252Cf Source Research Articles

Neutron Shielding for 252Cf Source: FLUKA Simulations and Measurements

Neutron Shielding for 252Cf Source: FLUKA Simulations and Measurements

The Feasibility of Passive Neutron Measurement for Dry Storage Safeguards Approach

This work investigates the feasibility of using passive neutron measurement in identifying the removal or diversion of spent nuclear fuels (SNFs) in dry cask storage. In particular, we adopted a tomographic approach to measure emitted neutrons along the cask’s periphery through a laboratory-scale demonstration. An Arktis S670e helium-4 (He-4) detector array was utilized to measure neutrons. A 1/10th model cask was also built to mock the TN-32 dry cask. The test bench was conducted with three <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">252</sup> Cf sources to exemplify different assembly loading configurations. The measured neutron intensity patterns were then used to create images that represent the different source configurations. Also, Monte Carlo N-Particle Transport (MCNP) simulation was carried out to intercompare the experimental results with computational ones. Here, we describe the details behind the proposed safeguards technique through laboratory-scale demonstrations and examine its potentials and limitations.

Absolute mass calibration of fission product distributions measured with the E-[formula omitted] method

The mass calibration of fission product distributions measured with the energy–velocity (E-υ) method constitutes a technical challenge. The energy loss of the fission fragments in the various dead layers of the spectrometer and other sources of pulse-height defects of the energy detectors are a significant source of systematic uncertainty in the mass calculations used to determine the measured fission product yields. In this study, the absolute calibration of the fission mass distributions is accomplished by measuring prompt γ-rays in coincidence with the fission fragments. This allows for the direct calibration of the mass spectra and eliminates any dependence on potentially complicated energy corrections. The first test measurement was performed at the Los Alamos Neutron Science Center employing the SPectrometer for Ion DEtermination in fission Research (SPIDER). SPIDER is a 2E-2υ spectrometer designed for measuring independent fission product yields from neutron-induced fission. In this test, the single-arm SPIDER (E-υ) system and an array of 252Cf sources were used. The single-arm system consisted of two time pick-off detectors for measuring the time-of-flight of the fission fragments and a double-sided silicon strip detector (DSSD) for measuring the kinetic energy. Characteristic γ-rays from fission fragments were detected using three high-purity germanium (HPGe) detectors. For the mass calibration, γ–mass coincidence events from twelve product isotopes were used. The measured FPYs from 252Cf spontaneous fission were found to be in excellent agreement with the evaluated data after applying the absolute mass calibration. From the γ–mass coincidence events, the mass resolution of the system was also extracted. An average mass resolution of ∼1.4 AMU (FWHM) for the light fragments and ∼2.4 AMU (FWHM) for the heavy fragments were found. This was the first in-situ calibration and characterization of the SPIDER spectrometer, which paves the way for high-quality FPY measurements with this instrument.

Study on NaI(Tl) crystal at −35 [formula omitted]C for dark matter detection

We present the response of a NaI(Tl) crystal in terms of the light yield and pulse shape characteristics at two different temperatures: 22 °C (room temperature) and −35 °C (low temperature). The light yield is measured using 59.54keV γ-rays from a 241Am source relative to the mean charge of single photoelectrons. At the low temperature, we measure a 4.7 ± 1.3% increase in the light yield compared to that at room temperature. A significantly increased decay time is also observed at the low temperature. The responses to nuclear recoil events are measured using neutrons from a 252Cf source and compared to those to electron recoil events. The measured pulse shape discrimination (PSD) power of the NaI(Tl) crystal at the low temperature is found to be improved in the entire energy range studied because of the increased light yield and the different scintillation characteristics. We also find a 9.23 ± 0.26% increased quenching factor of α-induced events, which is the light yield ratio of α recoil to electron recoil, at the low temperature. This supports the possibility of an increased quenching factor of the nuclear recoil events that are known to have similar processes of dark matter interaction. The increased light yield and the improved PSD power of nuclear recoil events enhance the sensitivity for dark matter detection via dark matter–nuclei interactions.

