General Overview
A. Promoted project subjects/Project subjects
The objectives throughout the research period and the principal results in FY 2006 are described here for the promoted
project subjects and project subjects.

1. Nuclear Technology: Supporting Foundations for a Stable Supply(1) High accuracy prediction of irradiation embrittlement in light water reactors and its code formation (promoted project subject): FY
2006 - FY 2008
[Objectives]
To develop a highly accurate prediction method for the irradiation embrittlement of a pressure vessel and a fracture toughness master curve method to ensure rational and highly reliable operation in preparation for the aging of light water reactors and topropose draft codes for academic societies and industrial associations.
[Principal Results]
・Using the knowledge acquired through the observation of microstructure monitoring specimens of pressure vessels at domestic
nuclear power plants, the embrittlement prediction method originally developed by the CRIEPI was further improved so that thedraft revision of JEAC4201, one of the Japan Electric Association Technical Standards, can reflect the improved method (Fig.1).
・A master curve fracture toughness test using small specimens was conducted for the steel used for domestic nuclear reactor pressurevessels and the effects of the size and shape of the specimens were clarified.
・A draft standard of the Japan Society of Mechanical Engineers (JSME) was prepared for evaluation of the soundness of low classpiping.
160140120

Predicted Value (C)100806040

Present Prediction Method
CRIEPI s Prediction Method200-20020

40

60

80

100

120

140

160

180-20

Actual Value (C)Fig.1 Prediction Accuracy of the CRIEPI s Embrittlement Prediction Method(2) Comprehensive measures to deal with the deterioration of light water reactors caused by thermal flow: FY 2006 - FY 2008
[Objectives]
To formulate management rules and to develop supporting technologies from the viewpoints of thermal flow and vibration
inside a reactor and piping for the purpose of the early realization of the reduced frequency of plant stoppage and an increased nuclear
reactor output.
[Principal Results]
・A test was conducted on the diameter of the droplets in high speed steam as the droplet size is important for the prediction ofdamage by droplet impact erosion. It was found that approximation is possible by the general form of exponential distribution toenable prediction of the maximum droplet diameter.
・Using the BWR s regional stability test facility (SIRIUS-F) which is capable of simulating void reaction feedback in detail andwhich was developed by the CRIEPI, a simulation test was conducted with ABWR and BWR-5. The results confirm that the impactof MOX fuel installation on the regional stability of the reactor core is small and that the prediction accuracy of ODYSY, which is
the licensing analysis code, is sufficiently high.
(3) Advanced assessment technology for SCC crack propagation of LWR materials: FY 2006 - FY 2009
[Objectives]
To contribute to the rational operation of LWR through the wider application of the maintenance standards, in turn based on
clarification of the data adoption criteria as the basis for the analysis of the crack propagation characteristics and the maintenancestandards, and upgrading of the crack propagation speed evaluation diagramme, etc.

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[Principal Results]
・It was found that the SCC crack length and depth distribution of low carbon stainless steel follow the logarithmic normal distribution and that the maximum value is determined by extreme statistics. These findings suggest that the time for SCC cracks to appear
can be predicted.
・“SCC Advanced Testing Facility for Environment of high Temperature water in Yokosuka (SAFETY)”, a large-scale facility forbase research, has finally been installed to establish the testing method to check the long-term soundness of the technology to deal
with the SCC of high aged light water reactors.
(4) Assessment of the biological effect of low dose radiation: FY 2006 - FY 2008
[Objectives]
To establish a more rational protection system for radiation by means of acquiring and presenting scientific data based on anepidemiological survey and animal experiments and properly assessing the risk of low dose and low dose rate radiation.
[Principle Results]
・Analysis of the findings of an epidemiological survey in high natural radiation areas in China confirmed that the cancer death rate
will not rise even if the radiation level is three to five times higher than that of Japan. The analysis results were presented to theICRP and IAEA, etc. to contribute to the formulation of protection standards for radiation.
・Data relating to the cumulative nature of the exposed dose was acquired and analysis of the carcinogenic and carcinoma inhibition
mechanisms confirmed that the metabolism of cell and other play an important role.
(5) Development of rationalized radiation safety ensuring method: FY 2006 - FY 2008
[Objectives]
To propose a method to judge the suitability of rationalized and advanced safety assessment and measuring technologies forsafety assessment standards so that radiation safety measures can be better explained to and understood by the public.
[Principal Results]
・The measurement performance test of the CRIEPI s clearance level measuring equipment (CLALIS) for metal and concrete waste
was completed and a plan for a field demonstration test leading to the commercial use of CLALIS was formulated.
・Calculation of the lasting period of an artificial barrier s performance and the shifting time to groundwater was conducted from theviewpoint of the risk theory with radioactive carbon which is important in terms of the extra depth disposal of radioactive waste andthe effect of such disposal method on dose assessment was clarified.
・The actual source efficiency which is important for the surface contamination assessment for radiation control was measured at theHot Laboratory and the compatibility with the surface contamination density criterion under the existing law was verified.
(6) Development of high level radioactive wastes disposal technology (promoted project subject): FY 2006 - FY 2008
[Objectives]
To develop a geological and groundwater environment survey and assessment technique, design and construction technologies for a rational disposal facility and a method to assess the performance and safety of barriers so that high level radioactive waste
can be safely disposed of.
[Principal Results]
・By collaboration research on sedimentary rock with JAEA s Honorobe in Japan and Mont Terri project in Switzerland, the applicability of the controlled drilling technology was verified and a method, which takes rock anisotropy due to cracks and other factors
into consideration, to measure initial stress was developed (Fig.2).
・The creep test for soft sedimentary rock was conducted under simulated in-situ conditions to clarify the effect of environmental
factors on the long-term creep characteristics.
・The heat resistance test was conducted for bentonite under simulated disposal site conditions to establish the degree of the impact ofheat.
・The design requirements for a disposal facility were identified for the use of various types of low alkaline cement as barrier materials. In addition, a cement crack control technique was proposed to the Nuclear Waste Management Organization of Japan.
Development of Controlled Drilling Technology

