Preliminary Safety Analysis Report Research Articles

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.

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.

DEMO – The main achievements of the Pre – Concept phase of the safety and environmental work package and the development of the GSSR

The EUROfusion Safety and Environmental Work Package (WPSAE) has the scope to progress the safety studies for the future EU DEMO reactor. A Generic Site Safety Report (GSSR) has been prepared to include all the steps necessary to cover the safety issues of the nuclear fusion plant, from the definition of principles and requirements, through the detection of the source terms at risk, selecting the postulated initiating events, analyzing the accidents, quantifying the doses to the population and investigating waste production and its management. Eleven GSSR volumes collate the studies performed. The final goal is to prepare safety documentation, as complete as possible, to initiate a preliminary safety report when the plant site is selected.In parallel, the safety studies are supporting the maturation of the design of DEMO, providing feedback on the technical choices for the machine, the selection of materials, the use of space, the equipment necessary to correctly manage the safety risks through continuous collaboration with the design teams of other DEMO Work Packages.The GSSR focused on safety and environmental issues, it did not encompass security issues that could impact the design and hence the safety performance of DEMO. Consideration of the interactions between safety and security will be needed in the development of the DEMO detailed design and its associate Preliminary Safety Analysis Report (PSAR).The main achievements of the Pre-Concept Phase are presented in this paper, by summarizing the contents of GSSR volumes. Completion of the work is foreseen during the future Concept Design Phase (2021–2027).

Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to power light water reactors (LWRs) in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion of U.S. high performance research reactors (USHPRRs) like the MITR. The Preliminary Safety Analysis Report (PSAR) for the MITR LEU conversion has been submitted and currently under review by the U.S. Nuclear Regulatory Commission (NRC). A transition core plan, from 22 fresh LEU fuel elements (i.e., beginning-of-life) gradually to 24 of them arranged in an equilibrium configuration, is expected to serve as an appendix chapter in the final SAR. The current study presents the fuel cycle development, which eventually leads to the transition core plan. The results confirm the equilibrium state, where both the shim bank movement (i.e., core reactivity) as well as the content of fissile materials stabilize, can be achieved by fixed-pattern fuel management. Fission density has been evaluated for a number of fully discharged LEU fuel elements, using a best-estimate approach. The fuel cycle calculations also generate power profiles at each core state. A steady-state thermal-hydraulic safety analysis has thus been performed, where onset of nucleate boiling (ONB) is considered as the safety criterion. The results confirm significant margin to ONB at all analyzed transition and equilibrium fuel cycle states.

Validation of COCOSYS 2.4v4 AIM module on various single effect and integral experiments

Abstract Nuclear power plant containment is the last barrier, as stated in the worldwide accepted in-depth safety principle, between a nuclear reactor and the environment. During various accident scenarios, fission products can exit the fuel and primary circuit, entraining the containment and later the environment. One of the most important fission products exiting the corrupted fuel pins is iodine. Iodine is a strongly bioactive chemical species with a very complex in-containment physical-chemical behavior. The fission product transport inside the containment as well as the release into the environment takes an important role in the preliminary safety analysis report (PSAR). Analyses presented in PSAR are calculated by numerous computational codes. The validity of these results should be checked. The only chance to do so is to validate the computational tool on validation experiments, which represent single effect behavior as well as a complex behavior on integral tests. Validation experiments presented in this paper are from an international Behavior of Iodine Project (BIP) which was conducted under the Organization for Economic Co-operation and Development – Nuclear Energy Agency (OEC-NEA). The single effect tests are illustrated by four representative experiments, namely the G-01 studying the iodine deposition and resuspension on steel coupon in gas phase, the G-04 studying the iodine deposition and resuspension on painted surface in gas phase, the G-06 studying the humidity effect on iodine deposition on painted coupons in gas phase and finally the AECL-2 studying the iodine deposition on painted coupons in water phase. The integral test is represented by the RTF P9T1, where the complex iodine chemistry including the response to pH change is studied. The scope of this paper is to reveal whether the COCOSYS code is capable to bring satisfactory results of a complex in-containment iodine behavior with recommended and basic setup of computational models and parameters.

