Effective Full Power Years Research Articles

Evaluation of neutron radiation damage in the VVER-1200 reactor pressure vessel

In nuclear reactors, the structural materials of reactor pressure vessel (RPV) are damaged by the radiation that produced by fission reactions. Neutron irradiation damage is one of the most critical factors for the degradation of the reactor pressure vessel (RPV). A displacement per atom (dpa) is a standard measurement for calculating neutron-and gamma-induced radiation damage in materials. The aim of this paper was to evaluate the neutrons radiation damage in the reactor pressure vessel of VVER-1200. In this paper, neutron damage factors, including fast neutron fluxes above 1 MeV, above 0.5 MeV, above 0.1 MeV, the damage rate (dpa/s), and Integrated damage [DPA for one effective full power year (EFPY)] were investigated for VVER-1200/AES-2006 reactor pressure vessel (RPV). The calculations are performed by using NJOY code together with the MCNP6 code at beginning of Cycle (BOC). The cross section library used in present calculations is based on ENDF/B-VII.1 library (ENDF71x). The NJOY code was used for generating DPA cross sections in an ACE formatted file that suitable for using in the MCNP6 code. The MCNP6 code was used for simulating the VVER-1200 reactor core and to calculate neutron DPA rates. MCNP6 simulation of the VVER-1200 reactor core identified the area where the RPV neutron radiation is maximized. The integrated damage for a full effective year was evaluated to be 2.128E-04 DPA/EFPD for the VVER-1200 RPV.

Feasibility Study on the Application of Boron Carbide for Long Term Reactivity Control in the LOTUS Small Fast Reactor

Abstract 200 MWth lead-cooled fast reactor (LOTUS) reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at Vietnam National University (VNU) University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refueling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the startup core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers are withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refueling. Good neutronic behavior of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add to the burnable poison rods.

Conceptual design of the double tube burnable poison for the next generation small modular reactor

In this study, a new application of an innovative burnable poison (BP), which is called Double Tube Burnable Poison (DTBP), is introduced. Two design cincepts of the DTBP were tested in a small modular reactor (SMR) to achieve a longer operating cycle with low soluble boron concentration. These two concepts can be loaded in different locations in the fuel assembly, which increases the degree of freedom of nuclear design. The flexibility of this design allows it to increase its poison effect with the ability to control the depletion speed of its absorber materials. The DTBP pins were combined with the Erbia pins to achieve a flat k-infinite letdown curve with a significant reduction of the excess reactivity. The assembly and core calculations were performed by DeCART2D and MASTER codes. The results showed that the combination of DTBP and Erbia core achieved a cycle length of more than 3.5 Effective Full Power Years. Also, the highest value of Critical Boron Concentration was reduced to 309 ppm within the design limitations of the moderator temperature coefficient and the maximum pin power peaking. The performance of this option is the best among the other options including the conventional BP designs.

Analysis of fuel performance under normal operation conditions of MicroURANUS: Micro long-life lead-bismuth-cooled fast reactor

An innovative ultra-long-life lead-bismuth eutectic (LBE)-cooled fast reactor called MicroURANUS has been designed to prevent severe accidents. It utilizes the remarkable natural circulation capability of the LBE coolant. Furthermore, it can be operated at 60 MWth without refueling for 30 effective full-power years (EFPY). In this study, the thermal and mechanical performance of the highest linear power rod of MicroURANUS was evaluated by modifying the fuel performance code FRAPCON-4.0, which is the normal operation fuel performance code for light water reactors (LWRs). A plutonium generation model was added to consider changes in the physical properties of the UO2 fuel. Furthermore, mechanical and oxidation models of 15–15Ti were added, and the coolant models were modified for liquid metal. According to the unique design of MicroURANUS, a maximum low linear power density of 12.5 kW/m ensures that the highest linear power rod can operate below the safety limit. A low fuel temperature provides a large safety margin for fuel melting, as well as low-pressure build-up fission gas release. In addition, low inlet and outlet coolant temperatures of 250 and 350°C cause the cladding to display a low degree of swelling (a maximum diametral strain of 1.5%) while maintaining high mechanical integrity with negligible cumulative damage fraction (CDF). The modified fuel performance code for lead-cooled fast reactors exhibited the capability to be utilized for fuel performance evaluation and design feedback of MicroURANUS.

