Development of 2012 Edition of JSME Code for Design and Construction of Fast Reactors: (4) Creep–Fatigue Evaluation Method for Modified 9Cr–1Mo Steel

This paper describes a creep–fatigue evaluation method for modified 9Cr–1Mo steel, which has been newly included in the 2012 edition of the JSME code for design and construction of fast reactors. In this method, creep and fatigue damages are evaluated on the basis of Miner’s rule and the time fraction rule, respectively, and the linear summation rule is employed as the failure criterion. Investigations using material test results are conducted, which show that the time fraction approach can conservatively predict failure life if margins on the initial stress of relaxation and the stress relaxation rate are embedded. In addition, the conservatism of prediction tends to increase with time to failure. Comparison with the modified ductility exhaustion method, which is known to have good failure life predictability in material test results, shows that the time fraction approach predicts failure lives to be shorter in long-term strain hold conditions, where material test data is hardly obtained. These results confirm that the creep–fatigue evaluation method in the code has implicit conservatism.

Derivation of Design Fatigue Curves for Austenitic Stainless Steel Grades 1.4541 and 1.4550 Within the German Nuclear Safety Standard KTA 3201.2

Titanium and niobium stabilized austenitic stainless steels X6CrNiTi18-10S (material number 1.4541, correspondent to Alloy 321) respectively X6CrNiNb18-10S (material number 1.4550, correspondent to Alloy 347) are widely applied materials in German nuclear power plant components. Related requirements are defined in Nuclear Safety Standard KTA 3201.1. Fatigue design analysis is based on Nuclear Safety Standard KTA 3201.2. The fatigue design curve for austenitic stainless steels in the current valid edition of KTA 3201.2 is essentially identical with the design curve included in ASME-BPVC III, App I (ed. 2007, add. July 2008 respectively back editions). In the current code revision activities of KTA 3201.2 the compatibility of latest in air fatigue data for austenitic stainless steels with the above mentioned grades were examined in detail. The examinations were based on statistical evaluations of 149 strain controlled test data at room temperature and 129 data at elevated temperatures to derive best-fit mean data curves. Results of two additional load controlled test series (at room temperature and 288°C) in the high cycle regime were used to determine a technical endurance limit at 107 cycles. The related strain amplitudes were determined by consideration of the cyclic stress strain curve. The available fatigue data for the two austenitic materials at room temperature and elevated temperatures showed a clear temperature dependence in the high cycle regime demanding for two different best-fit curves. The correlation of the technical endurance limit(s) at room temperature and elevated temperatures with the ultimate strength of the materials is discussed. Design fatigue curves were derived by application of the well known factors to the best-fit curves. A factor of SN = 12 was applied to load cycles correspondent to the NUREG/CR-6909 approach covering influences of data scatter, surface roughness, size and sequence. In terms of strain respectively stress amplitudes in the high cycle regime, for elevated temperatures (>80°C) a factor of Sσ = 1.79 was applied considering and combining in detail the partial influences of data scatter surface roughness, size and mean stress. For room temperature a factor of Sσ = 1.88 shall be applied. As a result, new design fatigue curves for austenitic stainless steel grades 1.4541 and 1.4550 will be available within the German Nuclear Safety Standard KTA 3201.2. The fatigue design rules for all other austenitic stainless steel grades will be based on the new ASME-BPVC III, App I (ed. 2010) design curve.

Structural Integrity for Nuclear Power Plant Against Excessive Load on Design Extension Condition

This paper introduces an example of structural integrity evaluation for Light Water Reactor (LWR) against excessive loads on the Design Extension Condition (DEC). In order to assess the design acceptance level of DEC, three acceptance criteria which are the stress basis limit of the current design code, the strain basis limit of the current design code and the strain basis limit by using Load and Resistance Factor Design (LRFD) method were applied. As a result the allowable stress was increased by changing the acceptance criteria from the stress basis limit to the strain basis limit. It is shown that the practical margin of the LWR’s components still keeps even on DEC by introducing an appropriate criterion for integrity assessment and safety factors.

Time-History Analysis of a Super-Large Storage Tank Under Seismic Excitation

In this paper, two finite element models are established for a super-large storage tank with or without a floating roof on the medium level. Time-history analysis with consideration of fluid-solid coupling for the deformation of tank wall and medium sloshing during or after an earthquake is performed with the emphasis on the effects of the floating roof. It is found that the upper part of tank is more sensitive to the earthquake action than the lower part. The wind girders and the reinforcing rings play a big role in limiting the radial deformation of the upper part of the tank wall. The floating roof has little effect on the tank wall deformation, but it is effective in suppressing the medium sloshing during the earthquake. After the earthquake, the radial deformation of the tank wall attenuates quickly, but the sloshing attenuation of the medium presents a slow progress and the floating roof inhibits the sloshing attenuation of the medium.

