Safety Analysis Report Research Articles

SCALE inventory and reactivity analysis as part of the Hermes 2021 PSAR review

The readiness of SCALE for comprehensive studies of pebble-bed reactors has been demonstrated through detailed analysis of a fluoride salt–cooled, high-temperature pebble-bed reactor (PB-FHR). The methods developed for pebble-bed reactor modeling in SCALE, particularly for inventory generation, have proven effective in gaining insights into the reactor physics of this advanced reactor. Excellent agreement with another code package has been observed, further highlighting SCALE’s strong performance.The SCALE results supported the US Nuclear Regulatory Commission’s construction permit application review of the Hermes low-power PB-FHR demonstration reactor. A SCALE model of the Hermes reactor was developed at Oak Ridge National Laboratory using information from the Preliminary Safety Analysis Report (PSAR) and supplemented with publicly available data. SCALE reactivity coefficient simulations reproduced PSAR results within 1σ statistical uncertainties. Sensitivity studies emphasized the importance of graphite specifications for accurate keff predictions.

Model predictive control of a grid-scale Thermal Energy Storage system in RELAP5-3D

This research delves into the integration and control of a Thermal Energy Storage (TES) system with a Small Modular Reactor (SMR), specifically the NuScale VOYGR SMR module in RELAP5-3D. The research methodology centered on modeling the NuScale VOYGR SMR, a light water pressurized water reactor (LWR) with a power output capacity of 77 MWe per module. The reactor and plant details were sourced from NuScale's final safety analysis report and supplemented by information from the NuScale website. The SMR plays a crucial role in energy generation, and to manage and dispatch the produced energy effectively, a robust storage system is essential. The proposed solution to this challenge is the implementation of the TES system. The selected TES for this research is a two-tank system. The study also employed Model Predictive Control (MPC) to optimize the operation of the TES system in conjunction with the SMR. Various simulations, including accident scenarios, were conducted to assess the system's response and performance. The research leveraged real energy demand data from the California Independent System Operator (CAISO) database and scaled it to reflect the power generation of a single SMR. The findings suggest that while integrating a TES system with an SMR can enhance the performance compared to a standalone SMR, certain scenarios might exacerbate the total power mismatch. The study provides insights into the potential of integrating TES systems with nuclear reactors and the challenges and considerations involved.

Large pipe break opening time in pressurized heavy water reactors

In the frame of the licensing process of the AtuchaII PHWR (pressurized heavy water reactor) the BEPU (Best-Estimate Plus Uncertainty) approach has been selected for issuing Chapter 15 of FSAR (Final Safety Analysis Report) dealing with accident analysis. In case of a LOCA (Loss of Coolant Accident) the assumption of instantaneous opening time is too conservative, especially for this kind of reactor with a positive coolant void reactivity coefficient.Analysis was performed to assess the ability of Atucha II nuclear power plant’s fast boron-injection system to prevent overheating of the fuel elements in the event of a Beyond Design Basis Accident (BDBA). The goal of the Analysis was to develop a Dynamic Break Opening which was used as the input for the LOCA analyses. The results show that the fast boron injection can effectively shut down the reactor, thereby preventing the overheating of the fuel elements.

NuScale-like SMR Model Development and Applied Safety Analyses with the Code Chain Serpent-DYN3D-ATHLET

NuScale is an integral pressurized water reactor (iPWR) operated with light water driven by natural circulation through two helical coil steam generators (HCSGs). This work reports the safety analyses of a boron dilution and steam line break accidental sequences in a developed plant computational model based on the specifications from the Final Safety Analysis Report (FSAR). Multi-physics calculations are performed with the code chain Serpent-DYN3D-ATHLET. A state-of-the-art multi-dimensional vessel topology is developed with ATHLET for the accurate representation of flow and temperature fields, as well as spatial core power distribution. The static calculation results show agreement with the reference values from the FSAR. The boron dilution sequence shows a homogeneous core power excursion by the boron feedback, and the reactor is tripped upon “high pressurizer pressure” signal. During the steam line break sequence, the affected HCSG depressurizes rapidly and the reactor is tripped upon “low main steam pressure” signal. None of the transients violate safety margins. The adopted 3-D vessel modelling approach and applied multi-physics calculations are able to capture both transient physics within the reactor domain, and conclude that symmetric arrangement of the HCSG tubes enhance coolant mixing and prevent a heterogeneous core power excursion.

