External Neutron Research Articles

Low Radon Cleanroom for Underground Laboratories.

Aim of a low radon cleanroom technology is to minimize at the same time radon, radon decay products concentration and aerosol concentration and to minimize deposition of radon decay products on the surfaces. The technology placed in a deep underground laboratory such as LSM Modane with suppressed muon flux and shielded against external gamma radiation and neutrons provides “Zero dose” space for basic research in radiobiology (validity of the LNT hypothesis for very low doses) and for the fabrication of nanoelectronic circuits to avoid undesirable “single event effects.” Two prototypes of a low radon cleanroom were built with the aim to achieve radon concentration lower than 100 mBq·m3 in an interior space where only radon-free air is delivered into the cleanroom technology from a radon trapping facility. The first prototype, built in the laboratory of SÚRO Prague, is equipped with a standard filter-ventilation system on the top of the cleanroom with improved leakproofness. In an experiment, radon concentration of some 50 mBq·m−3 was achieved with the filter-ventilation system switched out. However, it was not possible to seal the system of pipes and fans against negative-pressure air leakage into the cleanroom during a high volume ventilation with the rate of 3,500 m3·h−1. From that reason more sophisticated second prototype of the cleanroom designed in the LSM Modane uses the filter-ventilation system which is completely covered in a further improved leakproof sealed metal box placed on the top of the cleanroom. Preliminary experiments carried out in the SÚRO cleanroom with a high radon activity injection and intensive filter-ventilation (corresponding to room filtration rate every 13 s) showed extremely low radon decay products equilibrium factor of 0.002, the majority of activity being in the form of an “unattached fraction” (nanoparticles) of 218Po and a surface deposition rate of some 0.05 mBq·m−2·s−1 per Bq·m−3. Radon exhalation from persons may affect the radon concentration in a low radon interior space. Balance and time course of the radon exhalation from the human body is therefore discussed for persons that are about to enter the cleanroom.

Neutronics study of a subcritical system driven by external neutron source for 99Mo production

A subcritical system with the Low Enriched Uranium (LEU) solution was proposed to produce 99Mo, which is driven by an external D–T neutron source. The neutronics model of the subcritical system is approximately a concentric cylindrical structure with the maximum radius and height of the system as limited to 100 cm and 150 cm, respectively. The materials and dimensions of the zones were designed based on neutronics analysis including the sub-critical multiplication factor ks and distribution of neutron flux to achieve high production of 99Mo, as well as to keep the system in a safe and deep sub-critical state (i.e. 0.90<ks<0.95). With intensity of neutron source as 5 × 1013 n/s, the yield of 99Mo in the system can reach to 167 Ci/day, which could be used for around ten thousand nuclear medical procedures.

Evaluation and validation of external neutron source strength of operating Fast Breeder Test Reactor

Fast Breeder Test Reactor is one of the high flux operating test facility in India. The study aims to find out the source strength of the neutron source subassembly during its residence time in the reactor core and its contribution to the observed count rates in the startup neutronic channels which are ex-core 235U coated fission detectors. The entire power history during various fuel irradiation campaigns was considered in the calculation of neutron source strength. The study is experimentally validated by shifting the external neutron source subassembly to an elevated position in the subcritical core. The experiment is devised in such a way that inherent neutron source strength is not amplified and the count rates obtained in the detectors will be solely from the external neutron source. The count rates observed during the experiment at the ex-core detectors are compared with the simulated count rates during the lifted position of the source subassembly and found to be in good agreement.

Preliminary analysis of neutron time-delay histograms from Changvan latitude surveys

The name “Changvan” is given to a transportable neutron monitor, modernized by researchers in Thailand and housed inside a standard-size shipping container. It contains three neutron-sensitive proportional counters, which contain enriched 10BF3 gas, to detect cosmic ray showers. Primary cosmic rays are high-energy particles arriving from outer space. By colliding with atoms in the atmosphere, they produce a shower of sub-atomic particles, called secondary particles. Some particles can reach the Earth’s surface, including neutrons. The Changvan is designed to measure atmospheric neutrons during a latitude survey. The side counter tubes are ringed with lead producer to produce evaporation neutrons when secondary particles collide with it. The center counter lacks the lead producer but can receive neutrons from nearby counters. The reflector and moderator made from polyethylene thermalize the evaporation neutrons and protect against disturbance by external low-energy neutrons. In this work, we pursue methods to determine spectral changes in Galactic cosmic rays using data from the Changvan, avoiding the systematic inconsistencies of cross-station comparisons. We examine neutron time delay histograms from the three counter tubes, as recorded by specially designed readout electronics, where the time delay refers to the time interval between consecutive neutron detections in one counter. We perform a preliminary analysis of such histograms to measure the leader fraction (L) of neutrons that do not follow a previous neutron from the same primary cosmic ray. The electronic dead time will be considered in the L calculation.

