Nuclear accident and nuclear terrorism are of great threats to international security. The countermeasures against nuclear emergency situation should be deliberately considered before it happened. In accordance with experience from Fukushima Daiichi Accident, various kinds of complex situations must be considered at the phase of emergency preparedness. And all related schedules, including national laws and regulations, allocation of responsibility and authority, emergency plans and procedures, facilities, instruments, trainings and exercises must be examined. The appropriateness of the emergency preparedness and the emergency response capability need to be verified through emergency exercises. However, exercises are usually costly and labor-intensive. So, it is necessary to design a software system to help conduct emergency exercises simulation. A Nuclear Emergency Decision Deduction and Training Platform (NEDDTP) is designed in China Nuclear Security Center of Excellence, which aims at performing the simulation of nuclear emergency process and providing a training platform for emergency workers. 3D models, the environment of nuclear power plant (NPP), some physical numerical models for simulating and several typical nuclear emergency scenarios for training are included in the platform. NEDDTP realizes the function of digitization of scenario, emergency simulation, resource management, command and dispatch, process record and evaluation in exercise, and it also provide VR display for emergency workers.

An urgent lesson learned from Fukushima Daiichi accident is what can happen by natural disaster can also occur by human design. The accident raised a fear that terrorists could cause a similar accident by acts of sabotage against nuclear power plant (NPP) and it is noticeable that threats of terrorism for nuclear security are increased after the accident. When considering sabotage, the prime threat to nuclear power plants, due attention should be paid to sabotage by insiders. Generally, insiders are the individuals with authorized access to nuclear facilities in transport who could attempt unauthorized sabotage. They could take advantage of their access authority and knowledge, to bypass dedicated physical protection elements or other provisions [1]. Thus, we should value the catastrophic consequences of the attack or act of insider sabotage which may lead to loss of safety functions of NPP. International Atomic Energy Agency (IAEA) indicated that the physical protection system (PPS) of a nuclear facility should be integrated and effective against both sabotage and unauthorized removal. The primary PPS functions are deterrence, detection, delay and response. It is noticeable that if detection failed, delay and response would become invalid. Thus, detection of insiders’ sabotage should be enhanced. Considering current countermeasures of PPS to insiders’ sabotage, the most significant challenge is how to distinguish ordinary maintenance behaviors and malicious behaviors since some malicious behaviors may hidden in ordinary maintenance behaviors. It appears that hand behavior has high contribution to human activity and a significant portion of maintenance behaviors and malicious behaviors. In this study, we proposed a hand behavior detection algorithm for insiders’ malicious behaviors for nuclear security [2]. We focused on the fact that the hand shape is uniquely determined by the fingertip coordinates. First, the depth image of the hand was captured with Kinect v2, and after removing the five fingers were remained by removing the palm and wrist parts, and the five fingers were identified using the K-means clustering [3], and the farthest point of each finger from wrist pixel was taken as the fingertip coordinates. The fingertip coordinates of the five fingers were combined for 60 frames to be time-series data, and this was used as the training data of the neural network. Time-series data obtained from five kinds of behaviors of five hands was used for training data. For the machine learning method, the Stacked-Auto Encoder (SAE) [4–5] which is one of popular methods was used. It extracts the feature of input data at intermediate layer of the first stage. In the second layer, the extracted feature is input and its feature is extracted to be used as the input of the softmax layer for pattern classification. Meanwhile, a real-time fingertip tracking system was developed and time-series data of each fingertip was successfully obtained with 29.8fps using MATLAB whose CPU was Intel Xeon Processor E5-2630v4 (25M Cache, 2.20 GHz). Moreover, a time-series data analysis based behavior recognition method was developed and all assumed malicious behaviors were detected with high accuracy (82.555% in overall) and speed (0.0023 seconds per frame) in the same computing environment. Also, robustness of the behavior recognition method was verified.

Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant. Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model. The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.

