Development of the ISI Device for Fast Breeder Reactor MONJU Reactor Vessel

In-service inspection (ISI) is carried out to confirm the integrity of the main components of the Fast Breeder Reactor (FBR) “MONJU”. The weld-joints are examined by using an inspection device which has a glass fiber scope for visual examination and a horizontally polarized shear (SH) wave electromagnetic acoustic transducer (EMAT) for volumetric testing. The ambient temperature during the inspection is 200°C and the irradiation field is 10 Sv/hr. A new inspection device has been developed in order to improve the visual test performance, volumetric test performance and controllability of the inspection device reflecting the experience of the original test. In this paper, detail of the new inspection device and the test results of sensors such as the CCD camera, EMAT and bead sensor are reported. The paper also reports on the CCD camera cooling system and other components.

Installation of Valvtechnologies Ball Valves in Severe Service Applications Within the Advanced Test Reactor

The Advanced Test Reactor (ATR) located at the Department of Energy’s Idaho National Laboratory, is the most powerful test reactor operating in the United States rated at a design power of 250 MW(t). Operating cycles are nominally seven per year with outages that last 7 to 14 days, allowing time for routine plant maintenance and experiment insertions and manipulations. While the ATR pressurized water loops can operate at the same temperature and pressure requirements of a pressurized water reactor, the loops also have the ability to operate at higher conditions. Hence, it is critical to ensure that when component replacements are called for, they can meet or exceed design requirements of a typical power reactor, while continuing to satisfy the design requirements of the ATR experiment loops.

A Review of the PSA Model Truncation Convergence Issues at Some ENS Plants

This study provides the insights gained from the Probabilistic Safety Assessment (PSA) model update of several Entergy Nuclear South (ENS) plants with respect to truncation convergence based on the limited guidance on the issue in the industry. The industry rule of thumb, the ASME and NRC guidance and requirements on the subject have been reviewed. The recent model updates performed at some of the ENS plants (River Bend, ANO 1 and 2) considered these criteria. Based on the current criteria used in the industry for truncation convergence, the recent PSA model update results for the River Bend Station (RBS) and ANO-1 are not converging even at a low truncation limit of 1E−11/reactor-year (yr). Many improvements were introduced in the recent model updates and convergence was expected at higher truncation values. This paper discusses the issues identified that are related to the convergence of the PSA results at low truncation limits.

Corrosion Phenomena of Eurofeer Steel in Pb-17Li Stationary Flow at Magnetic Field

Search of new energy sources draws the increasing attention to use for this purpose of reactors. In the Europe some years the program EUROATOM uniting scientific of the many countries for the decision of constructive problems at designing of fusion reactors operates. One of the main things in this program is the problem of liquid metals breeder blanket behaviour. Structural material of blanket should meet high requirements because of extreme operating conditions. Therefore the knowledge of the effect of metals flow velocity, temperatures and also a neutron irradiation and a magnetic field on the corrosion processes are necessary. At the moment the eutectic lead -lithium (Pb-17Li) is considered as the most suitable tritium breeder material [1–3]. In turn as a structural material have been tested both many austenitic and ferritic-martensitic steels [2–4]. As the optimum variant is considered steel EUROFER 97, which corrosion rate in liquid Pb-17Li eutectic is the least [3,4]. However, these results have been received without taking into account influence of a strong magnetic field. At the same time, this influence should be essential, as because of change of hydrodynamics of a liquid metal flow, and because of interaction of a magnetic field with a ferromagnetic steel. It has been shown in [5,6] that the magnetic field leads to increase of corrosion rate for austenitic (316L) and martensitic (1,4914) steels. Experimental data for EUROFER 97, and also a theoretical substantiation of the phenomenon are absent, that creates essential difficulties for forecasting working capacity of blanket construction. The aim of presented work were the theoretical and experimental investigations of magnetic field influence on the corrosion of EUROFER 97 steel exposed to flowing Pb-17 Li in specific designed loop.

SCC Propagation Rate of Type 304, 304L Steels Under Oceanic Air Environment

Corrosion integrity of canister in the concrete cask for spent fuel storage is very important because the canister serves to maintain the sealability over the storage period of 40 to 60 years. Natural exposure and accelerated corrosion tests of conventional stainless steels for canister, that are Type 304, 304L, and 316(LN), for concrete cask’s canister have been conducted by using many three Point Bending (3PB) test specimens and compared. The SCC propagation rates in Type 304 and 304L at the natural condition were about 1.2E−12 to 1.8E−11 m/s at the K (Stress Intensity Factor) range of 0.6 to 9.0 MPa√m, and that of the accelerate test (60 degrees C, 95%RHS., filled with NaCl mist) were about 1.0E−10 to 3.5E−9 m/s at the K range of 0.3 to 32 MPa√m. The SCC propagation rates under both natural and accelerated conditions were independent with K. Both da/dt values of the direct exposure test and of the under glass exposure test were in the same scattering band.