Effects of Neutron Flux Distribution and Control Rod Shadowing on Control Rod Calibrations in the Oregon State TRIGA® Reactor

Control rod calibration experiment results for the Oregon State TRIGA® Reactor (OSTR) immediately following LEU conversion in 2008, and MCNP® 5 predicted rod worths from the 2008 LEU Conversion Safety Analysis Report (CSAR) are discussed. The reactivity worth of the four OSTR control rods is measured using the rod-pull method. Reactor power and period measurements in this method rely on the fission chamber power detector on the north side of the reflector. It is proposed that the location of the fission chamber and the neutron flux distribution in the core may result in an inaccurate reactor period measurement due to the asymmetry of the neutron flux distribution in the OSTR core. The asymmetry of the flux is believed to be more pronounced during super-criticality, resulting in errors in the time-of-power-rise measurements. As a result, control rod calibration experiments may under-predict or over-predict the reactivity worth of certain control rods. A time-independent Monte–Carlo method for the quantification of these effects is presented. Thermal flux maps at the core axial mid-plane are obtained from the model to inform discrepancies between predicted and observed results.

Development of the Reliability Assurance Program in a Brazilian nuclear power plant subsidized by a Reliability, Availability and Maintainability Model

The main objective of this work is to present a methodology for the development of a Reliability Assurance Program (RAP) specific to a PWR experimental nuclear installation, through the analysis of the installation and the development of a preliminary RAP subsidized by a Reliability, Availability and Maintainability (RAM) model. The study of an evaluation was carried out in the long-term decay heat removal of the studied experimental plant, whose data were used for application of the RAP. The necessary steps for applying the developed RAP are followed, using the data from the assessment of the studied plant, resulting in a list of components of significant risk for the Program, and in the following steps of sending the list to the experts panel, ranking of SSCs by the panel and development of the final list of significant risk SSCs for using the list in the optimization of the plant. The RAP subsidized by a RAM model will be able to work with the logical relationships between each component of the plant for their effects on energy generation and with the quantitative prediction of the magnitude of each contributor to the occurrence of high-level events, and the developed methodology can be applicable throughout the experimental plant. In this way, it will be possible to implement the RAP in the plant, which will provide a structured way to meet the regulatory requirements for its licensing. Also, it will be possible to complement the plant safety analysis report, which must contain the RAP.

Accident tolerant fuels behavior analysis for Pressurized Water reactor in steady state and transient conditions

Nuclear Fission Technology (NFT) is constantly developing to seek an increasingly safer operation and, consequently, increase the public acceptance of nuclear energy generation. Therefore, the development of new types of nuclear fuels and the search for different materials for cladding fuel rods has been the subject of research mainly to provide better efficiency in to carry energy in nuclear reactors and also to reduce possible damages in accident situations. In this way, this work aimed to compare the thermal response and behavior of Zircaloy 4, SiC reinforced with Hi-Nicalon type S (HNS) and FeCrAl alloy when they were considered as fuel cladding at steady state and transient conditions during operation of a PWR (Pressurized Water Reactor). Two thermal hydraulic models using the RELAP5 MOD3.3 and RELAP5-3D v. 4.3.4 codes and the reference data from the document Final Safety Analysis Report of Angra 2 PWR were used to perform the investigations. The work was developed in two steps. In the first one, the results of the Angra 2 nodalization in steady state conditions were compared with the Angra 2 reference data, obtaining good agreement. In the second step, the thermal response, when the cladding materials were changed, was evaluated in steady state and transient situations. The transients LOFA (Loss of Flow Accident), LBLOCA (Large Break - Loss of Coolant Accident), and coolant channel blockage were simulated in extreme conditions. The results have shown that substituting Zircalloy 4 alloy by FeCrAl or ceramic cladding SiC HNS presented a similar and better thermal response. Moreover, in another simulation, UO2 fuel has been replaced by U3Si2. The results showed lower temperatures at the fuel center in the rods with U3Si2 than those using UO2 as nuclear fuel, demonstrating the advantages of using U3Si2 fuel concerning the radial heat transfer in the rod.