A non-intercepting monitoring method for beam position on the target in an accelerator driven system

In an accelerator driven sub-critical system (ADS), an intense external neutron source provided by an accelerator coupled to a target is used to drive a sub-critical reactor. To ensure the stable operation of the reactor, it is of crucial importance to locate the beam spot on the target during the operation. In this paper, we have proposed a method for beam position measurement on the target’s beam entrance window. The proposed method is to measure the neutron flux outside the spallation target in the four directions of the east, south, west, and north. Since the neutron flux decreases with the distance between the detector and the beam center, the position of the beam center on the target’s entrance window can be extracted from the measured four neutron fluxes. The simulations based on FLUKA and OpenMC codes have indicated that the x and y positions of the beam center can be extracted from the measured four neutron fluxes. Since the four neutron detectors are fixed inside the reactor core, this method allows for continuous measurement of the beam position.

Fractional two energy groups matrix representation for nuclear reactor dynamics with an external source

Nuclear reactor dynamics is the study of the time-dependent behavior of neutron density in an exceeding reactor. Knowledge of however the density in a reactor, which can respond to numerous perturbations, is important to a safe and economical style. Unfortunately, this information is extremely tough to get. To study the behavior of nuclear reactors it is necessary to solve the time-dependent two energy groups neutron diffusion model which is a system of stiff coupled partial differential equations. The two energy groups point kinetics equations with delayed neutrons derived from the neutron diffusion equations in the presence of the time-dependent external neutron source is solved analytically. Since the fractional order differential operator is non-local, then the fractional model of the two energy groups point kinetics equations can be a good representation for the neutron density in the nuclear reactors. In this work, the two energy groups fractional point kinetics equations with multi-group of delayed neutron precursors (2EFPKE) are formulated in the matrix form and two techniques are proposed to solve this system of differential equation analytically, which based on Laplace transformation and eigenvalues and corresponding eigenvectors of the coefficient matrix. The analytical procedure to the proposed fractional model in nuclear reactor dynamics is described and investigated for different types of time-varying reactivity. The performance of developed methods has been tested for both fast and thermal neutron density in the case of the time-dependent external neutron source. In addition, the anomalous diffusion processes, sub-diffusion, and super-diffusion processes are discussed for various fractional order.

Регистр хронической лучевой болезни когорты работников ПО «Маяк», подвергшихся профессиональному облучению

Purpose: To present descriptive characteristics, and structure of the chronic radiation sickness (CRS) registry prospects of its use. &#x0D; Materials and methods: A registry for CRS diagnosed in workers of the nuclear production facility Mayak Production Association (PA) throughout the follow-up period of 1948–2018 was established within a medical and dosimetry database ‘Clinic’ of the Southern Urals Biophysics Institute.&#x0D; Results: The CRS registry includes 2068 cases: 1517 (73.4 %) in males and 551 (26.6 %) in females. Almost all workers (97.9 %) with CRS were hired at the Mayak PA in 1948–1954 and chronically externally and/or internally exposed to ionizing radiation. At a date of CRS diagnosis the mean cumulative red bone marrow absorbed dose of external exposure to gamma rays was 1.1 ± 0.7 Gy in males and 1.0 ± 0.6 Gy in females; the mean annual dose was 0.46 ± 0.33 Gy and 0.38 ± 0.22 Gy in males and females, respectively; maximum annual dose was 0.67 ± 0.46 Gy and 0.55 ± 0.34 Gy in males and females, respectively. The CRS frequency in the Mayak PA worker cohort significantly increased with the cumulative and mean annual RBM absorbed dose of external exposure to gamma rays. In the meantime, the CRS frequency was not associated either with a dose of external neutron exposure or with a dose of internal exposure to alpha particles from incorporated plutonium. &#x0D; Conclusion: The established CRS registry providing complete high quality demographical, medical and dosimetry information, together with available biological specimens, in future will allow: the updating of dose–response and dose–time–response relationships; the estimation of latent periods, risks and dose thresholds and associated uncertainties for CRS development; certain tissue reactions in lymphoid and haematopoietic tissues; and a better understanding of their development patterns and mechanisms, taking into account non-radiation factors.