Brittle fracture initiation in decommissioned boiling water reactor pressure vessel head weld

The brittle fracture initiation behavior and fracture toughness in the ductile-to-brittle transition region for a thermally-aged weld metal of a decommissioned reactor pressure vessel head (in operation at 288 °C for 23 effective full power years) and in the non-aged reference condition were investigated. The results show that brittle fracture initiated primarily from non-metallic inclusions. The correlation between fracture toughness and brittle fracture initiation type (inclusion debonding or breakage), initiator size, initiation location (as-welded or re-heated regions in the weld metal) were analysed. Despite that thermal ageing does not affect significantly the fracture toughness, it could promote the debonding as a brittle fracture primary initiation type. The influence of debonded inclusion on the evolution of cumulative damage and brittle fracture initiation was assessed using crystal plasticity modelling.

A way Analyzing Oxide Layer on an Irradiated CANDU-PHWR Pressure Tube Using an EPMA and X-ray Image Mapping

The oxide layer in samples taken from an irradiated Zr-2.5Nb pressure tube from a CANDU-PHWR reactor was analyzed using electron probe microanalysis (EPMA). The examined tube had been exposed to temperatures ranging from 264 to 306 ℃ and a neutron fluence of 8.9 × 1021 n/㎠ (E > 1 MeV) for the maximum 10 effective full-power years in a nuclear power plant. Measuring oxide layer thickness generally employs optical microscopy. However, in this study, analysis of the oxide layer from the irradiated pressure tube components was undertaken through X-ray image mapping obtained using EPMA. The oxide layer characteristics were analyzed by X-ray image mapping with 256 × 256 pixels using EPMA. In addition, the slope of the oxide layer was measured for each location. A particular advantage of this study was that backscattered electrons and X-ray image mapping were obtained at a magnification of 9,000 when 20 kV volts and 30 uA of current were applied to radiation-shielded EPMA. The results of this study should usefully contribute to the study of the oxide layer properties of various types of metallic materials irradiated by high radiation in nuclear power plants.

MicroURANUS : Core design for long‐cycle lead‐bismuth‐cooled fast reactor for marine applications

A core design of MicroURANUS, a long-cycle lead-bismuth-cooled fast nuclear reactor for marine applications, is presented. It aims to generate a power of 60MWth, which can be regulated during operation. MicroURANUS was designed to achieve a small burnup reactivity swing for 30 effective full-power years of a lifetime without refueling. To attain these goals, a unit cell study with uranium oxide fuel was initiated to lay a solid foundation for core design, owing to size constraints. A reflector optimization was also performed to minimize neutron leakage. MicroURANUS adopts onion zoning constructed by two enrichment zones for flattening the power and lengthening the core lifetime. The coolant is driven by electromagnetic pumps to achieve inlet and outlet temperatures of 250°C and 350°C, respectively. MicroURANUS analyses were performed using the Argonne Reactor Computation suite and the Monte Carlo code MCS. Core performance features were analyzed for criticality, power profiles, fuel isotope mass inventory, reflector coefficients, reactivity feedback coefficients, and shutdown margin. Additional thermal-hydraulic calculations were performed to confirm that the fuel and cladding temperatures were within the acceptable range. Furthermore, a load follow analysis using the quasi-static reactivity balance method confirmed the feasibility of regulating power by adjusting the inlet coolant temperature.

DESIGN OPTIMIZATION OF THE HOLOS-QUAD MICRO-REACTOR CONCEPT

The Holos-Quad micro-reactor concept is proposed by HolosGen LLC to generate 22 MWt with a lifetime of approximately 8 effective full power years (EFPYs) for civilian applications. The design is based on a very innovative high-temperature gas-cooled reactor concept using four Subcritical Power Modules (SPMs) that fit into one commercial 40-foot transport ISO container. A rigorous design approach was employed in order to ensure that all input parameters were fully investigated and the best solutions possible were considered. The first step of this approach consisted of properly defining the Holos-Quad design problem by identifying the design objectives, the operational constraints, and the input parameters. In the second step, a sensitivity analysis was performed to enable a preliminary investigation of the input parameters to identify the correlations among input and output variables. Finally, the design optimization was performed in the third step, employing a genetic algorithm to effectively explore the highly constrained input parameters and find global optimal solutions. The multi-objective optimization identified various core solutions that would be optimum solutions for different types of applications. For applications where economics matters less and the ease of transportation matters more, a core weight of ~15 tons could achieve a lifetime of ~3.5 EFPYs. For applications where economics matters more and the ease of transportation matters less, a lifetime of 8.3 EFPYs could be achieved with a total core weight of ~26.7 tons.