The natural circulation small modular reactor (NCSMR) is a 330 MW reactor which has no reactor coolant pumps (RCP) and no active safety injection systems at all. The reactor is mainly comprised of the reactor pressure vessel (RPV) with integral pressurize r and steam generator. RPV is enclosed by a vacuumed pressure containment vessel (PCV) and the PCV is submerged in the underground containment pool. A MELCOR model and corresponding input deck are developed for the RPV, PCV, and containment pool. The containment pool takes the role of ultimate heat sink (UHS) in accident situations. The containment pool may crack and leak in some critical accidents as the earthquake, leading to the severe accident of the reactor. A TMI-2 like SBLOCA in the RPV (stuck open RVVs) along with the containment pool crack (loss of ultimate heat sink) is simulated in the work. So me key parameters as the RRVs stuck open fraction, the PCV-SRVs open or not, the containment pool crack position would have large influence on the severe accident sequence. The sensitivity of these parameters to the accident sequence is analyzed in the work. According to the simulation results, the RPV pressure decreased with the RRVs stuck open. The depressurization of RPV accelerated with the RPV-SRV open fraction increase. The PCV pressure increased after that. Two cases as the PCV-SRV open after PCV pressure increase to 5 MPa, and PCV break while the RV d id not open, are analysis. The coolant discharge mass flo wrate in RPV and PCV were different in two cases, leading to the different degradation situation of the core. Since the containment pool is so important for the accident mitigation, sensitivity analysis is done for the containment pool crack position in the pool. The work will be meaningful in gaining an insight into the detailed process involved. One of the final goals of this work would be to identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents during the design process of NCSMR.

After the 2011 Fukushima accident, engineers of nuclear power plants are looking beyond the basic design requirements and ensuring that countermeasures are built in to avert possible nuclear accidents. In seismic probabilistic risk assessment (SPRA), uncertainties can be classified in two ways as aleatory uncertainties or epistemic uncertainties. To improve the reliability of SPRA, the difference in seismic response due to difference of building modelings related to epistemic uncertainty was focused on. Two modeling methods were used for a seismic response analysis: a three-dimensional finite-element model and a conventional sway-rocking stick model. Simulated input ground motions related to aleatory uncertainty were generated for the input waves. Then, the seismic floor response results of the various input ground motions of the two modeling methods were quantified. For the uncertainty quantification related to the different building modelings, a statistical analysis of the floor response results of the nuclear reactor building were further performed. Finally, for the quantification of the uncertainty in the fragility analysis for SPRA, the way to use of these results were discussed.

The new licensing standards were further improved by taking into account of lessons learned from the Fukushima-Daiichi nuclear accident, and countermeasures against severe accidents were newly required as regulatory items, where severe accidents were defined as serious accidents that occur under conditions exceeding design bases. Organic solvent fire in cell was defined as one of the severe accidents in nuclear fuel reprocessing facilities, which should be investigated, in order to establish methods for evaluating effectiveness of the countermeasures. One of the combustibles in the fire accident at reprocessing facilities is the organic solvent composed of 30% tributyl phosphate (TBP) and 70% dodecane. When the solvent burns, aerosol of soot and radioactive substances are released inside the facility. The aerosol causes a clogging of high-efficiency particulate air filters (HEPA filters) in a ventilation system of the facility, which increases a differential pressure of the filters. We have performed combustion tests simulating the fire accident. As one of interesting results of the tests, we observed, when most of dodecane in the solvent was burned out, a rapid increase in a differential pressure of a HEPA filter, which may cause its rupture. We also found a small amount of RuO 4 release from the burning solvent, which can pass through HEPA filters due to its volatility. These phenomena should be adapted in the effectiveness evaluations of the countermeasures against the fire accident.

The loss-of-reactor-level (LORL) is a typical accident condition causing a severe accident (SA) in sodium-cooled fast reactors (SFRs). In the loop-type SFR Monju, important cause of the LORL is the second coolant leakage at the low elevation of the primary heat transport system (PHTS), which occurs in cold standby in a different loop from that of the first coolant leakage in rated power operation because of excessive declining of the sodium level. This study developed an evaluation method for the LORL with considering countermeasures to prevent LORL: i.e., pumping sodium up into reactor vessel (RV) and interruption of sodium flowing out by siphon effect. To evaluate the effectiveness of the countermeasures, the transient behavior in the RV sodium level was analytically studied in representative accident sequences. The representative sequences with lowest sodium level were selected by considering combinations of possible coolant leakage positions. The evaluation result clarified that the LORL can be prevented by conducting the above-mentioned countermeasures to maintain the RV sodium level sufficient for the operation of decay heat removal system, even after the second coolant leakage of PHTS.

