Intergranular stress corrosion cracking behavior of niobium-added Type 308 stainless steel weld overlay metal in a simulated BWR environment

Type 308 stainless steel weld metal as an internal cladding of reactor pressure vessels for boiling water reactors is subject to postweld heat treatment during fabrication and can suffer sensitization depending on carbon and ferrite contents. This sensitization can be avoided by using niobium-added Type 308 weld metal (specified as Type 308 NbL) which was developed for one-layer overlay welding application. In the present study, stress corrosion cracking (SCC) behavior of heat-treated Types 308 and 308NbL weld metals in oxygenated high temperature pure water was evaluated by slow strain rate test and U-bend tests with and without crevice. Every test showed that Type 308NbL weld metals were highly resistant to SCC compared to ordinary Type 308 weld metals. In single U-bend test, one-layer overlay weld metals of Type 308NbL produced by electroslag welding process using wide strip electrodes were crack free over 23,000 h. The U-bend test data of ordinary Type 308 weld metals were successfully analyzed by an SCC reaction model. Using this analysis, the SCC life margin for Type 308NbL over ordinary Type 308 weld metals, expressed as a ratio of respective times to SCC initiation, was estimated to be about 36.     

Experimental studies of 4-inch pipe whip test under BWR loca conditions

Pipe whip tests or jet discharge tests have been performed at the Japan Atomic Energy Research Institute, which simulate the instantaneous circumferential guillotine break of primary coolant piping in nuclear power plants. The present paper describes the results of the pipe whip tests using test pipes of 4 inch diameter, under the BWR LOCA conditions, which were performed from 1979 to 1981. The tests were carried out at an initial pressure of about 6.8 MPa and an initial temperature of about 285°C.The test pipe was 114.3 mm (4 in) in diameter, 8.6 mm in thickness and 4500 mm in length. The four pipe whip restraints used in the tests were the U-bar type of 8 mm in diameter and fabricated from Type 304 stainless steel. The experimental parameters were the clearance (30, 50 and 100 mm) and the overhang length (250, 400, 550 and 1000 mm).The main purpose of these tests is to investigate the effects of the clearance and the overhang length on the pipe whip behavior. It has been clarified from the test results that a smaller clearance and a shorter overhang length causes the deformation of the pipe and restraints to be minimized, and the test pipe collapses near the setting point of the restraints with the overhang length of 1000 mm.     

BWR pipe crack control using hydrogen water chemistry: Status report on Dresden-2 program

One of the proposed remedies for intergranular stress corrosion cracking of stainless steel piping in BWRs is an alternative water chemistry called hydrogen water chemistry (H₂WC) that involves suppression of reactor water dissolved oxygen to ≤ 20 ppb via hydrogen injection to the feedwater in conjunction with control of conductivity to ≤ 0.3 μ mho/cm. A long-term verification program, over two or three 18 month fuel cycles, was started at Commonwealth Edison's Dresden-2 reactor in April 1983 (Cycle 9). This paper describes the results of the water chemistry changes, structural material and fuel evaluations, and plant radiation level changes during Cycle 9, which ended in October 1984.To date the results of the verification program are very encouraging. They indicate that the alternative water chemistry, based on hydrogen additions to the feedwater to suppress oxygen and low conductivity, can be maintained in a large operating BWR, and that it does mitigate IGSCC in stainless steel recirculation piping. Monitoring of fuel and plant materials will continue in Dresden-2 at least through Cycle 10 to confirm the absence of any unusual side effects of this remedy for IGSCC.     

Japanese efforts in relation to BWR pipe cracking

The stress corrosion cracking was found in a Japanese Commercial BWR power plant first in 1974, then also in the others. To overcome this problem, much effort has been made to improve the stress condition, material and environment which are known to be the three causative factors of IGSCC. Also several studies have been carried out from the fracture mechanics point of view on the stress corrosion cracking behaviors in the BWR primary piping. Our experiences of pipe cracking, development of remedies and/or countermeasures examples of their application to our units and the fracture mechanistic research on piping LBB are shown in the following.     

SCC testing of pipe materials in BWR environment

A test loop has been installed in Ringhals 1 BWR, including facilities for Constant Elongation Rate Testing (CERT) and Electrochemical Potential (ECP) measurements in primary reactor water at reactor operation temperature. The loop is designed as to minimize transport time for reactor water from the reactor pressure vessel to the specimens being tested. Thus the testing environment is representative of the primary piping systems of BWRs, also with regard to short-lived constituents like hydrogen peroxide.The test program, which is in progress, has covered seven tests during start-up conditions or during power operation with presently current reactor water chemistry. In this presentation only CERT testing results on heavily sensitized austenitic chromium—nickel stainless steel are presented, although many other materials have been tested.Results show sensitized austenitic stainless steel is more prone to intergranular stress corrosion cracking (IGSCC) in actual than in simulated BWR environment and that start-up environment is chemically more aggressive than power operation environment. Reproducibility of the CERT technique as used is excellent.     

