Feedwater Line Research Articles

Application of piping failure database to nuclear safety issues in Korea

This paper describes analyses and application of OPDE (OECD Piping Data Failure Exchange) database and NPFD (Nuclear Piping Failure Database) to some safety related issues for Korea nuclear power plants. Safety issues considered in this study include the piping rupture frequency assessment, the screening criteria of the leak before break application, and the augmented in-service inspection and the risk-informed ISI (RI-ISI). As a result of this study: (1) Piping rupture frequencies of the very small LOCA, feedwater line break events, and flood events were evaluated and used as initiating event frequencies for the PSA of Korean PWR plants; (2) Based on the review of OPDE DB and LBB screening criteria, LBB application to the pressurizer surge line, safety injection system, and shutdown cooling system were approved under the conditions that detailed thermal stress analyses should be performed to the systems. (3) For RI-ISI application, major degradation mechanisms in the reactor coolant system were analyzed based on the analysis of the OPDE and NPFD and augmented inspections were enforced to the pipe segments susceptible to degradation mechanism such as thermal stratification/fatigue.

Flow accelerated corrosion: Experience from examination of components from nuclear power plants

Flow accelerated corrosion (FAC) has affected a large number of components made from carbon steel in the secondary as well as primary circuits of light water reactors worldwide. This is primarily a corrosion process enhanced by (electro) chemical dissolution and mass transfer, rather than a mechanical process. It has caused rupture of all sizes of pipelines as well as other components carrying either single phase water or two phase (wet steam) flow. It is reported to be the only mechanism that has significant potential for large leaks in the secondary circuit. This paper describes the observations from examination of a number of components that were affected by FAC in Pressurized Heavy Water Reactors in the past 5 years. Failure analysis done on the 10% feed water line, non-return valve (NRV) in the auxiliary feed water system and on the affected pipeline components from the secondary water system of various reactors has been described in this paper. The surface features observed (the signature pattern of FAC) on the inner surfaces of the pipelines affected by FAC in the primary and the secondary circuits of nuclear reactors have been described. Signature patterns of FAC, on components severely affected by single phase FAC (scallops) and on components affected by dual phase FAC (tiger striping) are recorded and reported. Components where FAC degradation had just initiated showed that the patterns were not well defined. The signature patterns became evident only after the degradation had occurred to a large extent. FAC like features were not seen in stainless steel components in the secondary side. Case studies showing that thinning of carbon steel components is not always attributable to FAC in the secondary circuit of the reactors are also described and reasons of these thinning illustrated to show the importance of establishing the cause of each thinning case. The paper also discusses if the size of the scallops is possible to be correlated to the FAC rate.

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 °C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 °C). Additionally, the peak drywell temperature of 155.3 °C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 °C, which is below the pool temperature used for evaluating the net positive suction head of pumps of the RHR system and the Emergency Core Cooling Systems (96.7 °C). The peak containment pressure and temperature are well below the design value (386.1 kPaG and 171.1 °C). Containment integrity of Chinshan Plant can be maintained under the SPU condition.

Experimental study on the blowdown load during the steam generator feedwater line break accident in the evolutionary pressurized water reactor

Integral effect tests using the ATLAS facility were performed to obtain the thermal-hydraulic parameters such as dynamic and static pressures, local temperatures, and flow rates during a feedwater line break of a steam generator. The break of a feedwater line was simulated using a double rupture disc assembly in order to satisfy the requirements for the break opening time of around a few milliseconds. In the present study, estimated break opening time was less than 1.5ms and broken areas were 48.1% and 93.4% of the feedwater line, respectively. The maximum dynamic pressures of about 1.57bar were obtained inside of feedwater box that was closest to the break location of the feedwater line. After the break of the feedwater line, propagation of the pressure wave along the distance from the break location inside the steam generator was clearly and pertinently observed in all the tests. From a structural integrity point of view, however, the risk induced by this maximum dynamic load could be treated to be insignificant.Based on the present experimental data, an analysis was performed using the best-estimate thermal-hydraulic system code, MARS 3.1 with an aim of qualitatively validating the test results and also evaluating the blowdown load prediction capability of the best-estimate system code in the simulation of steam generator feedwater line break. Calculation results on overall trends of the dynamic pressure behavior are comparable with the test results which indicate the feasibility of dynamic load prediction using the MARS code. As the break opening time became shorter and the broken area became larger, the dynamic pressures were calculated to be significant values in the sensitivity analyzes.

