The 35Cl(n, p)35S reaction plays a key role in neutron dosimetry for Boron Neutron Capture Therapy, in the synthesis of the isotope 36S, whose astrophysical origin remains unresolved, and in the design of next-generation molten-salt reactors. Its relevance has motivated its inclusion in the High Priority Request List (HPRL) of NEA. The goal of this work is to determine the 35Cl(n, p)35S cross-section from thermal energy to 120 keV for the first time ever in a single measurement, thus reducing systematic uncertainties related to the normalization to the thermal value. This had been a subject of concern in previous evaluations of this reaction. We made use of the Time-of-Flight technique with microMEGAS detectors at Experimental Area 2 (EAR-2) of n_TOF facility at CERN. The 10B (n, α ) 7Li and 235U(n, f) reactions were used as references. Rutherford Back-scattering Spectrometry was performed at Centro Nacional de Aceleradores (CNA) in Sevilla, in order to accurately determine the masses of the irradiated samples. We obtained a thermal cross-section of 0.470 ± 0.009 barns. The 1/v energy dependence of the cross-section is observed up to the first resonance at 0.398 keV, the resonances up to 120 keV are analyzed and resonance parameters extracted using SAMMY. Maxwellian Averaged Cross-Section (MACS) was calculated for k_B T from 1 to 100 keV, and lower values compared to estimations from ENDF were found, e.g., 1.07 ± 0.20 mb at k_BT=30 keV. The thermal cross-section and first two resonances are in agreement with the latest evaluation in ENDF/B-VIII.1, while remarkably lower resonance strengths were found for high energy resonances.

For several years, CEA has been developing highenergy imaging and photon interrogation techniques in the CINPHONIE irradiation bunker (CHICADE facility, CEA IRESNE, Cadarache, France). A new linear electron accelerator (K15 LINAC from Varex) has recently been commissioned and its irradiation field requires precise characterization. The manufacturer data for the mean electron energy are 15 MeV, but we asked for a slightly higher dose rate, resulting in a higher end-energy that needs to be characterized. In addition, the knowledge of the beam current is essential to ensure the reliability of both experimental results and numerical simulations. For that purpose, dose rate measurements in a water tank and delayed gamma-ray spectroscopy of thin metal foils activated in the X-ray field were carried out. A numerical model of the LINAC internals was built with MCNP6.3 to calculate dose rates in water and reaction rates in irradiation samples. A thorough characterization campaign was carried out to validate and calibrate this MCNP model against experiments. Finally, a Bayesian framework was applied to fine-tune the electron source energy distribution, estimate the average LINAC current, and quantify the X-ray angular deviation. Finally, the best estimates are E = (17.8 +/- 0.2) MeV and I-beam = (107 +/- 6) & micro;A.

Pile noise experiments in current mode were successfully conducted on the 37-fuel-assemblies core configuration of the LR–0 reactor operated by the CV Rez (Czech Republic). The experimental prompt decay constant (αp) was found equal to 244.0 ± 8.3 s−1, the effective delayed neutron fraction (βeff) equal 712 ± 15 pcm, and the generation time (Λ) was derived at 29.17 ± 1.2 µs. All uncertainties are equal to one standard deviation. The effective delayed neutron fraction was derived using a Diven factor computed at 0.875 with TRIPOLI-4.12. A bootstrap method was extensively applied on data to derive the kinetic parameters. Namely to compute the uncertainties on the power spectral density spectra, and throughout the data fitting procedure. The differences with MCNP6.1 and ENDF/B-VIII.0 nuclear data library are equal to −1.6 ± 3.4 % and +0.23 % ± 2.19 % on αp and βeff, respectively. These results are analysed in the light of several other pile noise experiments.

The TRIGA Mark II research reactor at the Jo & zcaron;ef Stefan Institute is a key facility in the field of nuclear research, characterized by its versatility and applicability in a wide range of scientific disciplines. This document highlights its operational history, contributions to nuclear safety, education and various scientific applications, including advances in reactor and radiation physics, neutron activation analysis, environmental science and even contributions to the fight against the COVID-19 pandemic. It highlights the reactor's significant role in fostering international collaborations, improving computer modeling techniques for nuclear research, and providing invaluable educational experiences. The great versatility and applicability of the JSI TRIGA reactor is emphasized by its adaptability to various research needs and its ability to enable groundbreaking studies in both fundamental and applied sciences.