Detection of embedded fatigue cracks in Inconel weld overlay and the evaluation of the minimum thickness of the weld overlay using eddy current testing

This study evaluates the applicability of eddy current testing to the detection and sizing of fatigue cracks embedded in Inconel weld overlays. Welded plate specimens, which model head penetration welds and their weld overlays, are fabricated, and fatigue cracks are artificially introduced into the specimens. Eddy current inspections are performed using a uniform eddy current probe driven with 10 kHz, and all of the fatigue cracks are detected with clear signals. Subsequent numerical inversions estimate that the minimum thicknesses of the weld overlays are 1.47, 2.17, and 2.23 mm, whereas true thicknesses revealed through destructive testing are 1.51, 3.25, and 2.10 mm, respectively. Thicknesses are also evaluated using potential drop and ultrasonic testing methods; the results demonstrate that eddy current testing is the most efficient of the three methods.     

Fracture behavior of a neutron-irradiated stainless steel submerged arc weld cladding overlay

Stainless steel weld overlay cladding was irradiated at temperatures and fluences relevant to power reactor operation. The cladding was applied to a pressure vessel steel plate by the submerged arc, single-wire, oscillating-electrode method. Three layers of cladding were applied. The first layer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by excessive melting of the base plate. Charpy V-notch and tensile specimens were irradiated at 288°C to a fluence of 2 × 10²³ neutrons/m² (> 1 MeV).When irradiated, both types 308 and 309 cladding increased 5 to 40% in yield strength and slightly increased in ductility in the temperature range from 25 to 288°C. All cladding exhibited ductile-to-brittle transition behavior during impact testing caused by temperature dependent failure of the δ-ferrite phase. The type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation. Conversely, the impact properties of the specimens containing the highly diluted type 309 cladding, microstructurally similar to that produced during some off-normal welding conditions in existing reactors, experienced significant increases in transition temperature and drops of up to 50% in upper-shelf energy.     

A review on the clad failure studies

In this paper, an attempt has been made to systematically organize the research investigations conducted on clad tube failure, so far. Before presenting the review on the clad failure studies, an introduction to different clad materials has been added, in which the effect of alloying elements on the material properties have been presented. The literature on clad failure has been broadly categorized under the headings LOCA and RIA. The failure mechanisms like creep, corrosion and pellet-clad interaction have been discussed in details. Each subsection of the review has been provided with summary table, in which the studies are arranged in the chronological order. A small section on acceptance criteria for ECCS has also been included. The last section of the review has been dedicated to the core-degradation phenomena.     

Experimental studies of core flow fluctuations and neutron noise in a BWR

Using the instrumented fuel assemblies (IFA) installed in the Japan Power Demonstration Reactor (JPDR)-II core, fluctuations of the inlet and outlet channel flow rates were observed under both conditions of at-power operation and cold core flow circulation. The correlation analysis revealed that the flow fluctuations in any IFA channel showed almost uncorrelated cross-covariance function with other IFA channel flow. To explain the mechanism of the channel flow fluctuations, some hypothetical idea is introduced.     

The general corrosion behavior of ENiCrFe-7 weld overlay cladding material in deoxygenated high-temperature and high-pressure water

The general corrosion behavior of Alloy ENiCrFe-7 in deoxygenated high-temperature and high-pressure water was investigated. The results showed that the precipitates of Alloy ENiCrFe-7 included niobium carbide and Al-Ti-O compounds, and the weight gain increased fast firstly before 2250 h, then the weight gain slowed down. There were obvious large particles spread on denser oxide film after 3000 h exposure. Ni was present at a single chemical metallic Ni state, Feⁿ⁺ content of the outer layer was close to 60%, which was much higher than that of the matrix. The oxide film consisted of an inner layer and an outer layer, the inner layer was mainly composed of Cr₂O₃ and the outer layer was mainly composed of Fe₃O₄ and FeCr₂O₄. Finally, it is found that the preferential corrosion location of pitting was niobium carbide precipitates by in same site observation, while Al-Ti-O compounds was not dissolved in deoxygenated high-temperature and high-pressure water for 1500 h exposure. The size and number of the pitting was not significantly changed with increasing exposure time.     

Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72–1.92 × 10²⁰n/cm² (E>1 MeV) and 2.03×10²¹n/cm² (E>1 MeV) at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 ×10²¹ n/cm² (E>1 MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.     

Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72?1.92 × 10²⁰ n/cm²($\text{E > 1}$ MeV) and 2.03 × 10²¹ n/cm² ($\text{E > 1}$ MeV)at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 × 10²¹ n/cm² ($\text{E > 1}$ MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.     

Fracture margin of pipe with detectable crack by leakage

This paper describes a theoretical method for calculating a detectable crack size by leak detection systems in BWR plants. Crack opening areas for carbon steel pipes of various diameters containing circumferential through-wall cracks are analyzed. It was shown that large diameter pipes have a much higher safety margin, and that the 0.1A Criterion (10% of pipe cross-section) for postulated leak cross-sections gives a conservative estimate.     

高通量工程试验堆出口管道焊缝破口频率计算

近年来,国内外进行多项研究堆概率安全分析,其中管道破口导致的失水事故是堆芯损坏的重要风险来源。本文参考管道破口计算程序PRAISE(Piping Reliability Analysis Including Seismic Events)方法,选取压力壳型研究堆——高通量工程试验堆(High Flux Engineering Test Reactor,HFETR)的运行工况,对其反应堆冷却剂出口管道的焊缝进行分析,得到运行中该处焊缝发生各类破口的频率。     

Leak-before-break and plastic collapse behaviour of statically indeterminate pipe system with circumferential crack

Much research has been carried out on Leak-Before-Break (LBB) behavior of pipes with cracks. However, most studies have been made on statically determinate pipe systems. Few studies have been made on LBB behavior of statically indeterminate pipe systems. Most pipe systems in nuclear power plants have supports and restraints, thus they can be considered as statically indeterminate pipe systems. From above points of view, LBB and plastic collapse behaviors of statically indeterminate pipe with circumferential crack and compliance were studied in this paper. A new method is proposed to analyze and evaluate the LBB and plastic collapse behavior of a statically indeterminate structure. The pipe system of which one end is clamped and the other is supported with compliance was analyzed. The main results obtained are as follows: (1) By combining the limit analysis theory and elastic–plastic fracture mechanics, the effects of crack size, compliance and fracture toughness on load deflection behaviors to failure and structural integrity of statically indeterminate pipe system have been analyzed quantitatively and easily. (2) When a crack grows in a statically indeterminate pipe before plastic collapse, load drop conditions can be derived quantitatively, as a function of J_IC, dJ/da, flow stress, crack size, pipe span length, compliance and flexural rigidity of the pipe. (3) The analytic method developed in this research is useful and convenient to evaluate the LBB and tearing instability behavior of a statically indeterminate pipe system. (4) LBB resolves easily for statically indeterminate pipes with a crack, even when it does not resolve for statically determinate pipes with the same crack. That results from the fact that bending moment redistribution during the fracture process occurs easily for statically indeterminate pipe systems, and its redistribution restrains plastic deformation of the cracked weak section.     

BWR mechanics and materials technology update

This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR.In this paper, specific results and accomplishments are summarized to provide a broad perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration.     

An assessment of the BWR fuel box-lattice-cell homogenization procedures with the Tarapur BWR incore and excore measurements

A code system has been developed to provide the incorefuel-management guidelines to the Tarapur BWR reactors. Constant checking of the design calculational methods is rendered possible by the steady flow of operating data from the Tarapur units over the last few cycles. The operating data include cold/hot criticals and detailed flux/power maps. Besides these, the burnups and isotopic composition of a few irradiated fuel pins obtained by mass-spectrometric analyses, are also available for validation of the BWR core and lattice-cell modelling.The calculated eigen values for different power levels and at different core average burnups are found to have a spread of less than 0.25% ΔK. Analyses of a number of TIP measurements show that the core power distribution can be predicted in a satisfactory manner for uncontrolled fuel bundles and non-peripheral fuel assemblies (<10%). For prediction of cold-criticals the void-history effects are found to be unimportant.The pin burnups and isotopic densities of important U and Pu isotopes relative to ²³⁸U have been compared with mass-spectrometric measurements. The pin-burnup profile comparison is found to be good for fuel pins, which are not near water gaps. Deviation histograms of various isotopes are presented in this paper. ²³⁵U is predicted within ± 3% (r.m.s.). The total Pu is overpredicted by 5–8%, while the quality of Pu is predicted within ± 1.0% (r.m.s.).