Low temperature overpressurization analysis for CPR1000

An additional low temperature overpressure protection system which relies on the Pressurizer (PRZ) pressure relief valves has been put forward under the situation in which the Residual Heat Removal System (RHRS) is unavailable or isolated. Taking into account the risk of Reactor Pressure Vessel (RPV) brittle fracture and LOCA under cold overpressurization transients, it is urgent to strengthen the low temperature overpressure protection for in-service Pressurized Water Reactors (PWR) in China. In this paper, the simulation and analysis of two types of overpressure transients at low temperature during the shutdown process are carried out through the detailed Relap5 modeling of CPR1000. The results show that the place where the brittle fracture occurs firstly is the lower plenum. The sensitivities of the system temperature and the opening/closing time of the additional protection system's valve are analyzed. The system set point becomes more restrictive at lower coolant temperature. Additional threshold values of the PRZ pressure relief valve are given under different Effective Full Power Years (EFPY). The set point should be less than 5.5 MPa in the case of 5 EFPY and less than 3.7 MPa at 10 EFPY.The results can be applied to engineering practice, which can effectively improve the safety of CPR1000.

NEUTRONIC BENCHMARK ON HOLOS-QUAD MICRO-REACTOR CONCEPT

The Holos-Quad micro-reactor concept is proposed by HolosGen LLC for civilian applications to generate 22 MWt with a lifetime of approximately 8 effective full power years (EFPYs). The design is based on a very innovative high-temperature gas-cooled reactor concept using four Subcritical Power Modules (SPMs) that fit into one commercial 40-foot transport ISO container. Neutronics benchmarks were developed based on a preliminary version of the Holos-Quad design to confirm feasibility of the neutronics design of this non-traditional high temperature gas-cooled (HTGR)-type micro-reactor concept. Calculations were performed using Monte Carlo codes (SERPENT and OpenMC) as well as the PROTEUS high-fidelity deterministic code for two exercises based on a unit cell model and a full core model. The results obtained showed good agreement with all of the evaluated parameters for the unit cell problem and the full core problem. SERPENT and OpenMC display consistently good agreement in eigenvalue within 150 pcm for the unit cell benchmark and less than 270 pcm for the full core benchmark. PROTEUS eigenvalues showed relatively larger differences but still reasonable agreement with the SERPENT solutions. For the depletion benchmark, the observed eigenvalue differences between SERPENT and OpenMC were within 300 pcm throughout the depletion up to 100 GWd/MT, ensuring that the two codes used equivalent input parameters such as the recoverable heat values for fission reactions. This benchmark exercise confirms the neutronic feasibility and core performance of the preliminary Holos-Quad design.

SPARK-NC: A Lead-Bismuth-Cooled Small Modular Fast Reactor with Natural Circulation and Load Following Capabilities

In this study, a conceptual design was developed for a lead-bismuth-cooled small modular fast reactor SPARK-NC with natural circulation and load following capabilities. The nominal rated power was set to 10 MWe, and the power can be manipulated from 5 MWe to 10 MWe during the whole core lifetime. The core of the SPARK-NC can be operated for eight effective full power years (EFPYs) without refueling. The core neutronics and thermal-hydraulics design calculations were performed using the SARAX code and the natural circulation capability of the SPARK-NC was investigated by employing the energy conservation equation, pressure drop equation and quasi-static reactivity balance equation. In order to flatten the radial power distribution, three radial zones were constructed by employing different fuel enrichments and fuel pin diameters. To provide an adequate shutdown margin, two independent systems, i.e., a control system and a scram system, were introduced in the core. The control assemblies were further classified into two types: primary control assemblies used for reactivity control and power flattening and secondary control assemblies (with relatively smaller reactivity worth) used for power regulation. The load following capability of SPARK-NC was assessed using the quasi-static reactivity balance method. By comparing three possible approaches for adjusting the reactor power output, it was shown that the method of adjusting the coolant inlet temperature was viable, practically easy to implement and favored for the load following operation.