Background: The ^35Cl(n, p)^35S reaction is of special interest in three different applications. First, in Boron Neutron Capture Therapy due to the presence of ^35Cl in brain and skin tissue. Second, it is involved in the creation of ^36S, whose astrophysical origin remains unresolved. Third, in the designing of fast nuclear reactors of new generation based on molten salts. Purpose: To measure the ^35Cl(n, p)^35S cross-section from thermal energy to 120 keV, determine the resonance parameters in this range and Maxwellian Averaged Cross-Section (MACS). Method: We made use of the Time-of-Flight technique with microMEGAS detectors at Experimental Area 2 (EAR-2) of n_TOF facility at CERN. The ^10B(n, α)^7Li and ^235U(n, f) reactions were used as references. Rutherford Back-scattering Spectrometry technique was performed at Centro Nacional de Aceleradores (CNA) in Sevilla, in order to accurately determine the masses of the irradiated samples. Results: We obtain a thermal cross-section of 0.470 ± 0.009 barns. The 1/v energy dependence of the cross-section is observed up to the first resonance at 0.398 keV, the resonances up to 120 keV are analyzed and MACS calculated for k_B T from 1 to 100 keV. Conclusions: The ^35Cl(n, p)^35S cross-section has been obtained over a wide energy range for the first time, with high accuracy across the aforementioned range. The thermal cross-section and first two resonances are in agreement with latest evaluation in ENDF/B-VIII.1, while lower resonance strength was found for high energy resonances. These data are used to calculate the MACS for different k_B T.

As part of its R&D programs on large objects characterization, the Nuclear Measurement Laboratory at the CEA-Cadarache center has equipped its high-energy tomograph with a new linear accelerator (linac): a VAREX K15 (9 to 15 MV range). This linac delivers a very high dose rate: up to 130 Gy/min at 1 m from the target. Combined with a mechanical bench and optimized detectors, this X-ray source allows handling very large objects for radiographies and tomographies, up to 1600 mm in diameter and 5 t in mass [1]. Compared with the kilovoltage range, MV energies offer two advantages: higher photon flux and deeper penetration capabilities (steel from 100 to 400 mm). The new X-ray source has been fully characterized in terms of dose rate, focal spot size and photon spectrum using water attenuation measurements. Depending on the geometry of the object to be scanned, two detectors can be used. The first is dedicated to larger objects and is a lens-based detector with different scintillator screens (Gadox or CsI, as described in [2]) specially designed for this configuration (with a screen size up to 800x600 mm2). The second, a commercial flat-panel with a small pixel pitch (0.1 mm) is used for the smallest but densest objects, where the spatial resolution is critical [3]. The performance of these detectors is characterized, compared and discussed. The tomograph set-up and its final performance at low (9 MV) and high (15 MV) energies are detailed in terms of MTF curves, and contrast-over-noise ratio obtained on specific mock-ups. Examples of tomography on real objects (industrial packages produced by metal additive manufacturing or radioactive waste drums) are also presented. Finally, the main drawbacks in this energy range are listed and detailed: 1) the scattering background caused by the Compton effect, 2) the thickness of the scintillator, which must be optimized according to the spatial resolution or expected efficiency and 3) the size of the X-ray source (limited to 1.5 mm). To overcome these limitations, various studies are currently underway, and the expected solutions are presented and discussed.

Electron linear accelerators (LINACs) are versatile and powerful X-ray sources, that can be used in medical radiotherapy as well as in various industrial applications including non-destructive testing, imaging and security inspection. LINACs accelerate electrons by passing them through a series of oscillating electric fields within a vacuum tube. These high-energy electrons are then directed towards a metallic target, producing X-rays (bremsstrahlung radiation) when they decelerate upon impact. In the field of non-destructive radioactive waste characterization, high-energy photon imaging (radiography, tomography) is used on large cemented radiological waste containers, with a volume of the order of 1 m3, to check their integrity and assess their content [1][2]. However, for such packages, passive gamma-ray spectroscopy, passive neutron counting and even active neutron interrogation fail in measuring nuclear materials, like plutonium and uranium. Therefore, high-energy photon interrogation techniques is under study to detect and quantify nuclear materials through the detection of induced-photofission particles. For the past years, CEA has been developing high-energy imaging [3] and photon interrogation techniques in CINPHONIE irradiation bunker (CHICADE facility, CEA IRESNE, Cadarache, France). CINPHONIE was recently upgraded with a K15 Varex accelerator that can reach a maximum dose rate of 130 Gy/min at 1 m from the X-ray target [4]. For advanced techniques (high-energy photon and photoneutron activations, photofission, bi-energy imaging), it is paramount to simulate precisely the irradiation field. For that purpose, a numerical model of the LINAC internals was built (with MCNP 6.3). It aims at simulating photon and neutron fields in view to calculate dose rates and reaction rates in irradiation samples, waste packages, but also in the whole casemate. A thorough characterization campaign was carried out to validate and calibrate this MCNP model against various experiments, including dose rate measurements in a water tank and delayed gamma-ray spectroscopy of thin metal foils activated in the X-ray field. These experimental results were used to fine-tune the electron source energy distribution as well as to estimate the average beam current. Its high-energy part is indeed particularly crucial for photofission and bi-energy studies.