Probabilistic safety assessment of components

Obviously, precautions against damage of components by operational and exceptional loads have to be taken. This is performed according to the procedures and standards of the nuclear regulations, e.g. by using the strength and fracture mechanical parameters (warranty parameters) of the involved materials. Within this integrity proof by calculation, however, the effects of the uncertainties in the input parameters cannot be evaluated quantitatively. By considering the possible parameter scatter in the geometrical dimensions and loads while performing the integrity proof, the failure probabilities, and hence the caused uncertainties, become quantifiable. In addition, postulated flaw sizes can be considered and the effects of operational measures can be evaluated. This requires, on one hand, exact knowledge about the scatter of the decisive parameters, and on the other hand knowledge about their effects on the employed methods, and hence on the result of the calculation. In this paper the applied reliability theory is described. Furthermore, the failure probabilities, especially of the weld joints of a feed water line, were calculated and the available partial safety factors were analysed.

Assessment of Natural Circulation Interruption during Asymmetric Cooldown Transients

This paper presents assessment results for the natural circulation interruption (NCI) phenomenon during the cooldown phase in a nuclear pressurized water reactor. This phenomenon could take place because of several circumstances, such as an asymmetric cooldown after the loss of the forced primary flow. Under NCI conditions, the homogeneous boration of the reactor coolant system (RCS) and the connection of the RCS to the residual heat removal system could be hindered. Moreover, at very low or no primary flow rates and an operating safety injection system, a pressurized thermal shock could occur in the reactor vessel due to cold fluid stratification in the loops. It is therefore important to understand the cause of loop flow stagnation and to derive accordingly the appropriate operator actions to avoid such a phenomenon.The main goal of the current study is to assess the effect of a cooldown strategy upon the single-phase NCI occurrence. For this purpose, two scenarios with asymmetric cooling between the reactor cooling loops were investigated: The first one concerns a feedwater line break combined with a loss of offsite power (LOOP), while the second one is limited to the LOOP (or any other transient leading to the loss of the forced primary flow). The analyses were carried out using the best-estimate thermal-hydraulic system code CATHARE 2/V2.5_1mod8.1, developed by Commissariat à l’Energie Atomique, Electricité de France, AREVA, and Institut de Radioprotection et de Sûreté Nucléaire. The calculation results mainly emphasize the effect of the cooldown rate and the opening strategy of the main steam atmospheric discharge valve upon the occurrence of the NCI phenomenon.

C212 静的電気化学解析と動的酸化皮膜成長解析を結合したモデルによるPWR給水系配管の減肉評価(配管減肉評価法(1),OS5 保全・設備診断技術),動力エネルギーシステム部門20周年,次の20年への新展開)

In order to evaluate pipe wall thinning due to flow accelerated corrosion (FAC), FAC evaluation methods based on the coupled FAC model of static electrochemistry analysis and oxide layer growth analysis has been proposed. Applicability and accuracy of the FAC evaluation methods has been confirmed by comparing wall thinning rates calculated with the proposed methods with those measured for the secondary piping of a PWR plant. And then, it was confirmed that calculated wall thinning rates agreed with the measured ones within a factor of 2. The effects of water chemistry on corrosion rates were evaluated with the methods and the minimum oxygen concentration was calculated for suppressing corrosion rate. As a result of suitable amount of oxygen injection at the suitable location of the feed water line, whole region along the feed water line could be controlled out of corrosive conditions without any serious effects on SG corrosion.

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.

Development and application of a dual RELAP5-3D-based engineering simulator for ABWR

For any innovated plant design, the designed paper plant can be converted into a computer as a digital plant with advanced simulation techniques before being constructed into a real plant. A digital plant, namely engineering simulator, can be applied for: (1) verification of system design and system integration, (2) power test simulation, (3) plant transient and accident analyses, (4) plant abnormal and emergency procedure development and verification, (5) design change verification and analysis, etc. An advanced engineering simulator was successfully developed for the LungMen advanced boiling water reactor (ABWR) plant to support various applications before and after commercial operation. This plant specific engineering simulator was developed based on two separate RELAP5-3D modules synchronized on a commercial simulation platform, namely 3-Key Master. On this advanced LungMen plant simulation (ALPS) platform, major plant dynamics were simulated by two separate RELAP5-3D modules, one for reactor system modeling and the other for balance of plant (BOP) system modeling. Moreover, major control systems as well as emergency core cooling system (ECCS) were all simulated in great detail with built-in tasks of this commercial simulation platform. Different from real time calculation on training simulator, precision of engineering calculation is intentionally kept by synchronizing modules based on the most time-consuming one. During synchronization, each module will check its’ own converge criteria in each small time advancement. This plant specific advanced ABWR engineering simulator has been successfully applied on: (1) licensing blowdown analysis of feed water line break (FWLB) for containment design; (2) phenomena investigation of low-pressure ECC injection bypass during FWLB; (3) analysis of FW pump performance during power ascending; (4) verification of plant vendor's pre-test calculations of each start-up test.