Tumor 18F-BPA Uptake in PET and Boron Neutron Capture Therapy Efficacy in Locally Recurrent Head and Neck Cancer

Boron neutron capture therapy (BNCT) is high-LET radiotherapy, where boron-10 is first targeted to tumors using a carrier compound, often L-boronophenylalanine (L-BPA), followed by low energy external neutron beam irradiation. A high target dose from BNCT is predictive for tumor response and favorable survival in patients with head and neck (H&N) squamous cell carcinoma. Constant tumor-to-the whole blood boron concentration ratios are often assumed in dose calculation, but tumor-to-normal tissue (T/N) ratios obtained from tumor imaging with PET using 18F-BPA as the tracer might also be useful in dose calculations. We investigated the association between the T/N ratios obtained from 18F-BPA PET and efficacy of BNCT in patients with H&N cancer. 117 patients with inoperable, locally recurred H&N cancer were treated with L-BPA-mediated BNCT at the FiR 1 research reactor facility, Espoo, Finland, in February, 2003 to January, 2012. To estimate tumor BPA accumulation, 33 patients underwent 18F-BPA PET prior to BNCT. The T/N ratios were evaluated from static emission scans. Blood boron concentration was monitored with inductively coupled plasma atomic emission spectrometry during neutron irradiation. Normal tissue boron concentration was assumed to equal to that of the whole blood. Tumor response to BNCT was assessed with conventional imaging using the RECISTv.1.0 criteria. Thirty-three patients (male, 55%) participated in the study. Most (25, 76%) had squamous cell carcinoma, 4 (12%) sarcoma, and 4 adenocystic carcinoma. The median minimum radiation dose to the gross tumor, calculated with the measured blood boron concentrations and the T/N ratios from PET, was 14 Gy (RBE) (range, 7 to 78 Gy [RBE]), and when a constant tumor-to-blood ratio of 3.5 was assumed, 19 Gy (RBE) (range, 10 to 29 Gy [RBE]). Two patients were unevaluable for response to BNCT (died early). Nineteen (61%) of the 31 evaluable responded (had a CR or a PR). The median overall survival (OS) time estimate was 22.8 months; 2-year OS was 42%. The 18F-BPA PET-derived T/N values ranged from 1 to 18 (median, 2.7). The median OS for the patients who had a T/N ratio >2.7 was 12.3 months as compared to 27.7 months among those with a smaller ratio (log-rank p = 0.036). A higher than the median T/N ratio did not predict for tumor response to BNCT (p = 0.24). A high tumor T/N ratio derived from 18F-BPA-PET is associated with unfavorable survival after BNCT. Cancers with high 18F-BPA uptake in PET may often be biologically aggressive, which hypothesis requires further study.

Core power control of the ADS based on genetic algorithm tuning PID controller

The accelerator driven subcritical system (ADS) reactor, an accelerator driven transmutation research facility relies on control rod and external neutron source to realize core power regulation. This paper researches the dynamic and control characteristics of the ADS core under the external neutron source disturbance. The state space model of the ADS core with multi-input multi-output is established by linearizing the core nonlinear model with small disturbance. The response characteristics of the ADS core power control under external neutron source strength regulation are simulated by using genetic algorithm tuning proportional-integral-derivative (PID) controller and traditional PID controller. The results show that under different external neutron source relative strength, the effect of genetic algorithm tuning PID controller is better than that of traditional PID controller. The ADS core power can be well controlled by using genetic algorithm tuning PID controller.

Dynamic simulation of CIADS core power control based on the duty ratio of the proton beam

The China Initiative Accelerator Driven System (CIADS) subcritical reactor uses a superconducting proton linear accelerator to form a periodic pulsed proton beam, which excites the periodic external neutron source to realize core operation. To study the influence of the periodic pulsed proton beam on the power control of the core, a lumped parameter method is used to establish a nonlinear model of the reactor core, which is then transformed into a linear model. Considering the periodic nature of the pulsed proton beam, the duty ratio of the proton beam is introduced to represent the duration of excitation of the pulsed neutron source in a single period. To study the influence of the periodic proton beam on the core power and propose a control strategy, a CIADS core power simulation system is designed and established based on a proportional-integral-derivative (PID) controller. The results show that the duty ratio and amplitude play an important role in the changes in the core power. Under regulation by the PID controller, different duty ratios and amplitudes of the periodic beam can be used to achieve core power control.