LOENIEv2 is a long counter detector used for delayed neutron (DN) measurements. It is composed of sixteen 3He tubes embedded in a cylindrical high-density polyethylene (HDPE) matrix. Thanks to a special arrangement of the 3He tubes in three concentric rings, relative variations of the total efficiency as low as 2% can be reached over the [0.1 – 1 MeV] energy range. This paper addresses the development of a Monte-Carlo model of the detector and its validation thanks to calibrated neutron source measurements, performed at the NPL institute. These sources are in the form of small cylinders containing either a spontaneous fission material (252Cf) or a radioactive material producing neutron through (Α,n) reactions (AmLi, AmB, AmF, AmBe). These sources are well characterized in emission rate, spectrum and anisotropy so that they can used as standards for efficiency calibration. Using the JEFF-3.3 nuclear data library, the calculation of the detector absolute efficiency is validated within experimental uncertainties. The averaging of [C/E-1] values between the different sources end up to an estimated bias of -0.3% with an uncertainty of ±0.6% (1σ) over the range [0.5 – 4.1 MeV].

Large inconsistencies still exist in nuclear data libraries regarding the kinetic parameters of delayed neutron (DN) precursors. As an example, there is a 17 T_1/2 = (8.87 ± 0.10) s. Those results are consistent with the values recommended by the IAEA/CRP work and they come with reduced uncertainties compared with previously published results.

Kinetic neutron parameters are of fundamental importance in the field of nuclear reactor dynamics and control. Moreover, the precursor yield fraction and the neutron generation time for a given nuclear reactor are dependent on the properties of the reactor. Thus, in-pile experiments, such as oscillation experiments and noise experiments, are commonly conducted to measure those values. In this work, a method for determining the kinetic parameters of a reactor along with their covariance data from in-pile experiments is presented. It is performed by combining values of the reactor’s response function obtained from both oscillation and noise experiments over a wide range of frequencies. The method is carried out for the MINERVE zero power reactor (ZPR) using a reanalysis of both oscillation and noise experiments that were conducted in the MINERVE reactor in 2013 and 2014. Moreover, various advantages and disadvantages of performing multiple in-pile experiments and combining their results in order to obtain a single set of kinetic parameters along with their covariance data are considered. Some suggestions for the design of such in-pile experiments are also discussed.

In nuclear reactors operating in nominal conditions, delayed neutrons (DN) drive the kinetic behavior of the neutron population. Predicting this behavior requires therefore the knowledge of the DN yield per fission, average lifetime, and kinetic of their emission, which is described by precursors groups abundances. Characterizing the uncertainties of these macroscopic data is essential, especially as they are used to design the safety margins of reactors. Even for major fissile isotopes such as 235 U and 239 Pu, discrepancies still appear between the main nuclear data library. In 2018, CEA launched the ALDEN project (Average Lifetime of DElayed Neutrons), gathering several laboratories (CEA/DES and DRF, CNRS/LPSC, LP2i and LPC Caen, ENSICAEN, and University of Caen). This collaboration aims at producing new precise DN macroscopic data for reactor applications. Three experimental campaigns were conducted in 2018, 2019 and 2021 on the thermal fission of 233 U, 235 U and 239 Pu at Institut Laue Langevin in Grenoble (France), giving high quality results. In February 2023, a new campaign was performed at the PTB Ion Accelerator Facility in Braunschweig (Germany), focusing on the fast fission of 238 U. In particular, the DN yield was measured for eight energies between 1.5 and 19 MeV. This paper presents the experimental setup, that was adapted to the new facility and neutron spectrum. The methodology of analysis is also detailed, and some preliminary delayed neutron decay curves are showed.

