSB database on cavity evolution in 9-12Cr tempered martensitic steels (TMS), includes the results for both dual heavy and helium ion (DII), and High Flux Isotope Reactor (HFIR) in situ helium injection (ISHI) neutron irradiations at 500°C. These results were combined with literature single ion and fission neutron irradiation data to derive a model for the void volume fraction, fv, as a function of displacements per atom (dpa) and transmutant helium concentrations in atomic parts per million (appm). The scientific foundation for the paper is described in a companion paper entitled “Cavity Evolution and Void Swelling in Dual Ion Irradiated Tempered Martensitic Steels”. Here, we show that fv(dpa, He/dpa) is described by the incubation dose, dpai, for the onset of void growth, and the post-incubation growth rate, fv’(%/dpa). Both dpai and fv’ decrease with increasing He/dpa at > ∼ 5. The dpai is also lower for the ISHI neutron irradiations at the same He/dpa. Single heavy ion and fission reactor neutron irradiations, with low He/dpa ratios, have a much larger dpai. Based on a combined analysis of DII, single ion, ISHI and fission neutron data, we further show that the post-incubation fv data analyzed here have a common empirical curve shape, with fv’ reaching up to ∼ 0.2%/dpa at very high dpa. We also show that fv’ can be predicted based on a physical model of defect partitioning between evolving sinks. At 500°C and fusion relevant He/dpa ≈ 10, the best-fit model predicts nominal swelling, S = fv/(1-fv), of ∼ 1.1, 4.9 and 16% at 50, 100 and 200 dpa, respectively. The physically motivated, data-driven model includes estimated uncertainties for both dpai and fv’.

Simulated research on displacement damage of gallium nitride radiated by different neutron sources

Gallium nitride (GaN), one of the third-generation wide-bandgap semiconductors, offers significant application for advanced electronic devices utilized in neutron irradiation environments, like the defense, space, and aerospace, etc. In these applications, neutron irradiation-induced defects affect the properties of GaN and eventually degrade the performance of devices. In this work, neutron transport process in GaN is simulated by using the Monte Carlo-based code, Geant4 toolkit under four different irradiation conditions, e.g. high flux isotope reactor, high temperature gas-cooled reactor, pressurized water reactor, and atmospheric neutron irradiation. The energy spectra of primary knock-on atoms (PKA) in GaN and the corresponding weighted spectra under those irradiation conditions are analyzed. It is found that there is one unusual “peak” at around 0.58 MeV in the Primary recoil spectrum, regardless of the irradiation conditions. This peak is attributed to the neutron reaction of hydrogen nucleus, i.e., (n, p). Because of the remarkable (n,p) reaction cross-section of low-energy neutron, the intensity of this peak is related to the ratio of low-energy neutron to the total neutron spectrum. By comparing these PKA energy spectra in GaN, we can see that the PKA energy spectrum created under atmospheric neutron irradiation is similar to that in the high flux isotopic reactor. Specifically, the energy distribution of PKA is wide, and the magnitude of energy is lower than those under fission neutron irradiation conditions. In combination with the effects of nuclear reaction products on electrical properties, the high flux isotopic reactor is more suitable for simulating the irradiation of GaN in an atmospheric neutron energy spectrum environment. These above results can provide not only some insights into the evaluation of the degradation of GaN-based electronic devices under neutron irradiation, but also dataset for the study of radiation damage effect of GaN in simulated neutron environment.

Neutron irradiation effects on the mechanical properties of powder metallurgical processed tungsten alloys

Neutron irradiation effects on the tensile properties of Pure W, K-doped W, W–3%Re, and K-doped W–3%Re were examined under the US-Japan collaboration project PHENIX. The fission neutron irradiation experiments were carried out up to 0.74 dpa at 600 °C and 800 °C. Pure W (SR) showed irradiation hardening and loss of ductility after irradiation at 600 °C and 800 °C. K-doped W–3%Re (SR) also exhibited irradiation hardening but was ductile after irradiation. Characteristic points of the K-doped W–3%Re (SR) are small grain size and layered structure. The stress-relief treatment and the layered structure may improve the ductility of powder-metallurgy W alloys after neutron irradiation, so a combination of K-doping and Re addition would be beneficial to improve irradiation resistance.