3D Analysis of Hydrogen Distribution and Its Mitigation Using Passive Autocatalytic Recombiners (PARs) Inside VVER-1000 Containment

Hydrogen is a flammable gas that can generate thermal and mechanical loads which could jeopardise the containment integrity upon combustion inside nuclear power plants containment. Hydrogen can be generated from various sources and disperses into the containment atmosphere, mixing with steam and air following a loss of coolant accident and its progression. Therefore, the volumetric hydrogen concentration should be examined within the containment to determine whether a flammable mixture is formed or not. Codes with 3D capabilities could serve this examination by providing detailed contours/maps of the hydrogen distribution inside containment in view of the local stratification phenomenon. In this study, a 3D VVER-1000 as-built containment model was sketched in AutoCAD and then processed into GOTHIC nuclear containment analysis code for hydrogen evaluation. The model was modified to a great extent by installing 80 passive autocatalytic recombiners and locating hydrogen sources to evaluate the performance of the hydrogen removal system inside the containment on maintaining the hydrogen concentration below the flammability limit during a large break loss of coolant accident. 2D profiles and 3D contours of volumetric hydrogen concentration with and without PARs are presented as the simulation outcome of this study. The results were validated against the results of the Final Safety Analysis Report, which also demonstrates the effectiveness of the hydrogen removal system as an engineered safety feature to keep the containment within a safe margin. Detailed 3D contours of hydrogen distribution inside containment can be employed to evaluate the local hot spots of hydrogen, rearranging and optimising the number and location of PARs to avoid the hydrogen explosion inside containment.

Evaluation of the CNESTEN's TRIGA Mark II research reactor physical parameters with TRIPOLI-4® and MCNP

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor keff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the keff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the keff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the keff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the keff is not negligible and should be taken into account when performing S/U analysis.

The large-break LOCA uncertainty analysis in a VVER-1000 reactor using TRACE and DAKOTA

An uncertainty analysis is inevitable in simulating reactor performance, particularly in the accident analysis for the estimation of safety borders. In this paper, the application of the uncertainty analysis of a Large Break Loss of Coolant Accident (LBLOCA) from main coolant pipeline at the reactor inlet and outlet, and also pressurizer (PRZ) connecting pipe are presented. First, the model developed in TRACE is verified in steady-state conditions and LBLOCA using the Final Safety Analysis Report (FSAR) of Bushehr nuclear power plant. By the verified model, the uncertainty analysis is made respecting uncertainty in boundary conditions. Finally, the analysis is achieved using the surrogate models and Wilks’ method in DAKOTA. The 95/95 uncertainty band of the peak cladding temperature (PCT) and main reactor parameters are obtained and analyzed. The results indicate that the PCT uncertainty in LBLOCA at the reactor inlet is large, the PCT varies in the range of 910 to 970 °C. The drop temperature time of fuel cladding is in the range of 270 to 600 s for LBLOCA at the reactor inlet. The PCT uncertainty in LBLOCA at the reactor outlet and PRZ connecting pipe is low.

Identification and tracing of radionuclides in low- and medium-activity liquid radwaste sources of G.A. Siwabessy reactor

Abstract The liquid radioactive waste generated by the G.A. Siwabessy reactor (RSG-GAS) is categorized into low-activity liquid radwaste (LALR) and medium-activity liquid radwaste (MALR). The radionuclide content of both LALR and MALR can use as an indicator of the structural integrity of the reactor’s systems, structures, and components (SSC). To evaluate the degradation of the reactor SSC, the radionuclide species were identified, and their activities were measured using gamma spectroscopy. Based on the identified radionuclides, the process of their formation can be traced. The radionuclides identified in LALR were 24Na, 51Cr, 59Fe, 60Co, 65Zn, and 124Sb, while the radionuclides in MALR were 24Na, 51Cr, 58Co, 59Fe, 60Co, 65Ni, 65Zn, 89Kr, 90Kr, 109Cd, 131I, 132I, 140Ba, 137Cs, 146Ce, and several others. The radionuclides found can be classified into corrosion product activation (60Co, 65Zn, 51Cr, 59Fe, 24Na, 65Ni), topaz impurities activation (51Cr, 59Fe, 60Co, 65Zn), fission product (90Kr, 140Ba, 131I, 137Cs, etc.), and demineralized water impurities activation (51Cr, 59Fe, 65Zn, 60Co, etc.). After comparing the activity value of each radionuclide with the limit value in the safety analysis report document, we can conclude that the activity of each one is below the required level. It can infer that the structural integrity of reactor SSC is still well maintain. During routine monitoring, the radionuclide content in the primary coolant fluctuates depending on the reactor load. The concentration of radionuclides detected varies when a large or small number of research samples are loaded onto the core. Nevertheless, their activities remain within the required safety limits.