Drilling machineSea
Freshwater environment
Salt water environment

Fault/crack zone

Controlling drilling and logging device

The applicability of the controlled drilling technology was verified at the Honorobe siteby drilling a 700 m long well.

Fig.2 Development of Controlled Drilling Technology and Survey Technique (Entrusted by theAgency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry)3

(7) Development of low level radioactive waste disposal technology (promoted project subject): FY 2006 - FY 2008
[Objectives]
To develop a method to assess the radionuclide migration inhibition function and long-term durability, etc. of artificial barrier materials (cement and bentonite, etc.) to contribute to the extra deep disposal of low level radioactive waste and to propose treatment and recycling measures for demolition waste.
[Principal Results]
・Data was acquired on the hydraulic conductivity and swelling characteristic of bentonite which was altered by various ions eluted
from cementitious materials and waste materials. The basic alteration mechanism of cementitious materials was clarified.
・Highly dense cementitious materials were produced to further improved the already developed low alkaline cement and the material
and dynamic properties of thee materials were evaluated in an experimental manner. In addition, a new flow-type test method wasdeveloped to clarify the phenomenon of eluviation relating to the evaluation of the long-term durability of cementitious materialsunder the conditions of a disposal site environment.
・A business model and local symbiosis plan using a simulator were compiled based on a survey on the local industrial structure andphysical distribution, etc. so that a decommissioning and recycling similar for demolition concrete can be used for the disposal andrecycling of industrial waste.
(8) Development of storage and transport technologies for recyclable fuel, etc. (promoted project subject): FY 2006 - FY 2008
[Objectives]
To develop, propose and promote the practical application of an economical and highly reliable long-term large capacity
storage technology to assist the stable operation and a flexible pre-processing plan for nuclear power plants.
[Principal Results]
・The feasibility of a reasonable design for shielding for concrete cask storage facilities was verified using the radiation streaming
safety assessment method based on the probability theory. In addition, a new design concept for a shallow underground vault
storage facility was developed based on the assessment results of the heat removal and aseismic capacities.
・The rust and SCC generation test and the salt deposit test were conducted with canister materials for the concrete cask storage
system and the low possibility of SCC generation during a storage period of some 50 years was confirmed.
・Based on the results of the concrete cask earthquake resistance test conduced in FY 2005, assessment of the soundness of the spent
nuclear fuel in the storage vessels at an intermediate storage facility supported by pile foundations at the time of an earthquake wasconducted (Fig.3).
・In connection with a possible falling accident at the time of the delivery of metal casks to an intermediate storage facility, using thecommercial plants scale model, the flat fall and spinning fall tests were conducted from a height of 1 m and no leakage of the filled
gas inside the casks was confirmed.

A concrete cask not firmly lashedwas tested with three types ofseismic wave (El Cento, JMA
Kobe and artificial seismic wave.
The concrete cask did not fall
and it was clarified that theresponse of the stored fuel would
be contained within the elasticity
range.
Aseismic test of a real size concrete caskon the large-scale triaxial vibration table
at the Research Institute for Disaster
Prevention Technology in Germany (EDefence)

Fig.3 Aseismatic Performance Demonstration Test for Concrete Casks (Entrusted by the Nuclearand Industrial Safety Agency of the Ministry of Economy, Trade and Industry)(9) Development of metal fuel cycle technology (promoted project subject): FY 2006 - FY 2008
[Objectives]
To proceed with the development of metallic fuels and dry pyro-recycling technologies as optimal technologies for thefuture fast reactor cycle which promises a high level of nuclear proliferation resistance and reduction of the environmental load ofnuclear waste.
[Principal Results]
・As part of the process demonstration test using plutonium, continuous tests to recover uranium and plutonium from the conversion
process from oxides to metal and the electrolytic refining process were conducted and the relationship between the uranium andplutonium densities in molten salt and the electrolytic current was clarified under the condition of a material balance of some 95%
for uranium and almost 100% for plutonium.
・Cycling data in the transitional period from light water reactors to fast reactors was analysed to clarify the inter-annual changes of
the reprocessed amounts of the spent light water reactor fuel required for the transition to fast reactors and of the spent fast reactor
fuel and changes of the storage volume of plutonium and the spent fuel over the years.