Component design and accident analysis of fuel assembly for prototype GEN-IV sodium-cooled fast reactor

Abstract The Korea 150 MWe Prototype GEN-IV Sodium cooled Fast Reactor (PGSFR) project has reached the middle of its preliminary safety analysis report (SAR) design phase. The core design studies are being conducted by KAERI. Some design features for the core have been made to comply with the GEN-IV reactor objectives. The main objective to improve the safety of the core design. A negative sodium void worth was achieved thanks to a new fuel assembly design including U-Zr and U-TRU-Zr axially heterogeneous fuel rods, a large cladding/small spacer wire bundle, a sodium bonding between metal slug and cladding, and upper gas plenum above the active core region, and upper/lower reflectors. Finite element calculations were performed to increase the stiffness of the above core load pad and top load pad in order to analyze their effect on the core mechanical behavior during hypothetical radial core flowering and compaction events. More generally, all design choices for the PGSFR have been made with the permanent objective of minimizing the fuel assembly height to decrease the overall costs of the reactor and the fuel cycle. This paper describes the fuel assembly design for the GEN-IV core at the end of the preliminary SAR design phase and studies performed to optimize the performances. The basic design and fuel assembly analysis results were described in this work. In specially, the design analysis results under the accident condition is focused in this paper. As the results of that, the structural integrity of a fuel assembly was satisfied for the PGSFR fuel assembly mechanical design criteria.

Study on the DLOFC and PLOFC accidents of the 200 MWe pebble-bed modular high temperature gas-cooled reactor with TINTE and SPECTRA codes

Validation and verification (V&V) is important and receives significant attention during the design and analysis of the nuclear power plant. As the first commercial modular high temperature gas-cooled reactor (HTR) demonstration plant, during the regulatory review of the preliminary safety analysis report (PSAR) and the final safety analysis report (FSAR) of the 200 MWe Pebble-bed Modular High Temperature gas-cooled Reactor (HTR-PM), the authorities also pay high attention to the code validation and verification.Within the cooperation between the Nuclear Research Group (NRG) of Netherland and Institute of Nuclear and new Energy Technology (INET), Tsinghua University of China, a code to code verification was carried out between the TINTE code, a thermal-hydraulic design and accident analysis tool for the Pebble-bed High Temperature Gas-cooled Reactor (HTGR), and the SPECTRA code, a thermal-hydraulic analysis code developed at the NRG. Three typical accidents scenarios, two depressurized loss of forced cooling (DLOFC) accidents scenarios and one pressurized loss of forced cooling (PLOFC) accident scenario, have been analyzed both with the TINTE code and the SPECTRA code.The simulation results show good agreement between TINTE code and SPECTRA code. The results also indicate that, due to the inherent safety feature of the HTR-PM design, the maximal fuel temperature during above accidents would never exceed the limitation of 1620 °C, below which there is no additional damage for the TRISO particles, and the SiC layer of each particle can guarantee the fission-product-retention capacity. This work will be helpful to the further study of the HTGR, and more code to code verification is planned.

Using FRAPTRAN 1.5 code for safety evaluation of TVS-2006 fuel rod under transient conditions of VVER-1200 (AES-2006) reactor

This article shows the results of important thermal-mechanical parameters related to the TVS-2006 fuel rod design which were analyzed and evaluated by using FRAPTRAN1.5 code. Based on the data given in the Preliminary Safety Analysis Report (PSAR) of AES-2006 (Novovoronezh NPP-2 Power Unit No.1)and FRAP-T (Fuel Rod Analysis Program-Transient), FRAPTRAN1.5 calculations of TVS-2006 fuel rod behaviors in Loss of Coolant Accidents and Reactivity-Initiated Accidents conditions have been made. The calculated results related to safety criteria of fuel element and cladding temperatures, cladding stress and strain, fuel enthalpy, local oxide thickness, gap gas pressure, and elongation of fuel rod have been compared with the ones given in AES-2006 PSAR. A good agreement has been observed between AES-2006 PSAR andFRAPTRAN1.5 calculations.

Verification of TVS-2006 fuel rod design of VVER-AES2006 reactor under steady-state operating condition Using FRAPCON-3.5 code

The purpose of this paper is to discuss the independent verification of TVS-2006 fuel rod design used in VVER-AES2006 reactor (Novovoronezh NPP-2 Power, Unit 1), based on the acceptance criteria and the reference data given in the Preliminary Safety Analysis Report of the State Research, Design, Construction and Survey Institute “Atomenergoproekt” (PSAR) and the operation of VVER-1000 reactor. The calculations were performed using FRAPCON-3.5 code, including fuel temperature, cladding temperature, fission gas release, internal gas pressure, cladding stress and strain, fuel extension, fuel rod elongation, cladding creep rate, fuel swelling rate, cladding oxide thickness and hydrogen concentration. The results are compared with the calculated data using START-3 code in PSAR and the acceptance criteria required by Russian nuclear regulatory body. Despite some discrepancies, the results showed conformance with the calculated data given in the PSAR and meet the acceptance criteria.