Analysis of Low-Pressure ECC Injection Bypass of an ABWR During a Feedwater Line Break with RELAP5-3D/K

In the innovative design of the advanced boiling water reactor (ABWR), conventional recirculation loops are removed and replaced by multiple reactor internal pumps. Therefore, there is no major penetration of the reactor pressure vessel (RPV) below the elevation of the top of active fuel. As a result, an ABWR loss-of-coolant accident (LOCA) can have a decreased impact on reactor safety. Moreover, in the new RPV design the injection points of all the conventional low-pressure emergency core cooling (ECC) systems (ECCSs) are shifted out of the core shroud to the downcomer and feedwater line as a new low-pressure ECCS, namely, a low-pressure flooder (LPFL). Consequently, the net hydraulic head built inside the downcomer will be the only driving force to bring the low-pressure ECC water into the core shroud during a large-break LOCA. In the analysis of a feedwater line break with RELAP5-3D/K, it was occasionally found that the hydraulic head built in the downcomer might not be great enough to bring the ECC water into the core shroud, and when the mixture water column ascends above the elevation of the feedwater rings, all the water injected by the LPFL will be directly driven to the break on the feedwater line. Fortunately, the capacity of the remaining high-pressure ECC flow directly injected above the core is great enough, and this ECC low-pressure injection bypass phenomenon can be terminated once the high-pressure ECC injection is manually turned off. This phenomenon of low-pressure ECC injection bypass is unexpected in the ABWR design, and it is worth further investigation.

Application of an optimized AM procedure following a SBO in a VVER-1000

The University of Pisa was involved in investigations of an Accident Management (AM) procedure based on passive feed water injection. Some experiments were performed to validate this possibility (e.g. in LOBI and Bethsy facilities) and fully analyzed by thermal hydraulic system codes. Recent activities in which the University of Pisa is engaged (also as leader) are focused on VVER-1000 safety analyses. The idea is now to use the acquired knowledge to explore if a procedure based on passive feed water injection is applicable and can provide any benefits to the Russian design pressurised plant. The postulated accident is a station blackout, in such a way only passive systems are available. The proposed AM is based on secondary and primary side depressurisation in sequence. The secondary side depressurisation performed by the BRU-A valves has the scope to feed passively the SGs with the water left in the feed water lines and in the deaerators. The primary side depressurisation, via the PORV, is foreseen to keep the plant at the lowest pressure (to reduce the energy of the system) and to maximize the “grace time” of the plant. Three cases are here considered: no operator action, application of the optimized AM sequence, application of the AM procedure at the last time when it is effective. The intention of this paper is to show that in case of an unlikely event such a SBO the implementation of a strategy based on systems not designed for specific safety application can have a large impact on the “grace time” of the plant.

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors (V) Effects of Hydrazine and Dissolved Oxygen on Flow Accelerated Corrosion of Carbon Steel

Hydrazine and hydrogen co-injection into reactor water via the feed water line can mitigate stress corrosion cracking in BWRs. The effects of water quality variation due to hydrazine injection on flow accelerated corrosion (FAC) of carbon steel used in the BWR feed water line should be considered from the viewpoints of radioactive exposure, radioactive waste generation and plant integrity. Optimization of oxygen injection into feed water was considered to suppress FAC during hydrazine and hydrogen co-injection. In this report, corrosion tests of carbon steel were made in hydrazine and/or dissolved oxygen injected solutions (conditions: no pH adjustment by ammonia; temperature = 215°C; flow rate = 3.5 ms−1) to study the effects of dissolved oxygen and hydrazine on FAC. Hydrazine and oxygen co-injection suppressed the total corrosion rate by about 1/5 compared with the case of 0 μmol-kg−1 hydrazine and <0.06μmol.kg−1 oxygen injection. With hydrazine and oxygen co-injection, the inner layer, which is protective oxide, was close packed. Thickness of the inner layer also became larger compared with the case for only hydrazine injection or only oxygen injection. α-Fe2O3, which is thermodynamically more stable than Fe3O4, was formed in addition to Fe3O4 which was the single species formed for the case of only hydrazine injection. From these results, it was concluded FAC will be suppressed by keeping the oxygen concentration at the current level in BWRs (about 1 μmol-kg−1) if hydrazine and hydrogen co-injection is applied.