Design and Development of Neutronics and Thermal Hydraulics Modeling Code for ACP1000 Nuclear Reactor Dynamics in LabVIEW

An advanced neutronics and thermal hydraulics nuclear code, called GNTHACP code, is designed and developed in LabVIEW as Graphical Neutronics and Thermal Hydraulics toolkit for 1100 MWe Advanced Chinese PWR (ACP-1000) nuclear power plant. The reactor neutronics model is developed using a nonlinear point reactor kinetics model, while, the reactor thermal hydraulics model is developed based on nonlinear fuel and coolant temperature dynamics. The heart of the GNTHACP code is the control rod reactivity model. Control rod reactivity banks are comprised of four power compensation banks G1, G2, N1, N2 and one temperature compensation bank R. The reactivity control configuration of these banks is highly nonlinear, complex and challenging in nature. The control rod reactivity model as a function of G1, G2, N1, N2 and R banks is optimized using two distinct techniques. The control rod reactivity model is optimized using Simplex Linear Programming (SLP) technique under constraints of reactor power as safety limit and control rod speed as maximum speed limit in LabVIEW. The control rod reactivity model is also optimized and investigated using nonlinear Sequential Quadratic Programming (SQP) technique under same constraints in LabVIEW. All the models are integrated and the state-of-the-art virtual instruments (VIs) are designed for cost function optimization, configuring models and calibration of model parameters in LabVIEW. The integrated model as graphical coupled neutronics and thermal hydraulics modeling code is optimized and validated against the Final Safety Analysis Report (FSAR) and different parameters of interest are investigated and proved within design limits as reported with CORCA and CORTH benchmark nuclear codes. The proposed code is stable, highly efficient and accurate as compared to other nuclear codes.

Reliability Assessment of NPP Safety Class Equipment Considering the Manufacturing Quality Assurance Process

Quality and safety are intensely related and go hand in hand. Quality of the safety-grade equipment is very important for the safety of a nuclear power plant (NPP) and achieving production goals. During manufacturing of plant components or equipment, deviation from the design might occur at different stages of manufacturing for various reasons, such as a lack of skilled manpower, deviation of materials, human errors, malfunction of equipment, violation of manufacturing procedure, etc. These deviations can be assessed cautiously and taken into consideration in the final safety analysis report (FSAR) before issuing an operating license. In this paper, we propose a Bayesian belief network for quality assessment of safety class equipment of NPPs with a few examples. The proposed procedure is a holistic approach for estimation of equipment failure probability considering manufacturing deviations and errors. Case studies for safety-class dry transformers and reactor pressurizers employing the proposed method are also presented in this article. This study provides insights for probabilistic safety assessment engineers and nuclear plant regulators for improved assessment of NPP safety.

Investigating the Human Factor in Maritime Accidents: A Focus on Compass-Related Incidents

Accidents at sea can have devastating consequences, and investigating their causes is a complex and rigorous process. One of the most important factors contributing to these accidents is the human factor, which has received considerable attention from researchers in recent years. This article examines the role of the human factor in marine accidents, focusing the use and maintenance of compasses. Using data from scientific journals and safety analysis reports, the frequency and focus of research on this topic is analyzed and areas for further investigation are identified. International regulations, performance standards, and handling requirements relating to compasses are also discussed, and an analysis of compass-related accidents is presented. The survey conducted among seafarers indicates that they are satisfied with the performance of their compass, but there is a need for more advanced compass technologies and training on the proper use and maintenance of compasses. The survey shows that routine maintenance and calibration of compasses is crucial to minimize the possibility of human error and to prevent maritime accidents. In conclusion, we stress the importance of conducting further studies in this area taking into account the human factor in order to improve maritime safety measures.