The design assessment of limiting transient event by RETRAN-02 and VIPRE-01 for Lungmen ABWR

System thermal hydraulic and hot channel calculations are utilized in this paper to analyze the limiting transient event for Lungmen nuclear power plant (LMNPP). The difference of initial critical power ratio (ICPR) and minimum critical power ratio (MCPR), i.e., ΔCPR, is employed to quantify an impact caused by a transient to LMNPP. Based on the ΔCPR criterion, the Lungmen Preliminary Safety Analysis Report (PSAR) concluded that Load Rejection (LR) event was the most limiting event. However, Inadvertent Closure of One Turbine Control Valve (ICOTCV) event was specified as the most limiting core-wide transient event for LMNPP in the Lungmen Final Safety Analysis Report (FSAR). Moreover, the partial-arc TCV control mode is now applied for linear steam flow control, while the Lungmen FSAR utilized full-arc mode for ICOTCV transient analysis. Based on the above differences, it is necessary to re-analyze both LR and ICOTCV transient events for safety consideration. In this paper, ΔCPRs are calculated by the combination of RETRAN-02 and VIPRE-01 codes. The system transient parameters calculated by RETRAN-02 are furthermore submitted into VIPRE-01 for hot channel analysis. The hot channel model constructed by VIPRE-01 simulates the hottest fuel bundle. The critical heat flux (CHF) is calculated by a CHF correlation. Moreover, the different valve control modes are discussed to study the influence on ICOTCV transient. The analytic results by RETRAN-02 and VIPRE-01 are compared with Lungmen FSAR for assessment. The assessment result indicates that the RETRAN-02 and VIPRE-01 analytic modes applied in this paper are capable of simulating the transient behaviors reasonably. Based on the design changes mentioned above, the calculated maximum ΔCPRs are 0.18 and 0.20 for LR and ICOTCV by RETRAN-02 and VIPRE-01, respectively. It is concluded that, first, the impact caused by ICOTCV event is larger than that by LR event with the new fuel type design and full-arc control mode. Second, the adoption of partial-arc control mode is capable of controlling steam flow rate linearly. However it also increases the ΔCPR during ICOTCV transient. This issue should be highlighted after commercial operation.

The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02

A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using Lungmen startup test data.

Analysis of a Coupled Blowdown from Both RPV and BOP with RELAP5-3D/K for a Limiting FWLB of an ABWR for Containment Design

The limiting blowdown event for the design of an advanced boiling water reactor (ABWR) containment shifts from a conventional recirculation line break to a feedwater line break (FWLB) by implementing reactor internal pumps. As a result, coupled blowdown from both the reactor pressure vessel (RPV) and the balance of plant (BOP) is involved in the limiting FWLB. Coupled blowdown from both RPV and BOP for the FWLB of the Lungmen ABWR has been successfully analyzed using the advanced RELAP5-3D/K code. To simulate adequately both the RPV and BOP blowdown, the essential simulation scope of an ABWR includes the reactor system, the main steam and turbine systems, the condensate and feedwater systems, the protection system, and the emergency core cooling system. As compared to what was presented in the preliminary safety analysis report of the Lungmen ABWR, unexpected prolonged decays of BOP blowdown flow and enthalpy were calculated. The revised blowdown flow and enthalpy calculated by RELAP5-3D/K from both RPV and BOP breaks provide a new and solid basis for the final safety analysis of ABWR containment for the Lungmen plant, which is scheduled for commercial operation in 2011. The successful modeling of the entire RPV and BOP with RELAP5-3D/K and associated application to the FWLB licensing blowdown analysis indicate that the advanced RELAP5 code can extend its traditional reactor safety analysis to the simulation and analysis of the entire power generation and conversion systems.

ITER operating limit definition criteria

The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction. OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent. This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits.