This article presents an experimental effort to provide high-quality data to improve the evaluation of the 239 Pu delayed neutron yield in the thermal energy range. The set-up is composed of a long counter with sixteen 3 He tubes, a fast shutter system to produce irradiation cycles with short rising/falling times, and a miniature fission chamber containing 114μg of 239 Pu. The whole system was installed in the PF1B experimental zone of the Institut Laue-Langevin, which provides a cold neutron beam. The repetition of irradiation/decay cycles enables to saturate the delayed neutron precursors and to measure their yield through the observed activity, shortly after the beam-stop. The innovation of our measurement technique relies on the clear distinction between prompt and delayed neutron counting, thanks to boron absorbers, without the necessity to move the sample. In such a way, it is possible to normalize the counting of delayed neutron emission to the one of total neutron emission, based on the well-known value of the prompt neutron multiplicity. The present work provides a delayed neutron yield value of v d = 0.642(5)%. The latter is in 1σ agreement with the IAEA recommendation of 0.628(38)%, with a strongly reduced uncertainty thanks to our normalization technique.

The neutron capture cross section of 241 Am is an important quantity for nuclear energy production and fuel cycle scenarios. Several measurements have been performed in recent years with the aim to reduce existing uncertainties in evaluated data. Two previous measurements, performed at the 185 m flight-path station EAR1 of the neutron time-of-flight facility n_TOF at CERN, have permitted to substantially extend the resolved resonance region, but suffered in the near-thermal energy range from the unfavorable signal-to-background ratio resulting from the combination of the high radioactivity of 241 Am and the rather low thermal neutron flux. The here presented 241 Am(n,γ) measurement, performed with C 6 D 6 liquid scintillator gamma detectors at the 20 m flight-path station EAR2 of the n_TOF facility, took advantage of the much higher neutron flux. The current status of the analysis of the data, focussed on the low-energy region, will be described here.

Reactimeter algorithms enable real-time evaluation of the reactivity of a near-critical reactor from the signal of a neutron detector. They only depend on a few parameters defined by the inverse point kinetic equations, and which can be calculated by most neutron transport codes. However, the neutron population measured locally by a detector may not always lead to a good estimate of the reactivity. For example, this would be the case for an in-core detector placed just next to a moving control rod. In such special cases – or when measurement precision is an issue – costly simulations are needed to provide correction factors that account for variations in the spatial distribution of the neutron flux. This paper aims to provide an auto-corrected reactimeter algorithm (ACRA) in the special cases where rapid perturbations are applied to a core close to criticality. Such perturbations were successfully performed in the MINERVE reactor using a dedicated control rod to alternatively insert and withdraw a small sample of material under study. ACRA has been validated against 3D Monte Carlo neutron transport calculations using the signals of four in-core detectors located just next to the moving sample.

In nuclear research reactors, integral experiments are powerful tools to measure integral core parameters, such as the delayed neutron fraction. Within the scope of the point kinetic approximation, reactivity modulation experiments can be used for probing the reactor transfer function and then infer integral parameters of the core. In this context, Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA) and Ecole Polytechnique Fédérale de Lausanne (EPFL) have been collaborating for developing a probe device (PISTIL) and measurement setup adapted to the CROCUS zero power research reactor operated by EPFL (Lausanne, Switzerland). Despite some mechanical limitations of PISTIL, its maximum reactivity worth was measured with a good precision and repeatability using different methods (8.82 ±0.07 pcm), and its value is found rather close to the simulated one using TRIPOLI-4 (9.4 ± 0.4 pcm with JEFF-3.3). Above 1 Hz, the shape of the used modulation is pseudo-sinusoidal, with only a few well defined harmonics of excitation. The strongest harmonic only was analyzed using standard signal processing algorithms such as the Fourier transform and the Bartlett estimator. Twelve data points were produced in the range 0.5 Hz to 200 Hz, with uncertainty ranging from 1 % to 15 %. The prompt decay constant was measured at 150 ± 3 rad/s. Below 1 Hz, stepwise modulations were used with pseudo-random time sequences, which allowed exciting at once a large number of frequencies. Around 150 data points were produced in this particularly interesting frequency domain, between 1.6 mHz and 0.75 Hz, thanks to the use of three distinctive sequences with different base frequencies and overlapping ranges. The amplitude and phase of the RTF were measured satisfactorily, with uncertainties below 1 % for the strongest harmonics. The shape of the RTF was found consistent with the predictions of both JEFF-3.3 and ENDF/B-VII.1 libraries.