Vitrification Technology Development Plan in Tokai Reprocessing Plant

The Tokai Vitrification Facility (TVF) is the only operating vitrification plant in Japan, constructed and operated by JAEA, to vitrify concentrated high radioactive liquid waste (HALW) in the Tokai Reprocessing Plant (TRP). JAEA started TVF hot operation in 1995 and produced 218 canisters as of March, 2006. An existing melter is the second melter, which was installed from 2002 to 2004 in place of the first melter stopped its operation by damage of a main electrode. JAEA has estimated that the damage was caused by accumulation of noble metal. Therefore, melter bottom structure was improved to get better drain ability of glass containing noble metal. Completing the melter replacement, vitrification operation was restarted in October 2004 and produced 88 canisters successfully until the end of March 2006. Through these experiences, JAEA made basic strategy to achieve stable TVF operation: keeping stable operation of the existing melter preventing adverse effect by noble metal accumulation and developing a new advanced melter with long lifetime preparing for future exchange as the third melter. Based on the basic strategy, JAEA made a decade development plan of necessary key technologies and has started the development since 2005.

WWER Steam Generators Tubing Performance and Aging Management

At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned.

Using Center Hole Heat Transfer to Reduce Formation Times for Ceramic Waste Forms From Pyroprocessing

The waste produced from processing spent fuel from the EBR II reactor must be processed into a waste form suitable for long term storage in Yucca Mountain. The method chosen produces zeolite granules mixed with glass frit, which must then be converted into a solid. This is accomplished by loading it into a can and heating to 900 C in a furnace regulated at 915 C. During heatup to 900 C, the zeolite and glass frit react and consolidate to produce a sodalite monolith. The resultant ceramic waste form (CWF) is then cooled. The waste form is 52 cm in diameter and initially 300 cm long but consolidates to 150 cm in length during the heating process. After cooling it is then inserted in a 5-DHLW/DOE SNF Long Canister. Without intervention, the waste takes 82 hours to heat up to 900 C in a furnace designed to geometrically fit the cylindrical waste form. This paper investigates the reduction in heating times possible with four different methods of additional heating through a center hole. The hole size is kept small to maximize the amount of CWF that is processed in a single run. A hole radius of 1.82 cm was selected which removes only 1% of the CWF. A reference computation was done with a specified inner hole surface temperature of 915 C to provide a benchmark for the amount of improvement which can be made. It showed that the heatup time can potentially be reduced to 43 hours with center hole heating. The first method, simply pouring high temperature liquid aluminum into the hole, did not produce any noticeable effect on reducing heat up times. The second method, flowing liquid aluminum through the hole, works well as long as the velocity is high enough (2.5 cm/sec) to prevent solidification of the aluminum during the initial front movement of the aluminum into the center hole. The velocity can be reduced to 1 cm/sec after the initial front has traversed the ceramic. This procedure reduces the formation time to near that of the reference case. The third method, flowing a gas through the center hole, also works well as long as the product of heat capacity and velocity of the gas is equivalent to that of the flowing aluminum, and the velocity is high enough to produce an intermediate size heat transfer coefficient. The fourth method, using an electric heater, works well and heater sizes between 500 to 1000 Watts are adequate. These later three methods all can reduce the heatup time to 44 hours allowing production to be doubled and a more uniform heating.

Low Burnup Inert Matrix Fuels Performance: TRANSURANUS Analysis of the Halden IFA-652 First Irradiation Cycle

The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd·kgUeq−1 vs. 45 MWd·kgUeq−1 (40 MWd·kgUOXeq−1) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd·kgUeq−1) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed.

Development of System Based Code: Methodologies for Life-Cycle Margin Evaluation

The late Professor Emeritus Yasuhide Asada proposed the System Based Code concept, which intends the optimization of design of nuclear plants through margin exchange among a variety of technical options which are not allowed by current codes and standards. The key technology of the System Based Code is margin exchange evaluation methodology. This paper describes recent progress with regards to margin exchange methodologies in Japan.