[en] The present invention provides control rods for a BWR type reactor, in which stress exerted to a sheath is prevented by disposing a long stiffener incorporated with neutron absorbers in a wing. A long stiffener is secured integrally with the sheath over the longitudinal direction of the wing at the lateral center of the wing. In addition, spacers are secured to the long stiffener each at a predetermined distance in the longitudinal direction of the neutron absorbers, and water channels are formed between both of them. Since the long stiffener is disposed over the entire length of the wing at the lateral center of the wing, the gaps in the elongated U-shaped sheath are kept strongly to ensure the strength, and since the area of the welded portion between the sheath and the stiffener is increased, the stress exerted to the sheath caused by the difference of the heat coefficient expansion of the sheath and that of the neutron absorbers can be dispersed. (N.H.)

[en] A portion of irradiated fuel assemblies at a place where a reactivity effect is high, that is, at a place where neutron importance is high is replaced with standard fuel assemblies having a known composition to measure neutron fluxes at each of the places. An effective composition at the periphery of the standard fuel assemblies is determined by utilizing a calibration curve determined separately based on the composition and neutron flux values of the standard assemblies. By using the calibration curve determined separately based on this composition and the known composition of the standard fuel assemblies, an effective neutron multiplication factor for the fuel containing portion containing the irradiated fuel assemblies is recognized. Then, subcriticality is ensured and critical safety upon containing the fuel assemblies can be secured quantitatively. (N.H.)

[en] The device of the present invention rapidly detects leakage of primary coolants due to rupture of heat transfer pipes in a steam generator of a PWR type reactor. That is, a neutron detector is disposed, as a dose rate measuring system, to a secondary main steam system, a secondary main steam pipeline, or a turbine. A calculation processing system compares the data obtained therefrom and a normal state, to judge the presence of an abnormal symptom due to leakage and calculate radioactivity concentration in the main steams at a measuring point based on the dose rate. With such procedures, if a heat transfer pipe in the steam generator should be ruptured, radioactive materials in the primary coolants reach the position of the neutron detector in several seconds. Based on the result, progress of the leakage is forecast, to estimate the scale of the ruptured portion with lapse of time. Since neutrons are an object of the measurement in the device of the present invention, the device does not undergo influence of gamma rays released from a radiation source nuclide present in natural radiation rays and the primary system. (I.S.)

[en] The present invention provides a method of and a device for measuring gamma rays of fuel assembly, capable of suppressing lowering of accuracy in the measurement even if the rotational direction of the fuel assembly is not determined accurately. Namely, at least a pair of gamma ray detectors are disposed at positions on assumed planes spaced each at an equal distance from a plane including an axial center of the fuel assembly, and assumed so that the fuel assembly is made symmetric with respect to the right-to-left direction. Gamma rays of the fuel assembly are measured by signals formed by synthesizing the signals of the pair of detectors. With such a constitution, when the fuel assembly is displaced by a predetermined angle, even if effective sensitivity of one detector is increased, effective sensitivity of the other detector is lowered. Accordingly, the average value of both detectors (including weighted average value, not limited to a simple average value) is substantially constant. As a result, the distance from the assumed plane is made identical, but the pair of detectors can be allowed by ±5mm in the axial displacement of the height. (I.S.)

[en] In a method of evaluating a service life of control rods in a BWR type reactor, a not-irradiated control rod is used as a standard control rod, and a calibration curve is determined based on a measured value for a neutron transmission rate, a burning dependent transmission rate and burning dependent reactivity of the not-irradiated control rods. Reactivity of an irradiated control rod is evaluated based on the measured value for the neutron transmission rate and the calibration curve, to determine the extent of the depletion of the evaluation value of the reactivity from the reactivity upon no-irradiation, thereby evaluating the degree of the nuclear service life of the irradiated control rods approaching to a limited predetermined value. Even for a control rod using neutron absorbers of complicate nuclear property different from a B₄C control rod comprising only ¹⁰B, the nuclear service life can be evaluated by measurement and analysis. (N.H.)

[en] Short communication

[en] EPR (EPR: European Pressurized Water Reactor) has been developed to achieve higher output (1,600 MW) and longer plant life (60 years), compared with the conventional unclear reactors, and the first commercial reactor was introduced in Finland no.5/Olkiluoto no.3. The integrated mono-block design was applied for nozzle shell flange instead of welded conventional flange and nozzle shell. And due to set-on type nozzles, a 600 ton ingot was required for this part. JSW successfully completed the manufacturing of first nozzle shell with integral flange 11 months after melting. This report summarizes manufacturing technology and process, and properties of nozzle shell with integral flange. (author)

[en] The main parts of a metal cask for storage and transport of spent nuclear fuel consists of main body, neutron shield material and external cylinder. The forged main body has been manufactured as a cup shape by welding of 'forged body' and 'forged bottom plate' which are independently forged. JSW has developed the manufacturing technology of 'cask body integrated with bottom plate' which has no weld line with the goal of cost reduction, manufacturing period shortening and further reliability improvement. Manufacturing for the prototype of 'cask body integrated with bottom plate' has completed to verify mechanical properties and uniformity of the product which satisfy the specified values stipulated in JSME Code S FA1 2007 edition. Here, we report the manufacturing technology and obtained properties of 'cask body integrated with bottom plate'. (author)

[en] On the steel forging (SF) elements for steam generator (SG) of the pressurized water type light water reactor (PWR), from a viewpoint of upgrading in their improvements of design and materials, here were described on three materials such as integrated steel forgings, high strength steel forgings, and vacuum carbon deoxidisation (VCD) steel forgings. On production of SG, by using the integrated SF, not only structural soundness of SG is upgraded, but also inspections containing inspections under production and usage become easier, to bring minimization of maintenance inspection and reduction of exposure under operation. And, in order to reduce weight of SG and upgrade seismic resistance, SA508, a Cl.3a high strength SF (620 MPa class in tensile strength) is used for some nuclear plants. Here were introduced material properties of this SF and described its chemical components and heat treatment condition. And, as a method to reduce macro- and micro-segregation of materials and to upgrade homogeneity of material property, a method combined deoxidisation of steel due to carbon monoxide reaction with crystal grain minimization due to addition of aluminum was investigated. In addition, properties of a low Si-SA508 Cl.3 steel using this method was compared with that of usual SA508 Cl.3 steel. (G.K.)

[en] In recent years, the components of PWR such as a steam generator, closure head of pressure vessels and pressurizer have been replaced for the lifetime lengthening of nuclear power stations. For the closure head, we developed a head forging type integrated with flange superior to the conventional joint welding one fabricated by steel plate head and forged flange. This report described our manufacturing technology, material properties and manufacturing experience for the newly developed closure head. (author)