Current status and technical description of Chinese 2 × 250 MW th HTR-PM demonstration plant

After the nuclear accidents of Three Mile Island and Chernobyl the world nuclear community made great efforts to increase research on nuclear reactors and to develop advanced nuclear power plants with much improved safety features. Following the successful construction and a most gratifying operation of the 10 MW th high-temperature gas-cooled test reactor (HTR-10), the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed and designed an HTR demonstration plant, called the HTR-PM (high-temperature-reactor pebble-bed module). The design, having jointly been carried out with industry partners from China and in collaboration of experts worldwide, closely follows the design principles of the HTR-10. Due to intensive engineering and R&D efforts since 2001, the basic design of the HTR-PM has been finished while all main technical features have been fixed. A Preliminary Safety Analysis Report (PSAR) has been compiled. The HTR-PM plant will consist of two nuclear steam supply system (NSSS), so called modules, each one comprising of a single zone 250 MW th pebble-bed modular reactor and a steam generator. The two NSSS modules feed one steam turbine and generate an electric power of 210 MW. A pilot fuel production line will be built to fabricate 300,000 pebble fuel elements per year. This line is closely based on the technology of the HTR-10 fuel production line. The main goals of the project are two-fold. Firstly, the economic competitiveness of commercial HTR-PM plants shall be demonstrated. Secondly, it shall be shown that HTR-PM plants do not need accident management procedures and will not require any need for offsite emergency measures. According to the current schedule of the project the completion date of the demonstration plant will be around 2013. The reactor site has been evaluated and approved; the procurement of long-lead components has already been started. After the successful operation of the demonstration plant, commercial HTR-PM plants are expected to be built at the same site. These plants will comprise many NSSS modules and, correspondingly, a larger turbine.

Review of accidental safety studies for the European HCPB test blanket system

This paper presents a review of safety studies for accidental sequences in the European solid breeder test blanket module (TBM) system. These studies are the starting point for the Preliminary Safety Analysis Report of ITER, under preparation to get the construction permit first and then later the operation licence. In general the reduced inventory of activation products and tritium associated with the TBM system makes the impact of this test system almost negligible on the overall safety risk of ITER. Nevertheless, the possibility of jeopardizing the ITER safety concept has been analysed in connection to the consequences of specific accident sequences, e.g. the pressurization of the vacuum vessel due to the He coolant blow-down, the hydrogen production from the Be-steam reaction, the possible interconnection between the port cell and the vacuum vessel causing air ingress and the necessity to assure heat removal in the short and long periods. In the frame of this assessment, three LOCA sequences have been selected as representative of accidents judged to cover all scenarios envisaged in Cat II to IV events involving the TBM, namely, in-vessel LOCA, ex-vessel LOCA and in-box LOCA.

Full-Load-Rejection Design Analysis for the Lungmen Advanced Boiling Water Reactor Using RETRAN

The very first work on RETRAN model development and its application to plant design study for the Lungmen nuclear power station (LMNPS) is presented. Lungmen is the fourth nuclear power plant of the Taiwan Power Company (TPC). LMNPS has two advanced boiling water reactor (ABWR) units, each with a thermal power of 3926 MW. The preliminary safety analysis report (PSAR) of LMNPS is currently under review. The Lungmen RETRAN-02/MOD5 model was developed to provide support to TPC in the PSAR review and system design study. An analysis of generator load rejection with failure of all bypass valves was performed against the analysis in the PSAR to benchmark the Lungmen RETRAN model. One of the specific designs of LMNPS is that the reactor has the capability to withstand a full-load rejection or a turbine trip event without a reactor scram. An analysis of generator load rejection with all bypass valves was done to evaluate this design. The results show that this design is appropriate.

Overview of the Preliminary Safety Analysis of the National Ignition Facility

The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 individual laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a low hazard, radiological facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis report be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A Preliminary Safety Analysis Report (PSAR) has been approved, which documents and evaluates the safety issues associated with the construction, operation, and decommissioning of the NIF.

Performance Analysis of a Repository for Low and Intermediate Level Reactor Waste

AbstractIn Finland, utilities producing nuclear energy are responsible for the management of the radioactive waste, including final disposal. As regards low and intermediate level waste, the approach has been adopted to employ the power plant sites for locations of repositories. The repositories will be excavated at the depth of about 50 to 125 m in the bedrock of the two Finnish nuclear power plant sites, Loviisa and Olkiluoto.The performance analysis presented in this paper has been carried out for the Preliminary Safety Analysis Report (PSAR) of the Olkiluoto reposi- tory. A flexible model has been developed to estimate the release of radio- nuclides from waste packages and their subsequent transport through the engi- neered barriers in the repository. Gradual degradation of the engineered barriers is accounted for by altering parameters at fixed time points. Safety margins of the disposal concept have been evaluated by including disturbed evolution scenarios in the analysis.