Conceptual design and analysis of the HCCR breeder blanket for the K-DEMO

A helium-cooled ceramic reflector (HCCR) blanket was studied as a candidate breeder blanket for the Korean DEMO reactor. A detailed three-dimensional neutronics model based on a K-DEMO model and a simplified thermal-hydraulic model were established for the HCCR-DEMO blanket conceptual design analysis. The insertion of a graphite reflector resulted in more than 33 % reduction of the required total beryllium multiplier volume, while its influence on the tritium breeding ratio (TBR) was evaluated to be less than 2 %. The final overall TBR of the HCCR-DEMO blanket was estimated as being more than 1.15 even after operation of 20 effective full power years (EFPYs) without replacement.

Neutronics design study of an advanced lead-cooled modular nuclear reactor (ALMANAR)

A design of an advanced small modular reactor with long-life core is being studied in this paper. The core is designed to produce 45 MWth power with a lifetime of 20 years without refuelling. In order to achieve a compact design and have a good neutron economy, lead is considered as the coolant due to its excellent neutronics and thermo-physical properties. However, the lead coolant speed in the core is limited to 2 m/s to minimise the corrosion and erosion of the structural materials. On the other hand, U15N is used as the fuel which reflects excellent thermophysical properties and compatibility with lead. In this study, the neutronics properties of the core including the reactivity evolution during its lifetime, the control rod worth, and the fuel and coolant reactivity feedbacks are evaluated. It was found that ALMANAR could achieve a long-life core of about 22 effective full power years with very excellent inherent safety features. Monte Carlo Serpent code is used to perform the calculations in conjunction with the latest nuclear data library ENDF/B-VIII.0.

Neutronics design study of an advanced lead-cooled modular nuclear reactor (ALMANAR)

A design of an advanced small modular reactor with long-life core is being studied in this paper. The core is designed to produce 45 MWth power with a lifetime of 20 years without refuelling. In order to achieve a compact design and have a good neutron economy, lead is considered as the coolant due to its excellent neutronics and thermo-physical properties. However, the lead coolant speed in the core is limited to 2 m/s to minimise the corrosion and erosion of the structural materials. On the other hand, U15N is used as the fuel which reflects excellent thermophysical properties and compatibility with lead. In this study, the neutronics properties of the core including the reactivity evolution during its lifetime, the control rod worth, and the fuel and coolant reactivity feedbacks are evaluated. It was found that ALMANAR could achieve a long-life core of about 22 effective full power years with very excellent inherent safety features. Monte Carlo Serpent code is used to perform the calculations in conjunction with the latest nuclear data library ENDF/B-VIII.0.

Small modular reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part II: whole-core analysis

Civil marine reactors face a unique set of design challenges. These include requirements for a small core size and long core lifetime, a 20% cap on fissile loading, and limitations on using soluble neutron absorbers. In this reactor physics study, we seek to design a core that meets these requirements over a 15 effective full-power-years (EFPY) life at 333 MWth using homogeneously mixed all-UO2 and micro-heterogeneous ThO2-UO2 duplex fuels. In a companion (Part I) paper, we found assembly designs using 15% and 18% 235U for UO2 and duplex fuels, respectively, loaded into 13 × 13 pin arrays. High thickness (150 μm) ZrB2 integral fuel burnable absorber (IFBA) pins and boron carbide (B4C) control rods are used for reactivity control. Taking advantage of self-shielding effects, these designs maintain low and stable assembly reactivity with little burnup penalty.In this paper (Part II), whole-core design analyses are performed for small modular reactor (SMR) to determine whether the core remains critical for at least 15 EFPY with a reactivity swing of less than 4000 pcm, subject to appropriate constraints. The main challenge is to keep the radial form factor below its limit (1.50). Burnable poison radial-zoning is examined in the quest for a suitable arrangement to control power peaking. Optimized assemblies are loaded into a 3D reactor model in PANTHER. The PANTHER results confirm that the fissile loadings of both fuels are well-designed for the target lifetime: at the end of the ∼15-year cycle, the cores are on the border of criticality. The duplex fuel core can achieve ∼4% longer core life, has a ∼3% lower initial reactivity and ∼30% lower reactivity swing over life than the final UO2 core design. The duplex core is therefore the more successful design, giving a core life of ∼16 years and a reactivity swing of less than 2500 pcm, while satisfying all the neutronic safety parameters. In particular, one of the major objectives of this study is to offer/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to understand the underlying physics of the duplex fuel and ‘open the option’ of designing the functional cores with both the duplex and UO2 fuel cores.