Barite concrete-based cement composites for 252Cf spontaneous neutron and 60Co/192Ir shielding based on Monte Carlo computation

Barite concrete composite materials have been investigated for 252Cf spontaneous neutron and 60Co/192Ir gamma sources’ shielding using Monte Carlo computational method. The Particle and Heavy Ion Transport code System (PHITS) was used to compute the shielding properties of three different materials (barite concrete, barite cement, and barite aggregate) used as structural walls in fixed neutron & gamma industrial radiography for Non-Destructive Testing applications. The obtained results displayed good properties of barite concrete in shielding spontaneous neutrons emitted from the 252Cf source, as the effective dose drops about 108 times in only 140 cm wall thickness, and it was found to be about 10 times more effective than other materials investigated. In addition, the investigated gamma shielding properties of the barite concrete showed a relatively smaller wall thickness compared to the ordinary concrete. The decision-making process based on the ALARA principle of dose limitation showed that the use of barite concrete in such facilities is more effective than the use of barite cement and barite aggregate, for both gamma and neutron radiography shielding design. To achieve an average value of 1 μSv/h, the obtained result shows that 80 cm of Barite concrete is needed, while 125 and 130 cm of barite cement and barite aggregate are needed, respectively to shield the Co-60 source. Meanwhile, 50 cm of wall made of barite concrete is sufficient to cut down the effective dose rate to 1 μSv/h (for 50 Ci and 55 cm for 150 Ci 192Ir), which is an appropriate design for the public area adjacent to the industrial radiographic facility. It was therefore concluded from the obtained data that barite concrete is the most effective shielding material for radioactive sources (60Co, 192Ir, and 252Cf) used in radiographic applications.

NuCARIBU : An upgrade for the CARIBU facility at the Argonne Tandem Linac Accelerator System

The Californium Rare Isotope Breeder Upgrade (CARIBU) facility is changing the mechanism for creating neutron-rich fission products. Spontaneous fission from a 252Cf source has provided beams to support the low energy and accelerated-beams ATLAS programs. 252Cf has a 2.65-year half-life, requiring the source to be replaced every three years to maintain high beam intensities. Fabricating an appropriately thin 252Cf source to efficiently release the fission products has been challenging. The solution to these problems is nuCARIBU, a new system that provides neutron-induced fission on actinide foils. The Best Cyclotron B6P System (6-MeV proton beam at 0.5mA) is chosen, utilizing a multi-cusp negative ion source extracting into a cyclotron, which uses carbon foils to strip the H- ions to protons. These protons are delivered to a 7Li target to produce neutrons. The fast neutrons are moderated to thermal energies to induce fission in an actinide foil, providing neutron-rich fission products. These subsystems are described in detail along with comparisons between the current CARIBU facility and the presented nuCARIBU facility.This work was supported by the U.S. Department of Energy, Office of Nuclear Physics, under Contract No. DE-AC02-06CH11357

Boron-Coated Straws Imaging Panel Capability for Neutron Emission Computed Tomography for Source Localization Inside Radioactive Drums

The evaluation of fissile mass inside radioactive waste drums by neutron measurements is essential for radioactive waste management, nuclear safety, and criticality issues. However, passive and active neutron measurements can be strongly impacted by the uncertainty on the neutron source position within the drum and by the matrix effects. Therefore, an imaging panel proposed by Proportional Technologies Inc., and composed of seven boron-coated straw (BCS) detectors has been tested to localize neutron interactions, in view to reduce uncertainties associated with plutonium or uranium position inside radioactive waste drums. In a previous work, a numerical model of the imaging panel has been developed and validated from a comparison with experimental profiles obtained with a ²⁵²Cf source. In the first section, the feasibility of neutron emission tomography by a setup composed of five extended imaging BCS panels is demonstrated by numerical Monte Carlo simulation. The second section details the experimental validation of the neutron emission tomography. Measurements are carried out with AmBe and ²⁵²Cf sources located inside an empty 118 L drum by rotating the BCS imaging panel around it. Afterward, the Richardson&#x2013;Lucy deconvolution algorithm is applied to provide 2-D neutron source images for each angle. Finally, the 3-D images are reconstructed using the reconstruction toolkit (RTK) backprojection projection routine. The results demonstrate the capability of the BCS imaging to provide the 3-D location, i.e., axial and radial positions of one and two neutron sources. Furthermore, the first tests with this passive neutron measurement system show a satisfactory 3-D reconstruction for ²⁵²Cf and AmBe sources separated by 20 cm. Consequently, BCS imaging panels open interesting prospects to reduce the uncertainty associated with plutonium or uranium localization in neutron measurements. Work is undergoing to assess the capability of this system for 118 L drums filled with organic and metallic matrices. Additionally, further prospects concern the performance of other deconvolution and reconstruction algorithms.

