Seismic loads for crack stability assessment, in a review of leak-before-break (LBB) applicability

This study was performed to define the seismic loading conditions for use in the crack stability assessment of a BWR for the applicability of the leak-before-break (LBB) criterion. The LBB has been applied to the design of Class 1 piping in Japanese light-water reactors. Crack penetrated condition with detectable leak in the LBB applicability review is classified into the Level C service condition. Here an S₁-earthquake (maximum design earthquake) is currently assumed, rather than an S₂-earthquake (extreme design earthquake). In order to justify this assumption, the frequency of an S₁-earthquake occurring during coolant leakage due to crack propagation was determined. The frequency of coolant leakage from Class 1 piping must be less than that of the Level C service condition (2.5 × 10⁻² to 1 × 10⁻⁴ per year) in order to assume that an S₁-earthquake is appropriate. Accordingly, the frequency of coolant leakage from Class 1 piping was calculated using a probabilistic fracture mechanics (PFM). The results of this analysis indicate that the frequency of coolant leakage from Class 1 piping is less than the expected occurrence of an S₁-earthquake. As the results, it is concluded that the assumption of the seismic loading employed in an LBB applicability review should be appropriate.     

Influences of the dynamic strain aging on the J–R fracture characteristics of the ferritic steels for reactor coolant piping system

The leak-before-break (LBB) design of the piping system for nuclear power plants has been based on the premise that the leakage due to the through-wall crack can be detected by using leak detection systems before a catastrophic break. The piping materials are required to have excellent J–R fracture characteristics. However, where ferritic steels for reactor coolant piping systems operate at the temperatures where dynamic strain aging (DSA) could occur, the fracture resistance could be reduced with the influence of DSA under dynamic loading. Therefore, in order to apply the LBB design concept to the piping system under seismic loading, both static and dynamic J–R characteristics must be evaluated.Materials used in this study are SA516 Gr.70 for the elbow pipe and SA508 Cl.1a for the main pipe and their welding joints. The crack extension during the dynamic and the static J–R tests was measured by the direct current potential drop (DCPD) and the compliance method, respectively. This paper describes the influences of the dynamic strain aging on the J–R fracture characteristics with the loading rate of the pipe materials and their welding joints.     

Ontario hydro's leak before break approach: Application to the darlington (candu) nuclear generating station a

Ontario Hydro has developed a leak before break (LBB) approach for application to the large diameter heat transport piping for Darlington NGS A as an alternative to the provision of pipewhip restraints. This approach has been applied to pipe sizes which are equal to or greater than 530 mm (21 in. NPS). The proposed LBB approach incorporates assessments at several levels to provide assurance against catastrophic rupture. A comprehensive and systematic review of pipe failure mechanisms is considered the first important step in establishing role and applicability of the LBB concept. The elements integral to the approach are those related to demonstration of crack stability utilizing fracture mechanics methods and those related to leak rate predictions and leakage detection capability. For evaluation of crack stability the J-integral/tearing modulus (J/T) method has been selected. Results from an extensive material test program from actual heat transport piping, forgings, associated welds and heat affected zones as inpur to EPFM analyses provide the J-resistance and J—T curves. The details of EPFM analyses for a straight pipe with a circumferential crack and a piping elbow with a central longitudinal throughwall crack are presented here. Additionally, results of crack opening detail, the effects of crack face pressure, the predictions of LEAK RATE code and an assessment of the leakage detection capability are presented.     

Development of modified piping evaluation diagram for leak-before-break application to Korean next generation reactor

Recently, the piping evaluation diagram (PED) is accepted in nuclear industry for an efficient application of leak-before-break (LBB) concept to piping system at an initial piping design stage. The objective of this paper is to develop the modified PED, which can account for the variation of the material properties of the PED development stage and those of the assembly stage. For this purpose, a parametric study was performed to investigate the effect of stress–strain curve on the detectable leakage crack length and the effect of fracture resistance curve on the LBB allowable load. Finite element analyses were also performed to investigate the effect of stress–strain curve on the LBB allowable load. Finally, a modified PED was proposed as a function of crack length and the allowable safe shutdown earthquake load. The LBB analyses based on the modified PED are in good agreement with those based on the traditional PED. By adopting the modified PED, the variation of material properties can be considered in the LBB analysis and the computing times required for the application of LBB during the design process can be considerably reduced.     