Small modular reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part I: Assembly-level analysis

In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. In this reactor physics study, we seek to design a soluble-boron-free (SBF) and low-enriched uranium (LEU) (<20% 235U enrichment) civil nuclear marine propulsion small modular reactor (SMR) core that provides at least 15 effective full-power-years (EFPY) life at 333 MWth using 18% 235U enriched micro-heterogeneous ThO2-UO2 duplex fuel and 15% 235U enriched homogeneously mixed all-UO2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements.The assembly-level behaviours of candidate burnable poison (BP) materials and control rods are investigated. We examine gadolinia (Gd2O3), erbia (Er2O3) and ZrB2 integral fuel burnable absorber (IFBA) as BPs. We arrive at a design with the candidate fuels loaded into 13 × 13 assemblies using IFBA pins for reactivity control. Taking advantage of self-shielding effects, this design maintains low and stable assembly reactivity with relatively little burnup penalty. Thorium-based duplex fuel offers better performance than all-UO2 fuel with all BP options considered. Duplex fuel has ∼20% lower reactivity swing and, in consequence, lower initial reactivity than all-UO2 fuel. The lower initial reactivity and smaller reactivity swing make the task of reactivity control through BP design easier in the thorium-rich duplex core. For control rod design, we examine boron carbide (B4C), hafnium, and Ag-In-Cd alloy. All the candidate materials exhibit greater rod worth for the duplex design. For both fuels, B4C has the highest rod worth. In particular, one of the major objectives of this study is to offer/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to observe the neutronic performance of the proposed duplex fuel with respect to the UO2 fuel and ‘open the option’ of designing the functional cores with both the duplex and UO2 fuel cores.A companion paper will examine key physics and core safety analysis parameters in the whole-core environment.

SPARK-LS: A novel design of a compact liquid chloride salt cooled fast reactor based on tube-in-duct fuel concept

A novel design of a compact liquid chloride salt cooled fast reactor (SPARK-LS) was proposed in the present study as a distributed energy resource for easy transportation and convenient deployment in diverse applications, such as electricity generation and district heating. The SPARK-LS reactor with rated power of 40 MWth can operate for 10 effective full power years without refueling. An innovative tube-in-duct fuel concept with coolant inside the tubes and fuel outside was adopted to achieve higher fuel volume fraction and lower pressure drop when compared with traditional pin-type fuel. Neutronics and thermal-hydraulics design and preliminary safety analysis were accomplished to obtain an optimized core design by using the advanced reactor design and analysis code system SARAX. Titanium was adopted as the reflector material due to its low density, good corrosion resistance, large strength-to-density ratio and decreased albedo with the neutron spectrum hardening. The optimized core design featured with compact volume and light weight was proved to satisfy the design constraints of peak coolant temperature, peak fuel centerline temperature and coolant pressure drop. The negative coolant density reactivity coefficient in the full core was ensured during the whole core lifetime thanks to the compact core design and the adoption of the titanium reflector. Based on the quasi-static reactivity balance method, five anticipated transients without scram were analyzed to confirm the inherent safety feature of SPARK-LS. It was shown that passive shutdown could be achieved safely for all the transients and the residual decay heat after the shutdown could be removed automatically by the natural circulation of coolant.