Formation energy and diffusion barrier of point defects in three uranium compounds as nuclear fuels

During fission neutron irradiation, point defects are generated in uranium fuel, and the point defects inevitably have some effects on the properties of the uranium fuel. In particular, the point defects accumulate through diffusion, resulting in formation of body defects that degrade the properties of the fuel. Using the first principles calculation method, based on the density functional theory, we calculated the elastic properties of three kinds of uranium compounds (UO2, UN, and UC). The results are in agreement with calculated and experimental results of others. On the basis of verifying the accuracy of the calculation model, we investigated formation energy and diffusion barrier of three kinds of point defects (vacancy, self-interstitial atom, and hydrogen/helium (H/He) atoms) in the three kinds of uranium compounds. The formation energy of the U vacancy is lower than that of the O, N, and C vacancy in the three kinds of uranium compounds, whereas the interstitial U atom has a higher formation energy than the interstitial O, N, and C atoms. The formation energy of the interstitial H atom is much lower than that of the interstitial He atom in the three kinds of uranium compounds, and the H atom prefers to occupy the O, N, and C vacancy while the interstitial He atom prefers to occupy the U vacancy. The diffusion of the interstitial H atom in UO2 is easier than that of the interstitial He atom, whereas the diffusion of the interstitial He atom in UN and UC is easier than that of the interstitial H atom. The diffusion of the interstitial self-atom in UO2 is more difficult than that of the self-atom in UN and UC. The work provides a fundamental understanding of the point defects formation and diffusion to develop uranium fuel by controlling defect formation.

Development of benchmark reduced activation ferritic/martensitic steels for fusion energy applications

Reduced-activation ferritic/martensitic (RAFM) steel is the benchmark structural material for in-vessel components of fusion reactor. The current status of RAFM developments and evaluations is reviewed based on two leading RAFM steels, F82H and EUROFER-97. The applicability of various joining technologies for fabrication of fusion first wall and blanket structures, such as weld or diffusion bonding, is overviewed as well. The technical challenges and potential risks of utilizing RAFM steels as the structural material of in-vessel components are discussed, and possible mitigation methodology is introduced. The discussion suggests that deuterium–tritium fusion neutron irradiation effects currently need to be treated as an ambiguity factor which could be incorporated within the safety factor. The safety factor will be defined by the engineering design criteria which are not yet developed with regard to irradiation effects and some high temperature process, and the operating time condition of the in-vessel component will be defined by the condition at which those ambiguities due to neutron irradiation become too large to be acceptable, or by the critical condition at which 14 MeV fusion neutron irradiation effects is expected to become different from fission neutron irradiation effects.

Towards a programme of testing and qualification for structural and plasma-facing materials in ‘fusion neutron’ environments

Materials damage by 14.1MeV neutrons from deuterium–tritium (D–T) fusion reactions can only be characterised definitively by subjecting a relevant configuration of test materials to high-intensity ‘fusion-neutron spectrum sources’, i.e. those simulating closely D–T fusion-neutron spectra. This provides major challenges to programmes to design and construct a demonstration fusion reactor prior to having a large-scale, high-intensity source of such neutrons. In this paper, we discuss the different aspects related to these ‘relevant configuration’ tests, including: • generic issues in materials qualification/validation, comparing safety requirements against those of investment protection; • lessons learned from the fission programme, enabling a reduced fusion materials testing programme; • the use and limitations of presently available possible irradiation sources to optimise a fusion neutron testing program including fission-neutron irradiation of isotopically and chemically tailored steels, ion damage by high-energy helium ions and self-ion beams, or irradiation studies with neutron sources of non-fusion spectra; and • the different potential sources of simulated fusion neutron spectra and the choice using stripping reactions from deuterium-beam ions incident on light-element targets.

Microstructural changes of Y-doped V-4Cr-4Ti alloys after ion and neutron irradiation

High-purity Y-doped V-4Cr-4Ti alloys (0.1–0.2wt. % Y), manufactured by the National Institute for Fusion Science (NIFS), were used for this study. Heavy-ion and fission-neutron irradiation was carried out at temperatures 673–873K. During the ion irradiation at 873K, the microstructure was controlled by the formation of Ti(C,O,N) precipitates lying on the (100) plane. Y addition effectively suppressed the growth of Ti(C,O,N) precipitates, especially at lower dose irradiation to up to 4 dpa. However, at higher dose levels (12.0 dpa), the number density was almost at the same levels irrespective of the presence of Y. After neutron irradiation at 873K, fine titanium oxides were also observed in all V alloys. However, smaller oxide sizes were observed in the Y-doped samples under the same irradiation conditions. The detailed analysis of EDS showed that the center of the Ti(C,O,N) precipitates was mainly enriched by nitrogen. The results showed that the contribution of not only oxygen atoms picked up from the irradiation environment but also nitrogen atoms is essential to understand the microstructural evolution of V-4Cr-4Ti-Y alloys.