In order to improve the technical level of nuclear and radiation safety personnel, strengthen the nuclear safety culture, safeguard nuclear and radiation safety, China established Registered Nuclear Safety Engineer Qualification License System (hereinafter referred to as the registered qualification system). The registered qualification system has been implemented for more than ten years. The paper briefly discusses the background and situation of the construction of the registered qualification system, introduces the achievements and problems, and focuses on the analysis of the key factors restricting the development of the registered qualification system. Finally, some pertinent suggestions are given.

In the Fukushima Dai-ichi nuclear power station, Loss of Ultimate Heat Sink (LUHS) was caused by the great east japan earthquake and the subsequent tsunami [1]. It resulted in severe accident in three units. In that time, fuel damage in Spent Fuel Pool (SFP) were prevented by the various countermeasures such as makeup by pump truck and recovery of injection systems /cooling water system. In the past, Probabilistic Safety Assessment (PSA) has been developed with a focus on the reactor. After the accident, it has been acknowledged that SFP PSA is important to enhance the plant safety. In this study, probabilistic assessment is performed to suggest countermeasures for LUHS to SFP.

The generic concept of Security Controls, as initially deployed in the information security domain, is gradually used in other business domains, including industrial security for critical infrastructure and cybersecurity of nuclear safety I&C. A Security Control, or less formally, a security countermeasure can be any organizational, technical or administrative measure that helps in reducing the risk imposed by a cybersecurity threat. The new IAEA NST036 lists more than 200 such countermeasures. NIST SP800-53 Rev. 4 contains about 450 pages of security countermeasure descriptions, which are graded according to three levels of stringency. In order to facilitate and formalize the process of developing, precisely describing, distributing and maintaining more complex security controls, the Application Security Controls (ASC) concept is introduced by the new ISO/IEC 27034 multipart standard. An ASC is an extensible semi-formal representation of a security control (e.g. XML or JSON-based), which contains a set of mandatory and optional parts as well as possible links to other ASCs. A set of Application Security Controls may be developed by one company and shipped together with a product of another company. ISO/IEC 27034-6 assumes that ASCs are developed by an organization or team specialized in security and that the ASCs are forwarded to customers for direct use or for integration into their own products or services. The distribution of ASCs is supported and formalized by the Organization Normative Frameworks (ONF) and Application Normative Frameworks (ANF) deployed in the respective organizational units. The maintenance and continuous improvement of ASCs is facilitated by the ONF Process and ANF Process. This paper will explore the applicability of these industry standards based ASC lifecycle concepts for the nuclear domain in line with IEC 62645, IEC 62859 and the up-coming IEC 63096. It will include results from an ongoing bachelor thesis and master thesis, mentored by two of the authors, as well as nuclear specific deployment scenarios currently being evaluated by a team of cybersecurity PhD candidates.

A Probabilistic Risk Assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan, but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is intended to develop PRA methodology for a combination of low temperature and snow for a Sodium-cooled Fast Reactor (SFR) that uses the ambient air as its ultimate heat sink for decay heat removal under accident conditions. Annual excess probabilities of low temperature and of snow are statistically estimated based on the meteorological records of low temperature, snow depth and daily snowfall depth. To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., plugged intake and/or outtake for the DHRS and for the emergency diesel generator (EDG), unopenable door on the access routes due to accumulated snow, failure of the intake filters due to accumulated snow, possibility of freezing of the water in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around the intake and outtake induced by loss of the access routes, this study has investigated effects of electric heaters installed around the intake and outtake as an additional countermeasure. By using the annual excess probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence is failure of the electric heaters and loss of the access routes for snow removal against the combination hazard at daily snowfall depth of 2 m/day, duration time (snow and low temperature) of 1 day.