Coupled neutronic/thermal-hydraulic hot channel analysis of high power density civil marine SMR cores

Core average power density of standard small modular reactors (SMR) are generally limited to 60–65 MW/m3, which is 40% lower than for a standard civil PWR in order to accommodate better thermal margins. While designing a SMR core for civil marine propulsion systems, it is required to increase its power density to make more attractive for future deployment. However, there are obvious thermal-hydraulic (TH) concerns regarding a high power density (HPD) core, which needs to be satisfied in order to ensure safe operation through accurate prediction of the TH parameters.This paper presents a coupled neutronic/thermal-hydraulic (TH) hot channel analysis of a HPD 375 MWth soluble-boron-free PWR core using 19.25% 235U enriched micro–heterogeneous ThO2-UO2 duplex fuel and 16% 235U enriched homogeneously mixed all-UO2 fuel with a 15 effective full-power-years (EFPY) core life. To perform this analysis the hybrid Monte Carlo reactor physics code MONK is coupled with sub-channel analysis TH code COBRA-EN. This approach is used to investigate the feasibility of different HPD marine PWR concepts and to identify the main TH challenges characterising these designs. To design HPD cores of between 82 and 111 MW/m3, three cases were chosen by optimizing the fuel pin diameter, pin pitch and pitch-to-diameter ratio. These cases have been studied to determine whether TH safety limits are satisfied by evaluating key parameters, such as minimum departure from nucleate boiling ratio, surface heat flux, critical heat flux, cladding inner surface and fuel centreline temperatures, and pressure drop. The results show that it is possible to achieve a core power density of 100 MW/m3 for both the candidate fuels, a ∼50% improvement on the reference design (63 MW/m3), while meeting the target core lifetime of 15 EFPY and remaining within TH limits. The size of the pressure vessel can therefore be reduced substantially and the economic competitiveness of the proposed civil marine PWR reactor core significantly improved.

Radiation damage to Ni-based alloys in Wolsong CANDU reactor environments

Abstract Radiation damage due to neutrons has been calculated in Ni-based alloys in Wolsong CANDU reactor environments. Two damage parameters are considered: displacement damage, and transmutation gas production. We used the SPECTER and SRIM computer codes in quantifying radiation damage. In addition, damage caused by Ni two-step reactions was considered. Estimations were made for the annulus spacers in a CANDU reactor that are located axially along a fuel channel and made of Inconel X-750. The calculation results indicate that the transmutation gas production from the Ni two-step reactions is predominant as the effective full power year increases. The displacement damage due to recoil atoms produced from Ni two-step reactions accounts for over 30% out of the total displacement damage.

Development and application of safety assessment system in PFM analysis for RPV to study the influence of the neutron fluence estimate

The reactor pressure vessel (RPV) is the most important core container for the power generation system of a nuclear power plant. However, the vessel shell is prone to radiation embrittlement because of the operational environment of the power generation process which may lead to crack growth in the shell wall. For the purpose of estimating and ensuring the integrity of the RPV structure, the risk to and reliability of the structure after a period of operation needs to be analyzed under a variety of transient loading conditions.Probabilistic fracture mechanics (PFM) is a new method of analysis applied in a number of studies for estimating the safety of component structures. However, it is difficult for power plant technicians and engineers to evaluate the influence of circumstances arising under various accidental situations on the RPV by PFM analysis because of the lack of relevant experience. Therefore, this study first develops an analysis system which integrates the Oak Ridge Nuclear Laboratory's (ORNL's) PFM FAVOR computer code and analysis models of Taiwan's domestic RPVs. This system may offer engineers and researchers the ability to estimate and analyze the risk and reliability of both boiling water reactors and pressurized water reactor vessels quickly and correctly.Secondly, the system is applied to study the influence of the neutron fluence and sampling conditions on the rupture probability of the RPV at the end of license including 36 and 54 effective full power years (EFPY). The four types of methods used for carrying out the PFM analysis are the Extrapolation, Average, NLR2 and NLR1 methods, and four representative pressurized thermal shock (PTS) transients were considered as loading conditions. The results show that a lower rupture probability is obtained when using the non-linear regression methods to predict the neutron fluence and assess the RPV integrity. Furthermore, the difference in the resulting values will rise with increasing the operation period. Under the same conditions, the rupture probability sequence by size is Average > Extrapolation > NLR2 > NLR1, espacially at 54 EFPY.