Effect of neutron energy and fluence on deuterium retention behaviour in neutron irradiated tungsten

Deuterium (D) retention behaviours for 14 MeV neutron irradiated tungsten (W) and fission neutron irradiated W were evaluated by thermal desorption spectroscopy (TDS) to elucidate the correlation between D retention and defect formation by different energy distributions of neutrons in W at the initial stage of fusion reactor operation. These results were compared with that for Fe2+ irradiated W with various damage concentrations. Although dense vacancies and voids within the shallow region near the surface were introduced by Fe2+ irradiation, single vacancies with low concentration were distributed throughout the sample for 14 MeV neutron irradiated W. Only the dislocation loops were introduced by fission neutron irradiation at low neutron fluence. The desorption peak of D for fission neutron irradiated W was concentrated at low temperature region less than 550 K, but that for 14 MeV neutron irradiated W was extended toward the higher temperature side due to D trapping by vacancies. It can be said that the neutron energy distribution could have a large impact on irradiation defect formation and the D retention behaviour.

Prompt and delayed inelastic scattering reactions from fission neutron irradiation—first results of FaNGaS

A new instrument for fast neutron gamma spectroscopy has been installed and tested at the Forschungsneutronenquelle Heinz Maier-Leibnitz of the MLZ in Garching. A beam with the flux of about 108 cm−2 s−1 fission neutrons is available to irradiate small samples to study inelastic neutron scattering reactions in materials. A number of relevant elements have been investigated and new gamma lines were identified in comparison to existing data libraries. The method seems to be linear with respect to measurement time and sample mass. A particular advantage of fast neutron PGAA compared to thermal neutron PGAA appears in the analysis of materials with high thermal capture cross sections.

Protecting intestinal epithelial cell number 6 against fission neutron irradiation through NF-κB signaling pathway.

The purpose of this paper is to explore the change of NF-κB signaling pathway in intestinal epithelial cell induced by fission neutron irradiation and the influence of the PI3K/Akt pathway inhibitor LY294002. Three groups of IEC-6 cell lines were given: control group, neutron irradiation of 4Gy group, and neutron irradiation of 4Gy with LY294002 treatment group. Except the control group, the other groups were irradiated by neutron of 4Gy. LY294002 was given before 24 hours of neutron irradiation. At 6 h and 24 h after neutron irradiation, the morphologic changes, proliferation ability, apoptosis, and necrosis rates of the IEC-6 cell lines were assayed and the changes of NF-κB and PI3K/Akt pathway were detected. At 6 h and 24 h after neutron irradiation of 4Gy, the proliferation ability of the IEC-6 cells decreased and lots of apoptotic and necrotic cells were found. The injuries in LY294002 treatment and neutron irradiation group were more serious than those in control and neutron irradiation groups. The results suggest that IEC-6 cells were obviously damaged and induced serious apoptosis and necrosis by neutron irradiation of 4Gy; the NF-κB signaling pathway in IEC-6 was activated by neutron irradiation which could protect IEC-6 against injury by neutron irradiation; LY294002 could inhibit the activity of IEC-6 cells.

Low-Flux Neutron Radiation Detection Technology with High Sensitivity

Boron-containing plastic scintillator detectors have a high detection efficiency for low-intensity thermal neutrons and fast neutrons which is currently the preferred types of neutron detector. This article is based on Monte Carlo method, studied boron-containing plastic scintillator for neutron detection performance, and analysis the energy deposition flux characteristics and detection efficiency when low intensity fission neutron incident to the boron plastic scintillator. We obtain the low-flux neutron detector performance in a variety of neutron source energy, boron-containing plastic scintillator diameter and length. Results showed that, when the boron-containing plastic scintillator lengths increase, the energy deposition flux will increase. When the length and diameter is constant, increasing source strength can increase the energy deposition flux brought by the recoil proton to a certain extent. When the source intensity over after thermal neutrons, due to the decrease of the cross section, the energy deposition fluxes brought by the react of neutrons and will decrease. The results provide help for low intensity fission neutron radiation detection technology with high sensitivity.