226Ra irradiation to produce 225Ac and 213Bi in an accelerator-driven system reactor

This study aims to produce 229Th using an innovative nuclear reactor concept, i.e., accelerator-driven system (ADS) reactor. Herein, we investigated the feasibility of producing 229Th from neutron transmutation of 226Ra to expand the availability of 225Ac and 213Bi in a simple model of ADS reactor. ADS reactor comprises two zones, i.e., an inner zone with a fast neutron spectrum and outer zone with thermal neutron spectrum, which is a subcritical core coupled with an external neutron source. Transmutation behavior, activity, and mass ratio of the obtained isotopes were investigated using the Monte-Carlo tool. In addition with offering the capability, flexibility, and safety of radioactive waste transmutation, the proposed ADS model provides high 229Th yield and requires less time than a critical reactor with the same neutron flux and irradiated quantity of 226Ra.

Isotopic depletion calculation with generalized perturbation theory in subcritical systems driven by an external neutron source

Generalized Perturbation Theory (GPT) is used with the aim of proposing a perturbation method for isotopic depletion calculations. Changes in nuclide densities over time induce variations in the neutron flux and these lead to changes in reaction rates present in the evolution matrix. The proposed perturbation method treats these reaction rates as integral quantities and changes in these integral quantities as perturbations coming from changes in nuclide densities with fuel burnup. This method is applied to a subcritical system driven by an external neutron source. The results obtained by perturbation method show good agreement with the direct calculation. Considering that only a 1D problem was tested, other problems treating more energy groups and 2 or 3 spatial dimensions should be done in order to fully validate the method.

A Monte Carlo study of neutron contamination in presence of circular cones during stereotactic radiotherapy with 18 MV photon beams

High-energy photons are being used to treat different kinds of cancer, but it may increase the rate of secondary cancers due to the neutron contamination as well as over exposing of patients and medical staffs in radiation therapy Takam, Bezak, Marcu, and Yeoh, 2011, Radiation Research, 176, 508-520. Due to some difficulties in experimental measurements of neutron contamination, Monte Carlo method is an efficient tool to investigate dose parameters and characteristics in new techniques. The 18-MV photon beam of linac and circular cones have been simulated by MCNP5 code. Various parameters of photon and neutron including mean energy, flux, KERMA, the number of particles crossing a surface at a distance of 100 cm (SSD = 100 cm) as well as the change in photon and neutron spectrum as well as in intensity through the transmission in the circular collimators have been investigated. The results of this study show that the use of a circular collimator decreases neutron dose in the central axis, which is an advantage, but neutron contamination inducing small neutron dose is distributed all over the space. On the surface of phantom, photon dose rate is approximately equal to 3.41E7 (mGy/mA.min) for different collimators, but the neutron dose rate is 1.64E2 (mGy/ mA.min), 2.03E2 (mGy/ mA.min) and 2.52E2 (mGy/mA.min) for diameters of 12, 20 and 40 mm, respectively and it decreases by decreasing the diameter of the collimator. The neutron dose rate decreases from 9.68E7 and 9.74E7 (mGy/min.mA) for open field size 33 cm2 and 55 cm2 to 1.64E2 (mGy/min.mA), 2.02E2 (mGy/min.mA) and 2.52E2 (mGy/min.mA) for collimator diameter of 12 mm, 20 mm and 40 mm. It can be concluded that the use of circular collimators has an advantage of reducing neutron dose in the central axis. It should be mentioned that the off-axis neutron dose surrounding the collimator can be eliminated using an external neutron shield without perturbing the treatment field.