Background: The $^{14}\mathrm{N}(n,p)^{14}\mathrm{C}$ reaction is of interest in neutron capture therapy, where nitrogen-related dose is the main component due to low-energy neutrons, and in astrophysics, where $^{14}\mathrm{N}$ acts as a neutron poison in the $s$ process. Several discrepancies remain between the existing data obtained in partial energy ranges: thermal energy, keV region, and resonance region.Purpose: We aim to measure the $^{14}\mathrm{N}(n,p)^{14}\mathrm{C}$ cross section from thermal to the resonance region in a single measurement for the first time, including characterization of the first resonances, and provide calculations of Maxwellian averaged cross sections (MACS).Method: We apply the time-of-flight technique at Experimental Area 2 (EAR-2) of the neutron time-of-flight (n_TOF) facility at CERN. $^{10}\mathrm{B}(\mathrm{n},\ensuremath{\alpha})^{7}\mathrm{Li}$ and $^{235}\mathrm{U}(n,f)$ reactions are used as references. Two detection systems are run simultaneously, one on beam and another off beam. Resonances are described with the $R$-matrix code sammy.Results: The cross section was measured from subthermal energy to 800 keV, resolving the first two resonances (at 492.7 and 644 keV). A thermal cross section was obtained ($1.809\ifmmode\pm\else\textpm\fi{}0.045$ b) that is lower than the two most recent measurements by slightly more than one standard deviation, but in line with the ENDF/B-VIII.0 and JEFF-3.3 evaluations. A 1/$v$ energy dependence of the cross section was confirmed up to tens of keV neutron energy. The low energy tail of the first resonance at 492.7 keV is lower than suggested by evaluated values, while the overall resonance strength agrees with evaluations.Conclusions: Our measurement has allowed determination of the $^{14}\mathrm{N}(n,p)$ cross section over a wide energy range for the first time. We have obtained cross sections with high accuracy (2.5%) from subthermal energy to 800 keV and used these data to calculate the MACS for $kT=5$ to $kT=100$ keV.

The transfer function of a zero-power thermal reactor was successfully measured thanks to neutron noise techniques from 1 mHz to 160 Hz. During a month-long experimental campaign, the fluctuations of the neutron population in critical and subcritical configurations of the core were acquired using excore fission chambers and analysed through Cross-Power Spectral Density (CPSD) methodology. Firstly, the reactor’s kinetic parameters, i.e. prompt decay constant, effective delayed neutron fraction and generation time, were obtained at critical state. It required calibrating the reactor’s power, which was done by metal foil activation and measurement of the 235 U fission rate. Secondly, these parameters were used to estimate the groups’ abundances of delayed neutrons from the CPSD measured in a sub-critical state. Fitting data with a point kinetic model was done with Bayesian inference - CONRAD and Stan programs were used. A very good agreement was found between experimental abundances and the ones computed with TRIPOLI-4 Monte-Carlo transport code and JEFF3.1.1 nuclear data library. Uncertainties on prior abundances between 6 % to 101 %, held mainly by nuclear data, were lowered down to 4 % to 54 %, depending on the delayed group.

For safety reasons, it is important to simulate precisely the kinetic behavior of light water nuclear reactors (LWR) in nominal or incidental situations. In the point kinetic approximation, a few integral parameters allow describing the reactor's behavior. It is of interest to measure those parameters, such as the delayed neutron fraction and delayed neutron abundances, in low power research reactors loaded with benchmarked fuel con-figurations to compare with simulations results. In this context, Commissariat a` l'' Energie Atomique et aux ' Energies Alternatives (CEA) and Centrum Rez (CVR) have been collaborating towards measuring the delayed neutron fraction of LR-0, a VVER-type research reactor operated by CVR. This paper presents the first estimation of the delayed neutron fraction of the reactor loaded with six fuel assemblies, arranged in a small hexagonal lattice with a center cavity accessible for instrumentation. This configuration was chosen because it was very close to the so-called reference configuration with seven fuel assemblies, the one that was benchmarked in the NEA IRPheP Handbook. The measurement was performed in July 2021 using the CEA-developed acqui-sition system SPECTRON dedicated to pile noise experiments. Pile noise refer to a set of techniques which focus on the correlations induced by the fission process in the signals of neutron detectors. The experimental campaign was designed to meet two objectives: first to provide an estimation of the delayed neutron fraction; second to study the impact of the location of the neutron detectors on the quality of the results. Two fission chambers were used, one was fixed and located in the core center and the second was placed in various locations in the reflector. Several configurations were tested with the mobile detector progressively moved up to 20 cm from the closest fuel element. The reactor fission rate was obtained thanks to gamma spectrometry on metal foils (gold, nickel and iron). Simulations with TRIPOLI-4 (R) and JEFF33 libraries were performed to convert the saturated activities into fission rates. The delayed neutron fraction was estimated at 782 +/- 10 pcm and the generation time was measured at 39.1 +/- 0.5 mu s. The uncertainty balance is as follows: 2 % associated to the Diven factor and 1.5 % associated to the core's integral fission rate and 0.5 % for the statistical uncertainty. The reproducibility of the experiments was found very good, with a negligible dispersion amongst experiments (0.2 %). These results are valuable complements to the characterization of LR-0 reference neutron benchmark field.