Evaluation of high strength, high conductivity CuNiBe alloys for fusion energy applications

The unirradiated tensile properties for several different heats and thermomechanical treatment conditions of precipitation strengthened Hycon 3HP™ CuNiBe (Cu–2%Ni–0.35%Be in wt.%) have been measured over the temperature range of 20–500°C for longitudinal and long transverse orientations. The room temperature electrical conductivity has also been measured for several heats, and the precipitate microstructure was characterized using transmission electron microscopy. The CuNiBe alloys exhibit very good combination of strength and conductivity at room temperature, with yield strengths of 630–725MPa and electrical conductivities of 65–72% International Annealed Copper Standard (IACS). The strength remained relatively high at all test temperatures, with yield strengths of 420–520MPa at 500°C. However, low levels of ductility (<5% uniform elongation) were observed at test temperatures above 200–250°C, due to flow localization near grain boundaries (exacerbated by having only 10–20 grains across the gage thickness of the miniaturized sheet tensile specimens). Scanning electron microscopy observation of the fracture surfaces found a transition from ductile transgranular to ductile intergranular fracture with increasing test temperature. Fission neutron irradiation to a dose of ∼0.7 displacements per atom (dpa) at temperatures between 100 and 240°C produced a slight increase in strength and a significant decrease in ductility. The measured tensile elongation after irradiation increased with increasing irradiation temperature, with a uniform elongation of ∼3.3% observed at 240°C. The electrical conductivity decreased slightly following irradiation, due to the presence of defect clusters and Ni, Zn, Co transmutation products. Considering also previously published fracture toughness data, this indicates that CuNiBe alloys have irradiated tensile and electrical properties comparable or superior to CuCrZr and oxide dispersion strengthened copper at temperatures <250°C, and may be an attractive candidate for certain low-temperature fusion energy structural applications. Conversely, CuNiBe may not be preferred at intermediate temperatures of 250–500°C due to the poor ductility and fracture toughness of CuNiBe alloys at temperatures ⩾250°C. The potential deformation mechanisms responsible for the transition from transgranular to intergranular fracture are discussed. The possible implications for other precipitation-hardened alloys such as nickel based superalloys are briefly discussed.

Evaluation of irradiation facility options for fusion materials research and development

Successful development of fusion energy will require the design of high-performance structural materials that exhibit dimensional stability and good resistance to fusion neutron degradation of mechanical and physical properties. The high levels of gaseous (H, He) transmutation products associated with deuterium–tritium (D–T) fusion neutron transmutation reactions, along with displacement damage dose requirements up to 50–200displacements per atom (dpa) for a fusion demonstration reactor (DEMO), pose an extraordinary challenge. One or more intense neutron source(s) are needed to address two complementary missions: (1) scientific investigations of radiation degradation phenomena and microstructural evolution under fusion-relevant irradiation conditions (to provide the foundation for designing improved radiation resistant materials), and (2) engineering database development for design and licensing of next-step fusion energy machines such as a fusion DEMO.A wide variety of irradiation facilities have been proposed to investigate materials science phenomena and to test and qualify materials for a DEMO reactor. Some of the key technical considerations for selecting the most appropriate fusion materials irradiation source are summarized. Currently available and proposed facilities include fission reactors (including isotopic and spectral tailoring techniques to modify the rate of H and He production per dpa), dual- and triple-ion accelerator irradiation facilities that enable greatly accelerated irradiation studies with fusion-relevant H and He production rates per dpa within microscopic volumes, D–Li stripping reaction and spallation neutron sources, and plasma-based sources.The advantages and limitations of the main proposed fusion materials irradiation facility options are reviewed. Evaluation parameters include irradiation volume, potential for performing accelerated irradiation studies, capital and operating costs, similarity of neutron irradiation spectrum to fusion reactor conditions, temperature and irradiation flux stability/control, ability to perform multiple-effect tests (e.g., irradiation in the presence of a flowing coolant, or in the presence of complex applied stress fields), and technical maturity/risk of the concept. Ultimately, it is anticipated that heavy utilization of ion beam and fission neutron irradiation facilities along with sophisticated materials models, in addition to a dedicated fusion-relevant neutron irradiation facility, will be necessary to provide a comprehensive and cost-effective understanding of anticipated materials evolution in a fusion DEMO and to therefore provide a timely and robust materials database.