Study on the Control and Fabrication of Thermal Fatigue Crack for Reliability Enhancement of NDT in Reactor Coolant System

Damages in nuclear facilities during the operation of the Reactor Coolant System (RCS) are caused by cyclic loadings due to mechanical or thermal fatigue. Therefore, the development of an integrated technology including fabrication of standard specimens and their practical usage is needed to enhance the reliability of nondestructive testing for surge lines or main feed water lines. In this study, thermal fatigue cracks on STS 304 plates (t = 6mm) and tubes (O.D = 89.7mm. t = 7.7mm) for performance demonstration inspection were fabricated for the Reactor Coolant System (RCS) in nuclear power plants. In case of plates, cyclic thermal changes, from 20 to 450, were loaded on the V-notched specimens under tensile stressed conditions. The applied tensile stress was 1,700MPa at the notch portion. In the case of tubes, cyclic thermal changes, from 35 to 355, were applied on the V-notched specimens under compressive stressed conditions. In the case of plates, the initial crack was generated at 17,000 cycles and the depth of crack was 2.54mm at 22,000 cycles. Unlike the plates, in the case of tubes, the initial crack formed at 14,000 cycles and the crack penetrated the tube at 25,000 cycles. In this paper, shapes and fractographies of fabricated thermal fatigue cracks, and the used cyclic thermal loads are presented.

A Study on Ultrasonic Testing for Nondestructive Evaluation of Thermal Fatigue Crack in Pipelines

Pipelines of nuclear power plants undergo high pressure and temperature. Thermal stratification typically occurs in the surge line and the main feed water lines by flow and this stratification will initiate and propagate thermal fatigue cracks. This may cause rupture and leakage and it is a serious problem to nuclear power plants operation. Therefore it is very important to detect and measure thermal fatigue cracks. In this study, thermal fatigue cracks were generated in austenitic stainless steel specimens by a thermal cycle in notched pipes and weld jointed pipes. Ultrasonic techniques were used to evaluate the thermal fatigue crack depth. When ultrasonic waves propagate from an angle beam probe to thermal fatigue cracks, waves are reflected and diffracted. Crack depth was evaluated by the reflected signals from back wall and diffracted signals from the crack tip, but diffracted signals were too weak to detect so the reflected signals were more useful. The TOFD and dB drop methods were used in this study. The TOFD method is uses a time delay of diffracted signal from the crack tip. The dB drop method is an application of an amplitude decreasing rate by a probe moving distance.

Modeling of a Power Conversion System with RELAP5-3D/K and Associated Application to the Feedwater Blowdown Licensing Analysis for the ABWR Containment Design

The blowdown of feedwater (FW) line breaks (FWLBs) has been successfully analyzed by using the Appendix K version of RELAP5-3D. To adequately simulate a feedwater blowdown event, one must consider the main steam system, the turbine system, the moisture separator reheaters (MSRs), the main condenser, and the condensate and FW system as all are involved in the modeling scope. The essential components of the simulation scope include the steam header, the high- and low-pressure turbines, the MSR, the FW pump (FWP) turbines, the main condenser, the condensate and booster pumps, the FW heaters of six stages, the steam extraction of seven stages, and the turbine-driven FWPs. All of the components are connected by associated piping as designed.Regarding the FW blowdown analysis, blowdown mass and energy are the two most important parameters to be calculated. Several essential phenomena are involved in this FW blowdown event, which include critical flow at the break and the internal venturi, flashing of FW near the break, runout and coastdown of the FWPs, steam extraction to FW heaters and FWP turbines, flashing of saturated water initially stored inside the FW heater shell sides and MSR drain tanks, energy release from saturated water and system metal, and cold water transportation from the main condenser to the break. All the essential processes involved during FWLB can be well simulated by the advanced Appendix K version of RELAP5-3D. The blowdown analysis calculated by RELAP5-3D/K for the FWLB was contracted to provide a solid basis for the final safety analysis report containment design analysis for the Lungmen advanced boiling water reactor (ABWR) plant. The successful application of RELAP5 for the entire balance-of-plant simulation and associated FW blowdown analysis indicates that the advanced RELAP5 can extend its traditional reactor safety analysis to the entire power conversion system simulation and analysis.