Nondestructive measurements of residual 235U mass of Israeli Research Reactor-1 fuel using the Advanced Experimental Fuel Counter

In 2018, a measurement campaign took place with participants from Los Alamos National Laboratory (LANL), the Nuclear Research Centre-Negev (NRCN) and Soreq Nuclear Research Center (SNRC) at the Israeli Research Reactor-1 (IRR-1) in which 14 of the reactor’s used fuel assemblies (FAs) with varied amount of depletion were measured with the nondestructive assay instrument Advanced Experimental Fuel Counter (AEFC). Designed for safeguards purposes, the AEFC measures both neutrons emitted from the FA (passive neutrons) and fission neutrons induced by an external neutron source (in this experiment, 252Cf). Signals recorded with the AEFC include total neutron count rates (Singles), time-correlated neutron count rates (Doubles), and total gamma-ray count rates. The 235U content of the FAs was previously assessed by two independent methods: (1) measurement of the transparency of the FA to low-energy gamma rays from an activated rhenium source (rhenium gamma transmission, or the RGT method) and (2) calculation of the ∼30-year burnup history of the core using detailed three-dimensional Monte-Carlo core depletion calculations. The results from the FAs that had been measured via the RGT method were used to construct the calibration curves, which translate the AEFC count rates to 235U mass. Then, the calibration was evaluated using AEFC measurements of six additional FAs that were not measured via the RGT method. From the results, it was determined the Doubles calibration curve was more reliable than that of the Singles and follows a simple second-order polynomial fit for the whole range of residual 235U mass content, albeit with larger statistical uncertainty.Detailed uncertainties quantification was conducted for both the AEFC Singles and Doubles. This includes the analysis of statistical uncertainties, calibration uncertainty, and random uncertainties due to the sensitivity of the AEFC to several sources of uncertainty, namely the FA position, FA orientation, interrogation source position, and ambient pool temperature. An overall total uncertainty of 6 g of 235U is estimated for the Singles and Doubles, which is mainly due to calibration uncertainty (for the Singles) and statistical uncertainty (for the Doubles), and which constitutes 3%–6% of the 235U total mass in the FAs, depending on their level of depletion.

LQG/LTR controller with simulated annealing algorithm for CIADS core power control

The purpose of this study is to design the core power controller for China Initiative Accelerator Driven System which only relies on the external neutron source to realize core power adjustment. Based on the state space model of CIADS core, a Linear Quadratic Gaussian controller with Loop Transfer Recovery (LQG/LTR) is adopted as the strong robustness. As the parameters of traditional LQG/LTR controller are selected by judging the singular value curve, it is difficult to achieve the optimum. To find the optimal parameters of the LQG/LTR controller, simulated annealing algorithm (SAA) is used. With the state space model of CIADS core, the core power control system is established by using the LQG/LTR controller with SAA. The core power step simulation and the external neutron source disturbance simulation are carried out. The results show that the control effect of LQG/LTR controller with SAA is better than that of traditional LQG/LTR controller.

Subcriticality estimation by extended Kalman filter technique in transient experiment with external neutron source at Kyoto University Critical Assembly

The monitoring subcriticality by the extended Kalman filter (EKF) is evaluated in the presence of an external neutron source through transient analyses and compares with that by inverse kinetic method at Kyoto University Critical Assembly. Throughout the transient experiment, subcriticality is deduced with the use of neutron counts measured every 1 s. For ensuring the initial condition of the EKF technique, the basic transient experiment is carried out by dropping a control rod into the core at a critical state, and the result by the EKF technique shows remarkably accurate subcriticality as compared with the reference value obtained by the rod drop method with less fluctuation than by the inverse kinetic method. In the ADS transient experiment with the 14 MeV stable neutron source, the subcriticality is set as target subcriticality of the accelerator-driven system ($$k_\mathrm{eff}= 0.97$$); the external neutrons are then injected from outside the core. The EKF technique reveals significantly good agreement with MCNP6.1, whereas subcriticality by the inverse kinetic method fluctuates widely during the transient experiment. Thus transient experiment shows that the EKF technique is applicable to subcriticality monitoring with a high time resolution and a high degree of precision.

The Status of AMoRE Double Beta Decay experiment

AMoRE (Advanced Mo-based Rare process Experiment) is searching for neutrinoless double beta decay of 100Mo. A pilot experiment, AMoRE-Pilot, was operated with six 48dep1Ca100MoO4 crystals of total mass ~1.9 kg in a cryostat at the Yangyang underground laboratory (Y2L) with the minimum overburden of 700 m. Since AMoRE-Pilot started in August 2015, two major improvements in its configuration took place to reduce its background level. The first one was replacement of detector components by materials with lower radioactivity level. The second one was the installation polyethylene blocks together with borated rubber sheets to reduce external neutron backgrounds. Using the AMoRE-Pilot data in different configurations, background modelling of the AMoRE-Pilot was performed.