Defect clusters formed from large collision cascades in fcc metals irradiated with spallation neutrons

Fcc pure metals were irradiated with spallation neutrons (energies up to 500MeV) at room temperature to a neutron fluence of 1×1018n m−2 at KENS, High Energy Accelerator Research Organization (KEK). Defect clusters induced by large collision cascades were examined using transmission electron microscopy (TEM). In Au, large groups of defects included more than 10 clusters, and the damage zone extended over 50nm, which was larger than that induced by fusion neutron irradiation (<20nm). Although small stacking fault tetrahedra (SFT) are formed in subcascades by fission and fusion neutron irradiation, dislocation loops were also observed in the present experiments. Large dislocation loops (>10nm) were identified as vacancy type by the conventional inside–outside contrast method. Because of the low neutron fluence, spatial overlapping of collision cascades was ignored. Large vacancy loops are formed through cooperative reactions among subcascades in a single collision cascade with large recoil energy.

Defect structures before steady-state void growth in austenitic stainless steels

In the radiation damage process of austenitic stainless steels, there exists an incubation period before steady-state void growth, and the defect formation behaviors during that period strongly depend on alloy composition. Using the technique of positron annihilation lifetime measurement, the evolution of defect clusters during the incubation period in neutron, electron, and H-ion irradiations was studied for a variety of austenitic stainless steels including commercial and model alloys. The lifetime measurements indicated that in fission neutron irradiation to 0.2dpa at 363K, single vacancies were predominantly formed in the commercial alloys, SUS316L and Ti added, modified SUS316, while large voids were formed in Ni and Fe–Cr–Ni. After neutron irradiation at 573K, stacking fault tetrahedra and/or precipitates were detected in the commercial alloys, while large voids were detected in the model alloys. In the 30MeV electron irradiation to a dose of 0.012dpa, the effect of alloying elements on lifetime data was less significant at 353K, but a significant difference was found between model alloys and commercial alloys at 573K. The H-ion irradiation at 2MeV was also performed at room temperature. Defect evolution during the incubation period is discussed on the basis of the neutron, electron and H-ion irradiation results.

PI3K Activates NF-&amp;kappa;B Signal Pathway Which Prevents IEC-6 Cells from Neutron Irradiation

Neutrons are high LET rays. It damages the bodies more seriously than γ rays. Intestinal epithelial cells are very sensitive to neutron irradiation. While by far, there are no papers concerning PI3K regulation of NF-κB signal pathway and its effects in the intestinal epithelial cells damaged by neutron irradiation. To explore the change of nuclear factor-kappa B (NF-κB) pathway in intestinal epithelial cell No.6 (IEC-6) cell line induced by fission neutron irradiation and the regulation of PI3K, we set up IEC-6 cell lines model by subculturing. Three groups of IEC-6 were given: control group, neutron irradiation group, LY294002 treatment and neutron irradiation group. Except the control group, the other groups were homogeneously irradiated by 4 Gy fission neutrons. The LY294002 treatment group cells were given LY294002 before 24 h of neutron irradiation. At 6 h and 24 h after neutron irradiation of 4 Gy, the proliferation ability of the IEC-6 cells was assayed by MTT colorimetry, the apoptosis and necrosis rates of the IEC-6 cells were assayed by flow cytometry (FCM), and changes of NF-κB (p65), the inhibitor of kappa Bα (IκBα) and IκB kinase β (IKKβ) were assayed by Western blotting. We quantitatively analyzed the results by means of image analysis. At 6 h and 24 h after neutron irradiation of 4 Gy, their proliferation ability decreased and lots of apoptosis and necrosis IEC-6 cells were found; the changes of NF-κB and IKKβ in IEC-6 cells were decreased and that of IκBα increased. The injuries of LY294002 treatment and neutron irradiation group were more serious than the control and neutron irradiation group. We found that IEC-6 cells were obviously damaged and induced serious apoptosis and necrosis by neutron irradiation of 4 Gy. The NF-κB pathway in IEC-6 cells was activated by neutron irradiation which can protect IEC-6 against injury caused by neutron irradiation. PI3K can activate NF-κB pathway which protects IEC-6 cells from being damaged by neutrons.