Crack Growth Analysis of High-Pressure Equipment for Hydrogen Storage

Crack growth analysis (CGA) was applied to estimate the cycle life of the high-pressure hydrogen equipment constructed by the practical materials of 4340 (two heats), 4137, 4130X, A286, type 316 (solution-annealed (SA) and cold-worked (CW)), and type 304 (SA and CW) in 45, 85 and 105 MPa hydrogen and air. The wall thickness was calculated following five regulations of the High Pressure Gas Safety Institute of Japan (KHK) designated equipment rule, KHKS 0220, TSG R0002, JB4732, and ASME Sec. VIII, Div. 3. We also applied CGA for four typical model materials to discuss the effect of ultimate tensile strength (UTS), pressure and hydrogen sensitivity on the cycle life of the high-pressure hydrogen equipment. Leak before burst (LBB) was confirmed in all practical materials in hydrogen and air. The minimum KIC required for LBB of the model material with UTS of even 1500 MPa was 170 MPa·m0.5 in 105 MPa. Cycle life qualified 103 cycles for all practical materials in air. In 105 MPa hydrogen, the cycle life by KIH was much shorter than that in air for two heats of 4340 and 4137 sensitive to hydrogen gas embrittlement (HGE). The cycle life of type 304 (SA) sensitive to HGE was almost above 104 cycles in hydrogen, while the cycle life of type 316 (SA and CW) was not affected by hydrogen and that of A286 in hydrogen was near to that in air. It was discussed that the cycle life increased with decreasing pressure or UTS in hydrogen. This behavior was due to that KIH increased or fatigue crack growth (FCG) decreased with decreasing pressure or UTS. The cycle life data of the model materials under the conditions of the pressure, UTS, KIH, FCG and regulations in both hydrogen and air were proposed quantitatively for materials selection for high-pressure hydrogen storage.

Development of Regulation on Accident Management and Statistics Analysis on Accident for China Special Equipment

Regulations on safety supervision and accident management of special equipment since 1979 were analyzed in order to clarify the rules and characteristics of safety supervision for China’s special equipment. The accident management process of special equipment gradually tends to be standardized and procedural. In addition, the number development and accident evolvement trend of special equipment in China since 1979 were studied aiming at obtaining the characteristics and rules of the special equipment development, and a guideline is provided for the safety supervision of special equipment in China. On the whole, the accident rates and mortality rates per ten thousand equipments gradually decrease in the stages investigated in this paper, but compared with the corresponding accident rates of special equipment in developed countries, there is a space for the accident rates and mortality rates of China’s special equipment to decline.

The Further Development of National Safety Assessment Standard for In-Service Pressure Equipment Containing Defects

The fitness-for-service assessment is widely used to demonstrate whether identified defects or in-service deterioration threaten the structure integrity of the pressure equipment. This article introduces the development of the Chinese national safety assessment standard. The assessment procedure and assessment techniques of current Chinese national safety assessment standard GB/T 19624-2004 are demonstrated, and the improvement is proposed. In order to maintain availability of aging equipment and enhance the long-term economic performance of in service pressure equipment, a new Chinese national safety assessment standard system is methodology proposed based on the analysis of some international structure integrity procedures or standards. A new national standard fitness-for-service which focuses on the evaluation of in-service damage and degeneration will be established while GB/T 19624 is applicable for general defects.

Advanced Inlay System for Inlet/Outlet Nozzles of RV for Preventive Maintenance Against Alloy 600 PWSCC in Japanese PWR Plants

Because many nuclear plants have been in operation for ages, the importance of preventive maintenance technologies is getting higher. One conspicuous problem found in pressurized water reactor (PWR) plants is the primary water stress corrosion cracking (PWSCC) observed in Alloy 600 (a kind of high nickel based alloy) parts. Alloy 600 was used for butt welds between low alloy steel and stainless steel of nozzles of Reactor Vessel (RV), Steam Generator (SG), and Pressurizer (Pz). As PWSCC occurred at these parts may cause Loss of Coolant Accident (LOCA), preventive maintenance is necessary. PWSCC is considered to be caused by a mixture of three elements: high residual tensile stress on surface, material (Alloy 600) and environment. PWSCC can be prevented by improving one of the elements. MHI has been developing stress improvement methods, for example, Water Jet Peening (WJP), Shot Peening by Ultrasonic vibration (USP), and Laser Stress Improvement Process (L-SIP). According to the situation, appropriate method is applied for each part. WJP has been applied for RV nozzles of a lot of plants in Japan. However PWSCC was observed in RV nozzles during the inspection before WJP in recent years, MHI developed the Advanced INLAY system to improve the material from Alloy 600 to Alloy 690. Alloy 600 on the inner surface of the nozzles is removed and welding with Alloy 690 is performed. In addition, heat treatments for the nozzles are difficult for its structural situation, so ambient temperature temper bead welding technique for RV nozzles was developed to make the heat treatments unnecessary. This paper describes the specifications of the advanced INLAY system and introduces the maintenance activities which MHI has applied for three plants in Japan by March 2012.

Tests and Discussions on the Impact Work of the S30403 Austenitic Stainless Steel at 20°C and -196°C

In this paper, impact tests on the S30403 austenitic stainless steel at 20°C as well as −196°C were carried out. Both base plate and welded joint including weld and heat-affected zone are tested to measure the impact work KV2. It is found that when the temperature of the testing environment is decreased from 20°C to −196°C, both the impact work KV2 for the base plate and welded joint are decreased remarkably. Specifically, the impact work KV2 for the base plate decreases by 19–29% while that for the welded joint decreases by as much as 53.8%. In addition, impact tests with different size of specimens show the impact work KV2 with 5×10×55mm specimens is about 0.53–0.54 times that with 7.5×10×55mm specimens, much lower than 2/3, the ratio of two specimens’ cross section areas, indicating that rules in relevant steel or equipment standards regarding impact tests using small specimens need to be revised.