Optimization of fast neutron and X-ray pulse shape discrimination (PSD) using an EJ276G scintillator in high radiation mixed conditions

Organic crystalline, liquid, and plastic scintillators with low Z-number materials are commonly used in fast neutron detectors. Pulse shape discrimination (PSD) is used to distinguish neutrons from X-rays because these detectors are affected by X-rays due to Compton scattering during fast neutron measurement. PSD was performed with a 1 × 1 × 3 cm3 EJ276G plastic scintillator, SiPM, and DAQ (10-bit resolution, 400 MHz) using the charge comparison method, which compares the total charge (Q body) and the delayed charge (Q tail) from the peak. Most studies have been conducted under laboratory conditions, such as with 252Cf or 241Am-Be sources. This study will perform PSD in a 15 MeV electron accelerator with a total flux of 6.72 × 1014 #/sec and an X-ray-to-neutron ratio of up to 4071:1. The PSD was optimized before the measurement in high flux conditions because it was difficult to perform PSD due to the pile-up effect in these conditions. Optimization focused on the figure of merit (FoM), which is an index of PSD performance. To distinguish neutrons from X-rays under these conditions, lead shielding is necessary to reduce the X-ray-to-neutron ratio. We performed Monte Carlo N-Particle Transport Code (MCNP6) simulations to determine the optimal shielding thickness. Fast neutrons interact with lead due to its high density, which results in small neutron energy loss and direction changes by scattering. This effect can affect other pixels as noise. We plan to evaluate the neutron images under high X-ray flux conditions by comparing the images of the MCNP6 simulation results and the measurement results.

Pulse-Shape Discrimination of SiPM Array-Coupled CLYC Detector Using Convolutional Neural Network

Cs2LiYCl6: Ce3+ (CLYC) is a dual-mode gamma-neutron scintillator with a medium gamma-ray resolution and pulse-shape discrimination (PSD) capability. The PSD performance of CLYC is greatly weakened when coupled with silicon photomultipliers (SiPMs) because of SiPMs’ low detection efficiency for the ultrafast Core-Valence-Luminescence (CVL) component under gamma excitation. In our previous work, the PSD Figure-of-Merit (FoM) value was optimized to 2.45 at the gamma-equivalent energy region of the thermal neutron by using the charge comparison method. However, this value was reduced to 1.37 at the lower gamma-equivalent energy region of more than 325 keV, and neutrons were difficult to distinguish from gamma rays. Hence, new algorithms should be studied to improve the PSD performance at low gamma-equivalent energy regions. Convolutional Neural Networks (CNNs) have excellent image recognition capabilities, and thus, neutron and gamma-ray waveforms can be discriminated by their characteristics through a known training set. In this study, neutron and gamma-ray waveforms were measured with a 137Cs source and moderated 252Cf source via an SiPM array-coupled CLYC detector and divided into two groups: training and PSD testing. The CNN training set comprised 137Cs characteristic gamma-ray waveforms and thermal neutron waveforms that were discriminated by the charge comparison method from the training group. A CNN with two convolution-pooling layers was designed to accomplish PSD with the test group. The PSD FoM value of the CNN method was calculated to be 37.20 at the gamma-equivalent energy region of more than 325 keV. This result was much higher than that of the charge comparison method, indicating that neutrons and gamma rays could be better distinguished with the CNN method, especially at low gamma-equivalent energy regions.