Design of novel fractional order FPGA based reactor protection and safety controllers for ACP1000 nuclear power plant in LabVIEW

In this research work, an advanced most modern ACP1000 Nuclear Power Plant is addressed. An enhanced fractional order model of ACP1000 nuclear power plant is adopted with an addition of protection and safety systems. The whole plant model is developed by using innovative hybrid technology of Visual Basic, LabVIEW, Fractional Order and Field Programmable Gate Array (FPGA). A reactor trip system is designed and modeled using FPGA technology in LabVIEW. Plant parameters are systematically modeled and panels are designed in LabVIEW for reactor protection controllers. Twenty one reactor trip controllers are designed and modeled based on complex digital logics using FPGA programming in LabVIEW. Two fractional order trip controllers are designed for over temperature protection and over power protection in LabVIEW. FPGA based safety controllers are designed for Engineered Safety Features (ESF) in LabVIEW. For enhanced model of ACP1000 nuclear power plant, 374 systems are modeled in modular form in Visual Basic Environment. Nine process controllers are configured in ANFIS framework in LabVIEW. In this research work, process controllers are used in conjunction with protection and safety controllers using FPGA. The parametric display of simulations is carried out in Visual Basic. The closed loop performance of proposed protection controllers is evaluated under reactor trip and turbine trip while that of safety controllers are evaluated under inadvertent opening of safety valves of pressurizer. Various parameters are simulated for severe transient conditions and the results are evaluated and validated against reference design data and Final Safety Analysis Report (FSAR) of ACP1000 nuclear power plant. All the results are well within the trip and safety systems design bounds under abnormal and severe operating conditions.

Design of Fractional Order Sliding Mode Adaptive Fuzzy Switching Controllers for Uncertain ACP1000 Nuclear Reactor Dynamics

Advanced Chinese Pressurized Water Reactor (ACP1000) is a third generation load following nuclear reactor. ACP1000 is designed to control the reactor power by a sophisticated control rod mechanism under the base load normal operation of a nuclear power plant in Mode-G. To extend the normal operation of ACP1000 for load following condition, boron adjustment control is used in manual configuration. In this research work, model based two new controllers are designed for ACP1000 reactor dynamics. A nonlinear two-point reactor kinetics model is developed for two halves of the reactor core designated as top and bottom of reactor core. Reactor feedbacks model for two-point reactor kinetics model is developed with fuel temperature, moderator temperature, Xenon concentration, G-Bank control rod position, R-Bank control rod position and boron concentration feedbacks under normal operation of ACP1000. Two problems of the large reactor core of ACP1000 are Xenon oscillations and axial offset in core power distribution. To address these problems, two new controllers are designed for normal load following operation of ACP1000. One controller is designed to replace G1-Bank and R-Bank in Mode-G for reactor power control. The second controller is designed to replace G2-Bank in Mode-G for reactivity control and axial power distribution control. Originally, both reactor coolant average temperature controller and reactor power controller were adaptive controllers. Therefore, both new controllers are designed based on an optimized sliding algorithm using a dedicated fractional order sliding mode control oriented adaptive fuzzy logic control (FO-SMC-AFLC) synthesis scheme. The performance of the proposed closed loop controllers is evaluated for design step and ramp power transients. Both proposed controllers are validated against benchmark results reported in Preliminary Safety Analysis Report (PSAR) of ACP1000. The novel control design scheme is proved satisfactory for normal load following operation of ACP1000, and all the results are found well within design limits.

Estimation of public exposure during normal operation of unit-1 Barakah Nuclear Power Plant using GALE and HOTSPOT

In August 2020, the first Barakah Nuclear Power Plant (BNPP) unit was connected to the grid and started supplying electricity. The BNPP site consists of 4 pressurized water reactor units, and each unit generates electricity up to 1400 MWe power. Concerning safety, the radiological assessment of the radioactive release and the corresponding public exposure during normal operation is essential. In the present work, the radioactive release during normal operation, in gaseous form, is calculated considering the operating condition of the BNPP by using the GALE code for gaseous forms. Then, the HOTSPOT code is used to simulate the radiological dispersion, including the public exposure and Total Effective Dose Equivalent (TEDE). HOTSPOT code uses the Gaussian dispersion model to provide near-surface releases, short-range dispersion, and short-term releases. Finally, the results in the present work are compared with the values given in the Final Safety Analysis Report of the BNPP. For typical operating circumstances, this study evaluated gaseous source terms such as iodine, noble gases, radioactive particles, tritium, C-14, and Ar-41 and found that TEDE values are within the authorized limits for distances up to 80 kms from the power plant.