Evaluation of Nuclear Piping Failure Frequency in Korean Pressurized Water Reactors

The purpose of this paper is to evaluate the piping failure frequency based on the piping failure events in Korean pressurized water reactors (PWRs) until the end of 2003. Two types of the piping failure frequencies including the piping damage frequency and the piping rupture frequency are considered in this study. The piping damage frequency for the failed piping system was estimated by using the piping population data such as the weld count or the base metal count. The piping rupture frequency related to the initiating event in a probabilistic safety assessment (PSA) was evaluated by using both the Bayesian approach (Method 1) and the conditional rupture probability approach (Method 2). In the Bayesian approach, two methods using Jeffreys noninformative prior (Method 1-1) and prior distributions based on the results in NUREG/CR-5750 (Method 1-2) were considered. Thirty piping failure events in ASME safety class pipings of Korean PWRs were identified and analyzed in this study. The results showed that the piping damage frequency for the events ranged from 5.42E-3/cr.yr to 2.77E-5/cr.yr. Three kinds of initiating events including the very small LOCA, the feedwater line break, and the flood are evaluated for Korean PWRs. The results for the piping rupture frequency in Korean PWRs were as follows: 1) The mean piping rupture frequency of the very small LOCA event ranged from 3.6E-3/cr.yr to 1.2E-2/cr.yr, the feedwater line break event from 3.6E-3/cr.yr to 2.5E-2/cr.yr, and the flood event from 7.8E-4/cr.yr to 3.6E-3/cr.yr. The mean piping rupture frequencies of the very small LOCA and feedwater line break events were higher than that of the flood event by one order of a magnitude. 2) Method 2 gave conservative results in the very small LOCA and feedwater line break events compared to Method 1-1 or Method 1-2, while Method 1-1 gave conservative results in the flood event. 3) The order of magnitudes in the mean piping rupture frequencies of the very small LOCA, the feedwater line break, and the flood in Korean PWRs were similar to those in the U.S. PWRs.

MICROCLIMATE AND ENERGY CONSUMPTION IN COMMERCIAL, HOT-WATER AND STEAM HEATED GREENHOUSES FOR TOMATO PRODUCTION

MICROCLIMATE AND ENERGY CONSUMPTION IN COMMERCIAL, HOT-WATER AND STEAM HEATED GREENHOUSES FOR TOMATO PRODUCTION

Subcompartment Pressurization Analysis Following Feedwater Line Break Accident using Realistic Evaluation Model

Currently, a high energy line break in the Main Feedwater Piping Area of Kori Unit 1 Nuclear Power Plant would be not vented through the room doors and ventilation dampers. Therefore, modifications to the wall design should be done to demonstrate that the Main Feedwater Piping Area walls are structurally sound during the feedwater line break event. The modifications to the wall design are minimized through the Realistic Evaluation Methodology. The purpose of this paper is to present the pressurization analysis of the Main Feedwater Piping Area in the event of feed water line breaks using the Realistic Evaluation Methodology, which consists of simulation of the response of the Nuclear Steam Supply System using RETRAN-3D code and generation of the realistic conservative Mass and Energy release rates through the break using the enveloping pressure curve to bound the RETRAN-3D calculated break pressure, and calculation of the optimized differential pressure of the Main Feedwater Piping Area using COMPARE/MOD1A code. Based on the Realistic Evaluation Methodology, a blowout panel with vent flow area of at least 14 m2 will be designed and installed to help control the room pressurization and to meet the wall structural design criteria of 0.02Mpa.

Subcompartment Pressurization Analysis Following Feedwater Line Break Accident using Realistic Evaluation Model

Subcompartment Pressurization Analysis Following Feedwater Line Break Accident using Realistic Evaluation Model

Weapons-Grade Plutonium-Thorium PWR Assembly Design and Core Safety Analysis

A light water reactor (LWR) fuel assembly design consisting of a blend of weapons-grade plutonium and natural thorium oxides was examined. The design meets current thermal-hydraulic and safety criteria. Such an assembly would have enough reactivity to achieve three cycles of operation. The pin power distribution indicates a fairly level distribution across the assembly, avoiding hot spots near guide tubes, corners, and other sections where excessive power would create significant loss to thermal-hydraulic margins.This work examined a number of physics and core safety analysis parameters that impact the operation and safety of power reactors. Such parameters as moderator coefficients of reactivity, Doppler coefficients, soluble boron worth, control rod worth, prompt neutron lifetime, and delayed-neutron fractions were considered. These in turn were used to examine reactor behavior during a number of operational conditions, transients, and accidents. Such conditions as shutdown from power with one rod stuck out, steam-line break accident, feedwater line break, loss of coolant flow, locked rotor accidents, control rod ejection accidents, and anticipated transients without scram (ATWSs) were examined.The analysis of selected reactor transients demonstrated that it is feasible to license and safely operate a reactor fueled with plutonium-thorium blended fuel. In most cases analyzed, the thorium mixture had less-severe consequences than those for a core comprising low-enriched uranium fuel. In the analyzed cases where the consequences were more severe, they were still within acceptable limits. The ATWS accident condition requires more analysis.