Investigation of CLYC-6 for thermal neutron detection and CLYC-7 for fast neutron spectrometry

Highly enriched 6Li and 7Li Cs2LiYCl 6 (CLYC) scintillators were investigated in view of their application as thermal neutron counter (CLYC-6) and as fast neutron spectrometer (CLYC-7) in the recently developed B-RAD radiation survey metre, an instrument designed to work in regions of high magnetic fields, currently equipped with a probe for photon dose rate measurements and γ-spectrometry. Both crystals were irradiated with a 2.5 MeV mono-energetic neutron beam from a Deuterium–Deuterium (D–D) generator and characterised in terms of neutron/γ-ray (n/γ) discrimination capability by the Pulse Shape Discrimination (PSD) method, energy resolution and quenching factor. Both crystals exhibit an excellent n/γ discrimination, quantified by a Figure of Merit higher than 2. The measured thermal energy resolution of CLYC-6 is 6%. The quenching factor for the 6Li(n,t)α thermal neutron reaction is 0.65 and for the 35Cl(n,p)35S reaction is 0.91. The measured neutron detection efficiency per unit volume of CLYC-6 is 60% higher than that of a 3He proportional counter. CLYC-7 was tested as a fast neutron spectrometer below 10 MeV, when irradiated with continuum neutron spectra from Am–Be and 252Cf sources. A proton/α-particle (p/α) discrimination was performed exploiting the PSD technique, without any unfolding procedure. The resulting spectra match the corresponding ISO spectra better than the spectra without the p/α discrimination, especially in the range 3 MeV to 5 MeV. A more sensitive PSD algorithm might help to improve particle discrimination at lower energies, while different reaction channels arise at higher energies preventing a spectrometry analysis.

Computational Investigation of Neutron Field modified with Various Materials

Medical accelerators, such as cyclotrons for Positron Emission Tomography (PET) and linear accelerators (LINAC), generate neutrons as a byproduct during operation. Neutron fields in these facilities tend to vary depending on the facility design as well as the presence of various materials in the facility. Unfortunately, these variations in the neutron field cannot be accounted for by neutron survey meters that are typically deployed in the workplace for radiation monitoring. In this work, the researchers used Monte-Carlo-N-Particle (MCNP) 6.1 to investigate the extent of field modification with materials that are commonly found in accelerator facilities. A computational investigation was performed using the field of the 252Cf source in the Philippine Nuclear Research Institute Neutron Laboratory (PNL) that was modified using polyethylene, lead, concrete and water. Variations in the spatial and energy distribution of neutron fluence as a result of introducing each modifier were presented. It was found that the relative deviation in the neutron dose due to polyethylene, lead, concrete and water modifiers were 96, 47, 83 and 94%, respectively. The result of this work was used as a basis for experiments performed at the PNL to investigate the limitations of commercial neutron survey meters when conducting dose measurements in different neutron fields.

Characterization of young Californium sources by means of [formula omitted]-ray spectroscopy

The spontaneous fission and neutron emission properties of Californium sources serve as a unique tool to analyze and validate the quality of detection of fission products. State-of-the-art methods for determining the age of a young Cf source were re-examined and modified. Further improvement included the determination of the single Cf isotopes within the Cf source and the neutron emission rate of the source.The analysis was performed by comparing the simulated γ-ray spectra of the Nucleonica suite with measured data of a young Cf source employing an HPGe detector. In addition, with the evaluation of the 133.5 keV peak (mainly due to the β-decay Ce-144/Pr-144) in relation to a peak at 497.1 keV (mainly due to the β-decay Ru-103/Rh-103) or 724.3 keV (multiplet, dominated by Ru-105/Rh-105 and Ce-145/Pr-145), the age-determination method of the investigated Cf source could be considerably improved to a confined age of 20 ± 3 months.By means of decay simulations, some deficiencies of current techniques in determining the isotope abundances were introduced. In particular, the quantities of Cf-252 and Cf-249 were analyzed extensively. Consequently, the source’s neutron emission rate matched well the emission rate from the source’s certification. The difficulties of defining the abundance of additional Cf isotopes within the source, albeit with lesser importance, are also illustrated.

Comparative optimization of Be/Zr(BH4)4 and Be/Be(BH4)2 as 252Cf source shielding assemblies: Effect on landmine detection by neutron backscattering technique