Investigating neutron activated contrast agent imaging for tumor localization in proton therapy: a feasibility study for proton neutron gamma-x detection (PNGXD)

Proton neutron gamma-x detection (PNGXD) is a novel imaging concept being investigated for tumor localization during proton therapy that uses secondary neutron interactions with a gadolinium contrast agent (GDCA) to produce characteristic photons within the 40–200 keV energy region. The purpose of this study is to experimentally investigate the feasibility of implementing this procedure by performing experimental measurements on a passive double scattering proton treatment unit. Five experimental measurements were performed with varying concentrations and irradiation conditions. Photon spectra were measured with a 25 mm2, 1 mm thick uncollimated X-123 CdTe spectrometer. For a 10.4 Gy administration on a 100 ml volume phantom with 10 mg g−1 Gd solution placed in a water phantom, 1129 ± 184 K-shell Gd counts were detected. For an administered dose of 21 Gy and the same Gd solution measured in air, resulted in 3296 ± 256 counts. A total of 1094 ± 171, 421 ± 150 and 23 ± 141 K-shell Gd counts were measured for Gd concentrations of 10 mg g−1, 1 mg g−1 and 0 mg g−1 for 7 Gy dose in air. The signal to noise ratio for these five measurements were: 7, 15, 6, 3, and 0.2, respectively. The spectrum contained 43 keV Kα and 49 keV Kβ peaks, however a small amount of 79.5 and 181.9 keV prompt gamma rays were detected from gadolinium neutron capture. This discrepancy is due to a drop in the intrinsic detection efficiency of the CdTe spectrometer over this energy range. The measurements were compared with Monte-Carlo simulation to determine the contributions of Gd neutron capture from internal and external neutrons on a passive scattering proton therapy unit and to investigate the discrepancy in detected characteristic x-rays versus prompt gamma rays.

Conceptual design and uncertainty analysis of a typical 50 kW aqueous homogeneous reactor aimed for medical isotope production

Production of medical isotopes such as 99Mo, 90Sr, 131I and other fission products using Aqueous Homogeneous Reactors (AHRs) offers several advantages in comparison to other methods. In this method, after about one week operation, required isotopes can be directly extracted from the solution and the remaining fuel solution will be returned to the core for the next campaign. Moreover, AHRs are inherently safe and reliable due to their large negative reactivity feedbacks. This paper demonstrates technical features and conceptual scheme of the Iranian homogeneous reactor named RAHA, which is mainly used for the production of 99Mo/99mTc. Neutronic and thermal-hydraulic characteristics of the reactor were analyzed with MCNPX-2.6 and RELAP5/Mod3.3 codes, respectively. It was assumed that the reactor is fueled by Uranyl Nitrate, and is able to remove generated heat using natural circulation. The coolant circulation velocity was obtained as 6.5 cm/s. Four different types of neutron absorbers were considered as control rod material, and BORA resin with a diameter of 1.4 cm and Stainless-Steel cladding was chosen. As a new characteristic which may improve safety of the system, the possibility of reactor startup without any external neutron source for the case of fresh fueled core was assessed. Finally, effects of the possible uncertainties on some physical and material properties of the core were investigated and sensitivity of the effective multiplication factor relative to the uncertain parameters was also extracted using Spearman's formula.

State estimation of external neutron source driven sub-critical core using adaptive Kalman filter

Extended Kalman filter (EKF) and cubature Kalman filter (CKF) are proposed to estimate the state parameters of an external neutron source driven sub-critical reactor, including power level, reactivity, external neutron source, six-groups of delayed neutrons precursor densities, equivalent fuel temperature, average coolant temperature and nuclear densities of iodine, xenon, promethium, samarium nuclides. Parameter settings and matters needed attention in EKF and CKF are also analyzed, especially the relationship between model prediction covariance matrix and measurement covariance matrix. In order to effectively identify the maneuvering of the external neutron source and reactivity on the uncertainty of the prediction model, two adaptive algorithms are proposed to adjust the covariance matrix of the prediction model online. The results show that these two adaptive algorithms can effectively detect various maneuvering such as neutron source variation and reactivity insertion, and realize the optimal estimation of the reactor state using EKF method. However, CKF has divergence and non-convergence. EKF achieves good results in all parameters estimation.