[en] This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

[en] This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

[en] Injection of a water stream into a circular pipe flow of a different orientation appears in many important engineering applications, such as in nuclear reactor safety analysis, the mixing of two fluids at a T-shaped junction, etc. In this regard, the analysis of pressurized thermal shock (PTS) in the nuclear industry is a special issue. One of the major efforts concerning this issue is to predict the mixing behavior and stratification of a round cold jet from the safety high-pressure injection (HPIS) line with a hot loop flow circulating in the cold-leg pipe. Typically, in Westinghouse pressurized water reactor (PWR) systems, high-pressure water from 60 degrees injector is charged into the horizontal cold leg. The paper will discuss the 3D finite volume simulation, using CFD (computational fluid dynamic) code FLUENT, of the EPRI/Creare experimental results (one-fifth-scale model of PWR HPIS and cold-leg mixing), test numbers 42 and 46. The reason for this work is to validate FLUENT capability on benchmarked test results, as a preparation for its use in the more complex 3D finite volume analysis of full scale water mixing in the reactor vessel downcomer. It is part of the study mentioned to observe the most severe design basis accident sequences conditions for NPP Krsko and re-evaluate their PTS potential. The most important factors needed for PTS analysis are the following: the final temperature in downcomer, the temperature decrease rate, non uniform cooling of the reactor pressure vessel wall (characterized by cold plume of SI water) and the RCS pressure. (author)

[en] An event of water coolant ingress (ICE: Ingress of Coolant Event) into a tokamak vacuum vessel in a nuclear fusion experimental reactor is caused by coolant pipe rupture, for example, at the divertor plate. The event progresses from the pipe rupture, subsequent water injection into the vacuum, splashing on the hot plasma facing plates, water evaporation, and to chemical reactions with the material of the plasma facing components such as graphite. The event might lead to severe consequences of pressurization enough to break the pressure boundary and possible releases of radioactive tritium retained in the graphite protective tiles and/or activated materials to the outside of the vacuum vessel. Therefore, it is one of momentous issues in the safety design to evaluate the consequences of ICE. In the initial stage of ICE water injection and water vaporization occur in the vacuum vessel and the water column freezes under some condition. Experiments of water jet into vacuum vessel were carried out to examine thermohydraulic characteristics in the initial stage. The authors assumed a small break of the coolant pipe in the present experiments. Water was ejected from a circular pipe into a vacuum enclosure. The circular pipes with various inner diameters of 0.5, 1.0, 2.0 and 5.0 mm were used to simulate the rupture part of the coolant pipe. Injection pressures of water ranged from 0.1 to 0.5 MPa. The injection process of water into vacuum was observed with a video camera. It was found that water evaporated first, then water column froze, melted and injected repeatedly under the conditions of low injection and small inner diameter until the pressure inside of the vacuum enclosure exceeded the saturated pressure. For the high injection pressure of water, the water did not freeze and the pressure inside the vacuum vessel increased rapidly up to the saturated pressure. These results will be reduced into models appropriate for analysis

[en] We are developing a statistical safety evaluation method using the RELAP5/MOD3.2 code for the loss-of-RHR (Residual Heat Removal) event during the mid-loop operation. To confirm the code prediction performance for gravity injection which is a one of the mitigation measures for this event, the Bethsy 6.9a test was analyzed using RELAP5/MOD3.2. In the analytical results, water mass flow rate into the pressurizer in the early period of the transient event was overpredicted. But water mass flow rate into the pressurizer was able to be decreased by artificially giving the circulation flow between core and the core bypass region. (author)

[en] A Reactor Core Isolation Cooling (RCIC) steam line break in the RCIC pump room or in the torus will effect the temperature and pressure of the High Pressure Coolant Injection (HPCI) pump room. The extent of the pressurization and heat up of the HPCI room is dependent upon the location of the break as well as the size of the penetrations between the HPCI room and the room where the break occurs. The concern is that a RCIC steam line break could lead to the HPCI room heating up to temperatures above the environmental qualification (EQ) limits of the equipment. The EQ limit is 185 F. If this temperature is reached in the HPCI room, a RCIC steam line break could lead to both the RCIC and HPCI systems being inoperable. This is possible as the torus compartment and Emergency core cooling systems (ECCS) rooms are all located at the same plant elevation level. There are penetrations between these rooms that would act as steam flow paths in the event of a line break or leak. Two different break locations were analyzed using the GOTHIC Containment Analysis Package. The first break was modeled in the RCIC room itself. This should be the limiting break. The volume of the RCIC room is approximately one tenth that of the torus, which is the location of the second break modeled. This difference in volume will result in different heat ups as well as different pressures of the room with the break and connected rooms

[en] The separation between the low-pressure moderator and high-pressure coolant in CANDU reactors makes it possible to utilize the moderator as a heat sink for emergency heat removal. This paper presents a review of a passive moderator-cooling concept that can be used under both normal and upset conditions. The paper also presents recent work completed realization of this concept. (author)

[en] Short communication

[en] In quantifying a plant-specific Poisson event occurrence rate λ in Probabilistic Risk Assessment, it is sometimes the case that either the corresponding Poisson exposure time t or the observed number of events x (or both) are uncertain. We present several methods which account for uncertainties in both x and t when using Bayesian methods to estimate λ. A gamma prior distribution on λ is considered. While the methods formally require numerical integration, a computationally convenient approximation is provided to implement them in practice. A numerical example concerning the rate of failure to operate of the high pressure coolant injection system of commercial boiling water reactors is used to illustrate the methods

[en] In PNGS 'B', the D₂O Storage Tank is pressurized with a Cover Gas (Helium) to ensure that a pressure higher than 45 kPa(g) exists at the suction of the Pressurizing pumps in any operating condition. This requirement is easily met during power operation with the combination of a minimum Cover Gas pressure of 30 kPa(g) and the hydrostatic head of the coolant, but it has been difficult to meet during the cooldown of the units. According to the operating procedures, prior to a unit cooldown the Cover Gas pressure is increased to 200 kPa(g) with a recommended D₂O Storage Tank level of 2.1 m (minimum of 1.9 in). During the cooldown, approximately 33 Mg of coolant are transferred from the D₂O Storage Tank to the Heat Transport System (HTS) to compensate for the coolant's density change. This amount of coolant is transferred by the Pressurizing pumps and is required to ensure that HTS pressure remains under control during the cooldown. However, the existing Cover Gas addition system is unable to supply enough Helium to maintain a pressure of at least 30 kPa(g) during the coolant pump out, and consequently, cooldowns are often interrupted or slowed down. Unfortunately, the cooldowns are interrupted when the HTS pressure and temperature conditions are adverse to the Pressure Tubes, causing the unnecessary use of thermalcycles. Thus, a numerical model of the Cover Gas in the D₂O Storage Tank was developed to understand the phenomenon and propose a modification of the cooldown evolution to avoid the use of thermalcycles. (author)