Monte Carlo simulations were used to model a portable Neutron backscattering (NBT) sensor suitable for detecting plastic anti-personnel mines (APMs) buried in dry and moist soils. The model consists of a 100 MBq 252Cf source encapsulated in a neutron reflector/shield assembly and centered between two 3He detectors. Multi-parameter optimization was performed to investigate the efficiency of Be/Zr(BH4)4 and Be/Be(BH4)2 assemblies in terms of increasing the signal-to-background (S/B) ratio and reducing the total dose equivalent rate. The MCNP results showed that 2 cm Be/3 cm Zr(BH4)4 and 2 cm Be/3 cm Be(BH4)2 are the optimal configurations. However, due to portability requirements and abundance of Be, the 252Cf-2 cm Be/3 cm Be(BH4)2 NBT model was selected to scan the center of APM buried 3 cm deep in dry and moist soils. The selected NBT model has positively identified the APM with a S/B ratio of 886 for dry soils of 1 wt% hydrogen content and with S/B ratios of 615, 398, 86, and 12 for the moist soils containing 4, 6, 10, and 14 wt% hydrogen, respectively. The total dose equivalent rate reached 0.0031 mSv/h, suggesting a work load of 8 h/day for 806 days within the permissible annual dose limit of 20 mSv.

Effect of Using Different Data Libraries and Simulation Codes on the Calculation of Spectra and Operational Quantities for the D2O-252Cf Source at PTB

The neutron reference field produced by a heavy-water moderated 252Cf source is used at PTB for calibrating neutron-measuring devices. Knowledge of the precise neutron spectrum is very important for the investigation of operational radiation protection quantities. Recently, new fission spectrum data of 252Cf has been proposed based on the latest nuclear data library version. At PTB, earlier calculations for the D2O moderated 252Cf neutron source were carried out more than 20 years ago, thus updated and more detailed calculations are required. In this paper, a detailed simulation model of the PTB moderated 252Cf source assembly has been prepared and investigated using new spectral data and two different Monte Carlo transport codes MCNP6.1 and PHITS3.22, with ENDF/BVIII.0, ENDF/B-VII.1, ENDF/B-V, and ENDL85 evaluated nuclear data libraries. The results show that the evaluated nuclear data libraries influence the calculated operational quantities by (3-5) %. The dosimetric quantities calculated with the PHITS code and the ENDF/B-VII.1 data library agree well with the MCNP6 results.

ELIGANT-GN — ELI Gamma Above Neutron Threshold: The Gamma-Neutron setup

Here we present the ELI Gamma Above Neutron Threshold setup recently implemented at the Extreme Light Infrastructure – Nuclear Physics. The main goals of the detector array are presented, as well as the mechanical design and the performance of the individual detectors. In addition, we have performed Geant4 simulations for realistic beam conditions. Finally, we show the performance of the array as a whole and outline the analysis procedure for complex experimental measurements by reproducing the angular correlations of neutrons following spontaneous fission of a 252Cf source.

Fast neutron spectra measurement in a copper using a 252Cf standard neutron source

The presented paper focuses on the measurement of fast neutron leakage spectra in the copper assembly using an intense standard 252Cf(sf) source. The neutron spectra were measured by a stilbene organic scintillation detector and compared with MCNP6.2 calculations using copper nuclear data from the ENDF/B-VIII.0, ENDF/B-VII.1, JEFF-3.3, JEFF-3.2, JENDL-4.0 and CENDL-3.1 libraries in the energy range of 1–10 MeV. Discrepancies appear in the intermediate neutron energy range between experiment and calculation. The best agreement with an experiment was achieved using ENDF/B-VIII.0 library in the region of 1–4.5 MeV. In the region of 4.5–10 MeV, the best agreement was achieved using JENDL-4.0 or CENDL-3.1 library. The largest discrepancies in the whole energy range were found for calculations involving JEFF-3.3 copper evaluations.

Characterization of a medium-sized CLLB scintillator: single neutron/gamma detector for radiation monitoring

The use of a single neutron/gamma detector is an interesting solution to detect and identify gamma emitters and also special nuclear materials (SNM), being able to discriminate between the two kinds of particles and also to perform good-resolution gamma spectroscopy.In this framework, we present a comprehensive characterization of a medium sized (2" × 2") CLLB (Cs2LiLaBr6:Ce) scintillation detector, in order to give the necessary information to assess its deployment in applications regarding homeland security and radiation monitoring. In particular, the parameters studied are: energy resolution, full-energy peak gamma efficiency, time resolution, thermal neutron/gamma discrimination capability, decay time of the signals, high counting rate performance and minimum detectable activities (of 137Cs and 252Cf sources). We employed digital nuclear electronics combined with a pulse shape discrimination algorithm to acquire and analyze the data. We compared our results with reported data for smaller CLLB scintillators, finding good agreement. Experiments were combined with Monte Carlo simulations (using GEANT4 v10.6.0 and MCNP5 v1.60) in order to complement the characterization. The obtained results suggest that the 2” × 2” CLLB detector offers better performance with respect to other scintillators of the same size such as NaI(Tl), CsI, CeBr, etc. which are commonly used in a radiation monitoring systems.