Cybersecurity incidents are stressful, complex in nature, and frequently are not systematically considered in daily tasks. When correctly managed, operational readiness procedures make sure the availability of data required to successfully and quickly recover from a security incident while lessening the adverse effect. Therefore, protective measures, such as implementation of data diodes, are playing an essential role in defending I&C systems. In addition, applicability of the newest forensic and digital evidence-related standards to the nuclear domain is being evaluated. Results of such evaluation are being considered in the 3D and 2D modeling of cybersecurity relevant assets. The development of the new IEC 63096, downstream standard of IEC 62645, will also support the proposed evaluation and modeling. However, IEC 63096 covers not only forensic and incident management-related security controls but also a broad range of cybersecurity controls. This paper will further explore the security degree-specific selection and overall assignment of forensic-related security controls for the nuclear domain. Results from ongoing prototype developments will be used to demonstrate possible alternative selections and assignments along with their contribution to different security metrics.

The introduction of soft controls into the main control room may provide a more convenient environment for operations, but also may introduce new types of human errors and new risks into the nuclear power plants. In this paper, taking the Chemical and Volume Control System (CVCS) at pressurized water reactor nuclear power plant as an example, the task analysis of operating procedures for starting the CVCS under a normal shut-down operation condition, the potential failure modes during operators executing each step of operating procedures and the dependency analysis of sub-tasks are presented. Furthermore, the sensitivity analyses are conducted for identifying probability importance and critical importance of each model parameter. The countermeasures for preventing and reducing the human errors of soft controls are discussed.

Nowadays cyber security assurance is one of the key challenges of safety critical software based NPP I&C (Nuclear Power Plants Instrumentation and Control) systems requirements profiling, development and operation. Any I&C system consists of a set of standard software (SW), hardware (HW) and FPGA components. These components can be selected and combined in different ways to address the particular control and safety assurance related tasks. Some of them are proprietary software (PS) and commercial off-the-shelf (COTS) components developed previously. Application of such components reduces the level of safety and cyber security, because they can contain vulnerabilities that were created intentionally. In this case, targeted attacks can lead to a system failure. National Vulnerability Database (NVD) and other open databases contain information about vulnerabilities which can be attacked by insiders or other intruders and decrease cyber security of NPP I&C systems. In this paper, we propose a safety assessment technique of NPP I&C systems, which consists of the following procedures: 1. Analysis of I&C architecture to assess influence of OTS component failures on dependability (reliability and safety) of the system. For that purpose, FMEDA or similar techniques can be applied. As a result, three-dimension criticality matrixes (CM) (with metrics of detection, probability and severity) are developed for different components (SWFCM and HW/FPGAFCM). 2. The IMECA-based assessment of OTS components and their configuration. In this case, CMs (SWICM and HW/FPGAICM) describe the degree of failure component influence on cyber security. 3. Joining of criticality matrixes (SWFCM and HW/FPGAFCM, SWICM and HW/FPGAICM), impact analysis of components depending on degree of influence on cyber security and safety as a whole. 4. Developing of Security Assurance Case and selecting of countermeasures according to safety (cyber security)/costs criteria. The developed tool supports creation of criticality matrixes for each analyzed component of the system and I&C as a whole. Joining of criticality matrixes allows creating common matrix for system cyber security and functional safety. The tool supports decision making to optimize choice of countermeasures according to criterion of safety and security/cost criterion.

Among the countermeasures to reduce the impacts of cable fires, installing automatic fire detection and suppression equipment may be applicable for cable trays due to the overcurrent incident considering a single failure of the circuit breaker. To establish a method to construct such systems effectively, we applied two types of automatic fire suppression systems using gaseous and bubble foam agents. For cable tray fire suppression tests, we used the overcurrent test equipment installed in the high power short-circuit testing facility in the CRIEPI (Central Research Institute of Electric Power Industry) in Japan. Several series of tests subjected to 2kA-class overcurrent of six times the allowable current were performed using FR (Flame Retardant)/ Non-FR high-/low-voltage power cables. During these tests, to reproduce the cable fire, one target cable was continuously energized by ring transformers, which confirmed the reliable fire suppression capability of the cooling and suffocation.