Fractional order modelling and robust multi-model intelligent controllers’ synthesis for ACP1000 nuclear power plant

In this research work, a third generation Pressurized Water Reactor (PWR) type Nuclear Power Plant (NPP) of 1100 MWe rating called Advanced Chinese PWR (ACP1000) NPP is adopted. The research and development work is mainly encompassed of addition of third steam generator loop, hybrid programming platforms, fractional order modular modelling and intelligent controllers. The ACP1000 NPP is comprised of primary loop, secondary loop and balance of plant (BOP). The ACP1000 nonlinear dynamics is precisely modelled with greater efficiency using fractional order differential equations. The entire plant is modelled into eighteen modules in a fractional order framework in Visual Basic (VB) environment. The eighteen modules are governed by nine controllers. Controllers are configured in a fashion of an adaptive neuro-fuzzy inference system (ANFIS). An innovative multi input single output (MISO) ANFIS virtual instrument (VI) is designed for training and configuring controllers in LabVIEW. The developed MISO ANFIS VI in combination is implemented for multi-input multi output (MIMO) ANFIS controllers in parallel computing framework. The variable transfer between fractional order multi- modular framework in VB (FO- MMF-VB) and robust intelligent ANFIS controllers in LabVIEW (RI-ANFIS-LV) is established in client server configuration and system exec VI for calling external code in VB. This hybrid programming framework of Visual Basic and LabVIEW is actually a development of LabVIEW Visual Basic Integration (LabVb) Toolkit used to evaluate the closed loop transient performance of ACP1000 nuclear power plant. Various parameters are simulated for dynamic transient simulations under load manoeuvring from 100% power to 50% power. The results are analysed and validated against benchmark design data and Preliminary Safety Analysis Report (PSAR). All the results are within the protection systems design limits under normal operating conditions.

Application of Best Estimate Plus Uncertainty (BEPU) Methodology in a Final Safety Analysis Report (FSAR) of a Generic Plant

The licensing process of a nuclear power plant is motivated by the need to protect humans and the environment from ionizing radiation and, at the same time, sets out the basis for the design and determining the acceptability of the plant. An important part of the licensing process is the realization of accident analysis related to the design basis, which should be documented in the Final Safety Analysis Report (FSAR). There are different options on accidents calculation area by combining the use of computer codes and data entry for licensing purposes. One is the Best Estimate Plus Uncertainty (BEPU), which considers realistic input data and associated uncertainties. Applications of BEPU approaches in licensing procedures were initiated in the 2000s, first to analysis of Loss of Coolant Accident (LOCA), and then to the accident analysis as a whole, documented in Chapter 15 of the FSAR. This work has as main objective the implementation of BEPU methodology in all analyses contained in FSAR, through the homogenization of the analytical techniques and identification of key disciplines and key topics in the licensing process.

Radiological consequence analysis for hypothetical accidental release from Nigerian Research Reactor-1

Radiological dispersion study is a key element in safety analysis report (SAR) of every nuclear facility for the purpose of emergency response planning. In this work, computational approach was used to determine the total effective dose and ground deposition at critical positions onsite and offsite of the Nigerian Research Reactor-1 (NIRR-1) facility which will be useful in the ongoing development of final SAR for NIRR-1 Low Enriched Uranium (LEU) core. In the methodology used, NIRR-1 LEU core was depleted with TRITON module of SCALE 6.2.3 code and the fission inventory in the core was calculated after a continuous operation at full power of 231.931MWD/MTU for 918 Effective Full Power Days (EFPD) at an operation regime of 3 h per day, 3 days per week and 48 weeks per year. Hot Spot was employed for atmospheric transport and dose calculations with consideration of different accidental scenarios in which 20%, 30%, 60% and 100% gaseous inventory was hypothetically released into the atmosphere. From the results obtained, the total effective dose to maximum exposed workers at 10 m and maximum exposed members of public at 300 m from the reactor were 3.10mSvand0.43mSv respectively for the worst-case scenario with 100% release while the maximum ground deposition was 5.5×106Bq/m2 with corresponding maximum ground shine dose rate of 7.5×10−4mSv/hr. This results are at least one order of magnitude below the dose limits recommended by the International Commission on Radiological Protection (ICRP) and indicate that the present LEU core of NIRR-1 is unlikely to cause any detectable health effect on workers and members of public in the event that 100% of its gaseous inventory is released into air in the environment. Hence it could unequivocally be said that the population is safe from the operation of NIRR-1 in its present location.