Gamma-blindness & neutron detection efficiency assessments with centrifugally tensioned metastable fluid detectors in low-scatter environment

A range of tests and associated analyses were performed in a state-of-art low scatter environment to demonstrate and validate the performance of the (Medium) Centrifugally Tensioned Metastable Fluid Detector (MCTMFD) for gamma blind detection of fast and epithermal neutrons for a variety of configurations, using Cf-252 and Am-Li isotope neutron sources. All measurements were performed in the Low Scatter Room (LSR) of Radiological Exposures & Metrology Laboratory at Pacific Northwest National Laboratory (PNNL). Experiments were performed to evaluate intrinsic neutron detection efficiency, ɛint n, the intrinsic gamma-neutron detection efficiency, ɛintγn, and the gamma absolute rejection ratio for neutrons (GARRn) as a function of tensioned metastable state negative pressure, Pneg, varying from 2.5 bar towards 9 bar. The specific embodiment of the MCTMFD utilized for the experiments at PNNL included a 16 cm 3 sensitive volume configured for either fast neutron detection using Decafluoropentane (DFP) as the detection medium, or fast and thermal neutron detection (using DFP with a chemical mixture composed of natural boron). Experiments were performed with Cf-252 neutron sources ranging in intensity from 1 × 104 n/s to 3.45 × 105 n/s, an Am-Li neutron source with intensity of 2.5 × 104 n/s, used in both a bare configuration and in combination with a 30 cm OD moderating sphere of D2O, with and without a 0.5 mm Cd shell, and intense Cs-137 and Co-60 gamma sources offering dose rates ranging from 27 mR/h to 53 R/h. It was found that the MCTMFDs remained gamma-blind (ɛintγn < 10−11) even under 53 R/h gamma background; the associated intrinsic fast neutron detection efficiency (ɛint n) varied from ∼0.2% for Pneg ∼2.5 bar, ∼10% for Pneg ∼5.5 bar and, ∼30% for Pneg ∼9 bar for a Cf-252 fission spectrum. In comparison, for the softer (Am-Li) spectrum, the ɛint n varied from ∼1% for Pneg ∼5.5 bar and approached the ɛint n values for Pneg states above 7 bar. The same MCTMFD when using a borated detection fluid with 20% B-10, offered a ×10–20 improvement in ɛint n for Pneg states ranging from 4.5–5 bar, for detecting epithermal neutrons from the Am-Li source (placed within a 30 cm OD, D2O moderator with and without a 0.5 mm Cd shield).

A Monte Carlo study on attenuation characteristics of colemanite- and barite-containing resources irradiated by 252Cf source against neutron–gamma photon

A Monte Carlo study on attenuation characteristics of colemanite- and barite-containing resources irradiated by 252Cf source against neutron–gamma photon

Special Nuclear Material Identification Through One-Minute Measurement with a New Backpack Radiation Device in Real Scenario Conditions

The constant concerns in global nuclear safety, aimed at deterring and combating the illicit trafficking of Material Out of Regulatory Control (MORC) and its possible use in criminal acts has raised the necessity of new detection solutions with higher efficiency and resolution to provide a high level of accuracy in the report to the authorities. Today’s radioactive isotope identifiers perform gamma spectroscopy identification and, sometimes, neutron counting. This paper presents the results of measurements performed in real-scenario conditions with a new type of portable radioactive isotope identifier for the detection and identification of both gamma and neutron sources. Its singular features are: 1) the capability to identify sources through the detection of neutrons, discriminating spontaneous fission sources (Cf-252), α-n sources (Am/Be, Am/Li) and nuclear material containing mix of isotopes of plutonium or uranium 2) the capability to make cross correlation between gamma and neutron measurements to achieve a higher level of accuracy in the identification of SNM that emits both neutrons and characteristics gammas. The test results are compared with international standards. The device exceeds the standard performance by triggering a neutron alarm for a 20.000 n/s Cf-252 source at a five times greater distance than the ANSI N42.34 one.