The electrical disturbance in Forsmark 2006 [3] led to increased attention being paid to the power supplies of nuclear power plants and their role in safety system reliability, both nationally and internationally. Since then numerous disturbances similar in nature have occurred in the electrical power supply which raises questions whether best available technology (BAT) has been utilised in the design and analysis of the electrical power supply of the safety functions of nuclear facilities. On repeated occasions this type of disturbances has had an impact on redundant parts of several safety systems due to functional dependencies between these. The frequency of these occurrences has been unexpected. The Swedish Radiation Safety Authority (SSM) has decided to write this document to clarify the regulators position on this issue. The document is also intended to support SSM:s assessments and evaluations of the Licensees efforts regarding degraded power supplies. An assessment of nuclear power plant electrical power systems is necessary in the light of the past years’ operational experience [1–7], where disturbances in the electrical power supply on repeated occasions have caused a power supply with degradation severe enough to challenge plant safety. As the potential consequences of such a degraded power supply can be severe it must be proven that the frequency of such occurrences is tolerably low. Furthermore, it is important to consider experiences from known situations with degraded power supplies, to enable a reasonable approach to identify and take counter-measures based on the root-cause and ensure utilisation of best available technology. A sufficient approach to enable prevention, protection and mitigation against this type of disturbances has been difficult to identify. Actual events and conditions causing a degraded power supply have often been complex in nature and difficult to anticipate, wherefore events and conditions which has not yet occurred are difficult to foresee. For this reason it is deemed most effective to identify and implement proportional measures that enhances the independence of the power supplies, such that a degraded power supply with a higher reliability is prevented from propagating to multiple parts of the safety systems. In this memorandum, SSM describes a state-based approach to analysing electrical power system functionality in different states of degraded power supply. The approach is intended to identify potential design weaknesses and measures to enhance robustness. Such an approach is viewed as more favourable in facilitating the identification of such measures, which may otherwise be neglected due to an estimated low frequency of occurrence, or missed due to incomplete identification of possible events and conditions. Furthermore this document describes how an assessment of electrical power system design can be performed, where the lowest common denominator from operational experience e.g. [1–7] is identified and counteracted. Actual occurrences of degraded power supplies, which all have been “unknown during the event identification process” but “well-known electrical phenomena”, can be described as unidentified degrading conductive disturbances .

Fukushima Daiichi NPP accident would be terminated, if sufficient accident countermeasures, such as water proof door, mobile power, etc [1, 2]. In case of Europe, it had already installed the heat removal system and filtered containment venting system (FCVS) from the lessons of TMI and Chernobyl Accidents. The new regulatory standard in Japan, the filtered vent system (FCVS) should be installed, and prevent the radioactive material in case of the severe accident and the overpressure breakage prevention of a primary containment vessel (PCV) and also the robustization of the FCVS. The authors examined the severe accident process in the 2 nd unit of Fukushima Daiichi NPS, and found the vent by FCVS should be done before water injection into the core. The PCV spray and water injection into the pedestal basement should be also the countermeasures to the severe accident. Countermeasures for an intentional aircraft collision should be installed too. Upon occurrence of a severe accident (SA), vent gas with radioactive fission products is blown out to a scrubbing pool through numerous venturi nozzles. Mist in steam moves upward to a metal fiber filter through a multi-hole baffle plate. After the mist is removed by that filter, radioactive methyl iodine (CH 3 I) is captured on the surface of a molecular sieve or AgX, made from zeolite particles with silver coating. A FCVS visualized test facility was installed at Hokkaido University. An AgX filter is used down-stream of the scrubbing pool and metal fiver filter. Thickness of AgX filter is very important parameter to obtain enough decontamination factor (DF). The DF for the radioactive iodine exceeds 10,000 at bed depth (AgX filter thickness) greater than 75mm.