Nuclear materials characterization in 870 L cemented waste drums is challenging due to significant neutron and gamma ray attenuation. Among the existing non-destructive approaches, photon active interrogation is of particular interest. A LINAC accelerates electrons to energies greater than 10 MeV, leading to highly penetrating interrogating bremsstrahlung photons that can reach nuclear materials in the depth of the 870 L waste drum and allow the detection of delayed gamma rays emitted by photofission products. Because the interrogating photon and neutron fluxes are extremely high during irradiation, measurements with high purity germanium detectors are usually performed after a long irradiation time, for example by moving the detector or the waste drum. Therefore, such measurements are only sensitive to delayed gamma rays associated to photofission products having sufficient long half-life of at least tens of seconds, which cuts a valuable signal coming from fission products with a shorter half-life. In this paper, we report gamma rays measurements of very short lived photofission products using a 3 in x 3 in cylindrical LaBr3(Ce) fast scintillation detector. A depleted uranium sample was irradiated by a pulsed bremsstrahlung source with maximum energy 17.5 MeV in a macro pulsing mode (1 s irradiation, 2 s cooling). This allowed to acquire a strong delayed gamma signal within a time window of 2 s.A clear signature of nuclear materials was observed even in the depth of a concrete mock-up, with an excellent signal-to-noise ratio beyond 3.5 MeV where the background of nonnuclear activated materials is negligible. These results are compared to a semi analytical model that qualitatively agrees but underestimates the measurements by a constant factor 1.7, probably caused by a wrong beam intensity normalization. Indeed, in this work the LINAC pulse time structure (macro pulsing) was different from its nominal working point used to characterize the beam (usual pulsing mode). Since it allows direct calculations of the detector response to delayed gamma-ray photofission measurements, this model is a fast calculation alternative to time-consuming Monte Carlo simulations, in view of further studies that will follow this feasibility demonstration.

Background: The 14N(n, p) 14C reaction is of interest in neutron capture therapy, where nitrogen-related dose is the main component due to low-energy neutrons, and in astrophysics, where 14N acts as a neutron poison in the s process. Several discrepancies remain between the existing data obtained in partial energy ranges: thermal energy, keV region, and resonance region. Purpose: We aim to measure the 14N(n, p) 14C cross section from thermal to the resonance region in a single measurement for the first time, including characterization of the first resonances, and provide calculations of Maxwellian averaged cross sections (MACS). Method: We apply the time-of-flight technique at Experimental Area 2 (EAR-2) of the neutron time-of-flight (n_TOF) facility at CERN. 10B(n, & alpha;) 7Li and 235U(n, f ) reactions are used as references. Two detection systems are run simultaneously, one on beam and another off beam. Resonances are described with the R-matrix code SAMMY. Results: The cross section was measured from subthermal energy to 800 keV, resolving the first two resonances (at 492.7 and 644 keV). A thermal cross section was obtained (1.809 & PLUSMN; 0.045 b) that is lower than the two most recent measurements by slightly more than one standard deviation, but in line with the ENDF/B-VIII.0 and JEFF-3.3 evaluations. A 1/v energy dependence of the cross section was confirmed up to tens of keV neutron energy. The low energy tail of the first resonance at 492.7 keV is lower than suggested by evaluated values, while the overall resonance strength agrees with evaluations. Conclusions: Our measurement has allowed determination of the 14N(n, p) cross section over a wide energy range for the first time. We have obtained cross sections with high accuracy (2.5%) from subthermal energy to 800 keV and used these data to calculate the MACS for kT = 5 to kT = 100 keV.