4・Metal fuel containing transuranic nuclides which had undergone the irradiation process at the Phenix Reactor was transported fromthe French Nuclear Agency to the Transuranic Element Research Institute in Germany to commence non-destructive testing.
(10) Establishment of design assessment technology for its practical application to small fast reactors: FY 2006
[Objectives]
To apply for an advance review by the NRC based on the safety assessment of a sodium cooled ultra-small fast reactor (4Sreactor) and the development of innovative technologies with a view to contributing to the establishment of a fast reactor design
assessment technology.
[Principal Results]
・Technical documents were jointly prepared with the manufacturers for the application for an advance review of the 4S reactor byUS NRC and a technical scenario was created for the safety assessment to obtain the necessary licence.

2. Advanced Maintenance Technology: Rational Operation of Electric Facilities(1) Establishment of a gas turbine hot parts integrity assessment technology (promoted project subject): FY 2005 - FY 2009
[Objectives]
To develop a technology capable of assessing the remaining life of gas turbine hot parts (mainly the rotor and stator blades)
based on the usage temperature of individual parts to thoroughly reduce the maintenance cost.
[Principal Results]
・The applicable temperature range of the temperature estimation method to thermal barrier coating (TBC) based on the state of thecut surface degradation of the corrosion resistance coating layer was clarified to enable estimation of the temperature of gas turbine
blades with film cooling path system and coated with TBC. At the same time, temperature estimation was conducted by means ofcomputer simulation and the suitability of the above-mentioned method was confirmed through comparative analysis. Moreover, anew TBC degradation indicator was introduced to upgrade this method.
・“The maintenance planning optimisation support system” was further improved to enable the more flexible planning of the parts
rotation to respond to the demands of users.
(2) Development of remote monitoring and integrity assessment system for power generation plants: FY 2006 - FY 2008
[Objectives]
To develop a tool capable of monitoring the secular deterioration of equipment, assessing the impacts of changing equipment efficiency on plant efficiency and assessing the state of quantitative damage to boiler materials by means of analysing thethermal material balance through real time remote monitoring so that the thermal efficiency of power generation plants can bemaintained/improved along with improvement of the efficiency of maintenance work.
[Principal Results]
・To make the already developed “analysis of the thermal efficiency of power generating system (EgWin)” more easily usable atactual power plants, its macroscopic functions were developed and the user support system was expanded.
・Trial operation commenced with a system which analyses the thermal efficiency by receiving the operating data of a power plant
on-line and then produces a database from the analysis results. In addition, “plant computer data analysis tools” was developed to
conduct heat absorption analysis at each heat transfer surface and balance evaluation of the combustion gas flow based on thereceived plant computer data.
・A basic technique to individually evaluate the performance of the compressor and turbine, which are element equipment of a gas
turbine (GT), using such a dimension-less number as the Mach number was developed for its application to the inference of thecauses of a declining pressure ratio of a GT compressor.
(3) Development of non-destructive assessment methods for power generation components and steel structures (promoted projectsubject): FY 2004 - FY 2008
[Objectives]
To develop crack detection, analytical assessment and monitoring methods for power generation components and steel structures to contribute to the preservation of the reliability and cost reduction of maintenance work.
[Principal Results]
・The growth behaviour of voids and minute cracks was clarified by means of the creep fatigue test of boiler piping high chromium
steel welding joint inside a scanning electron microscope and a simulation method based on such behaviour was proposed. Throughcomparison with the fatigue test results, it was verified that this method is capable of predicting growth behaviour.
・Using a ultrasonic propagation simulation programme, the measuring conditions of the convenient and low cost ultrasonic inspection method (SPOD method) developed by the CRIEPI were optimized and the ability of this method to measure some 2 mm deep
fatigue cracks, which are difficult to measure by other existing methods, was verified.
・The development of “Bending and Internal Pressure on real structural samples” was completed. This facility is one of the world slargest base research facilities which enables the internal pressure and bending load testing of piping specimens equivalent to thereheat pipes of boilers at thermal power plants.
(4) Establishment of an autonomous demand area power system technology (promoted project subject): FY 2006 - FY 2007-12-08[Objectives]To clarify the impacts of the mass connection of distributed power sources in a wide area and to establish a control technology for the demand area power (distribution) system in order to ensure efficient energy supply through coordination between system

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