During a Loss of Coolant Accident (LOCA), the high energy jet from the break may impinge on surrounding surfaces and materials, producing a relatively large amount of fibrous debris (mostly insulation materials). The debris may be transported through the reactor containment and reach the sump strainers. Accumulation of such debris on the strainers’ surface can cause a loss of Net Positive Suction Head (NPSH) and negatively affect the Emergency Core Cooling System (ECCS) capabilities. The U.S. Nuclear Regulatory Commission (U.S.NRC) initiated the Generic Safety Issue (GSI) 191 to understand the physical phenomena involved in this type of event, and help develop the tools to prove the safety and reliability of the existing Light Water Reactors (LWR) under these conditions. Some nuclear power plants have already adopted countermeasures in an attempt to limit the effect of the debris accumulation on the ECCS performance, by replacing or modifying the existing strainer configurations. In this paper, two different strainer designs have been considered and sensitivity analysis was conducted to study the effect of the approach velocity on the pressure drop at the strainer caused by the debris accumulation. The development of the fibrous beds was visually recorded in order to correlate the head loss, the approach velocity, and the thickness of the fibrous bed. The experimental results were compared to semi-empirical models and theoretical models proposed by previous researchers.

As revealed by Fukushima Daiichi nuclear disaster, countermeasures against severe accidents in nuclear power plants are an urgent need. In particular, from the viewpoint of protecting containment and suppressing diffusion of the radioactive materials, it is most important to install filtered venting devices to release high pressure contaminated gas to the atmosphere with elimination radioactive materials in the gas. A Multi Venturi Scrubber System (MVSS) is one of the filtered venting devices, and used in European reactors [1, 2]. One of the main components of the MVSS is a Venturi Scrubber (VS). It is considered that a dispersed or dispersed annular flow is formed in the VS by a self-priming phenomena. In the self-priming phenomena, the liquid was suctioned from a surrounding region of the VS to the inside of the VS. And a part of the radioactive materials are eliminated through the gas-liquid interface of the dispersed or annular dispersed flow. Therefore, to consider the MVSS operation characteristics, it is important whether to occur the self-priming or not and the liquid flow rate of the self-priming of the VS. The objective of this paper is to understand the self-priming phenomena of the VS for the filtered venting. And theoretical analysis and experiment were conducted. By comparing these results, we discussed about the mechanism of the self-priming phenomena. As results, the self-priming phenomena in the VS was confirmed and, at a high gas flow rate, the suspension of the self-priming is confirmed experimentally and theoretically.

Primary Water Stress Corrosion Cracking (PWSCC) is one of important ageing issues in PWR (Pressurized Water Reactor) primary systems. It has been pointed out that high concentration dissolved hydrogen may lead to occurrence of PWSCC. The authors have proposed to substitute hydrogen by methanol as a fundamental countermeasure of PWSCC. So far corrosion tests of stainless steels and Zircaloy-4 in methanol solutions at 320 °C were conducted under γ-ray irradiation and without irradiation. The test results show that methanol is promising. In the present paper, γ-ray irradiation experiments of methanol solution at 320 °C were done up to 100 kGy. A study on the radiolysis of methanol solution is important from two aspects. One concerns corrosion of structural materials. The radiolysis of methanol may result in formation of harmful compounds to the structural materials, such as carboxylic acids. It is necessary to know the yields of such compounds. The other concern is possible polymerization of methanol and formation of organic polymer deposit on fuel claddings. Large amount of the deposit on fuel claddings should be avoided to keep integrity of fuel claddings. Therefore, it should be clarified whether gaseous species are major products and whether polymerized species of methanol such as ethylene glycol is formed. After the γ-ray irradiation of methanol solution, following species were analyzed: CO 2 and H 2 , methanol, formaldehyde, formate and acetate, and ethylene glycol and glycerin. Without γ-ray irradiation, the major process of the thermal decomposition of methanol at 320 °C is oxidation of methanol by water and generation of one CO 2 molecule and three H 2 molecules. Under γ-ray irradiation, the decomposition of methanol is accelerated; little methanol remains after 10 kGy irradiation. The major product is CO 2 , and polymerization of methanol unlikely occurs. After methanol is completely decomposed, the hydrogen yield still increases. The reducing environment is maintained. Probably, transient organic species play important roles. The addition of low concentration methanol may be sufficient to maintain reducing environment of the PWR primary systems.