Fracture mechanics evaluation of small diameter piping considering the latest experimental results

The results obtained from investigations carried out on austenitic piping of small nominal diameter (DN80 and DN50) are introduced and discussed together with their assessment using fracture mechanics methods. Essential results are summarised as following. The pipes with flaws (fatigue crack) down to a depth to a_max/t=0.51 (DN80) as well as a_max/t=0.62 (DN50) and a circumferential extension of results 2α=120° reached bending angles up to 26°. The ASME collapse load (test collapse load) was exceeded considerably and the experimental maximum load could not be reached. Failure due to a leakage or rupture did not occur in any test. The maximum crack extension was 0.69 mm (DN80, a_max/t=0.51) resp. 0.3 mm (DN50, a_max/t=0.62). The experimental maximum load can approximately be assessed by the limit analysis. The fracture mechanics approximation methods GE/EPRI and LBB/NRC calculated a/t=0.4 and 2α=120° initiation loads above the experimental maximum load for pipes containing flaws. These results confirmed the procedures for the proof of integrity of small diameter piping by updating information on load, deformation and failure behaviour of austenitic piping damaged with circumferential flaws. Using these results may formulate a final safety concept for the proof of integrity of small diameter piping by completing the current concepts.     

压力管道破前漏分析的一种简化方法

介绍了一种用于核反应堆管道和压力容器破前漏（ＬＢＢ）分析的简化方法，它主要以线弹性断裂力学的基础，将ＬＢＢ分析中应力强度因子，裂纹张开面积和泄漏率等计算以解析公式表达出来。这种方法使用方便，而且满足有关的国家标准和国际规范的要求，适用于一些应力分布和几何形状比较简单的管道和压力容器，或用于ＬＢＢ性质的近似估计。     

Plastic collapse and LBB behavior of statically indeterminate piping system subjected to a static load

The plastic collapse and LBB behavior of statically indeterminate piping system were investigated in this study, compared with those of the statically determinate piping system. Special attention was paid to evaluate the crack opening displacement after a crack penetrated wall thickness. The main results obtained were as follows:

1. The reduction of ultimate strength caused by a crack was relatively small in the statically indeterminate piping system. The main reason is thought to be that a sufficient redistribution of the bending moment occurs in this system.

2. A method to evaluate the crack opening displacement after crack penetration in a pipe with a non-penetrating crack was proposed. From this method, it was known that the crack opening displacement could be evaluated by using the incremental plastic rotation angle.

3. The acceptable defect size considering the deformation of a pipe was estimated by comparing the plastic moment at the defective part and the gross yielding moment at the non-defective part.

Results of reliability test program on light water reactor piping

The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event.In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping.In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.     

A new method for detecting pressure tube failures in Indian PHWRs

For the annulus gas system (AGS) of the standardised Indian pressurised heavy water reactor, an elaborate pressure tube (PT) crack monitoring and detection system is envisaged to ensure safety through leak-before-break. The parameters that are monitored relate to the detection of D₂O moisture leaking in from the primary heat transport (PHT) system through a cracked PT. Since a slow build-up of moisture in the AGS may also occur for reasons other than PT failure, it is desirable that a diverse measurement technique should be available. This paper suggests such a technique, based on the observation that a small reference concentration of fission gases is normally present in the annulus gas. This concentration would change sharply upon PT failure, when the heavy water from the leaking PHT system releases the dissolved fission gas content into the annulus. This paper presents a theoretical study of the parameters that influence the build-up of fission product noble gases in the AGS and shows that leakage rates as low as 10 g h⁻¹ from a PT crack can be detected in a few tens of minutes by this method. This is expected to substantially increase the available time between the leak detection and the PT failure, thus serving as an important tool in meeting the leak-before-break criterion of a critical component in PHWRs.     

Leak-before-break assessment of CANDU pressure tube considering leak detection capability

The leak before break (LBB) concept is employed as defense in depth in CANDU reactors in order to avoid an unstable failure in pressure tubes. The LBB in pressure tubes can be ensured when detection, confirmation and location of the leak are carried out and the reactor is placed in a depressurized condition before the crack exceeds critical crack length. Leak detection and location is provided by the annulus gas system (AGS). Therefore, the evaluation parameters for LBB assessment, such as leak detection and leak location capabilities, should be made available through AGS performance test. Recently, the AGS in-situ tests with a simulated moisture injection were carried out in one of the CANDU reactors in Korea. This paper presents the LBB assessment performed taking into account the leak detection capability and the modification of operating procedure therefrom.     

Effect of dynamic strain aging on the leak-before-break analysis in SA106 Gr.C piping steel

The characteristics of dynamic strain aging (DSA) on material properties used in leak-before-break (LBB) analysis were discussed. Using these material data, the effect of DSA on the LBB analysis was estimated through the evaluation of leakage-size crack and flaw stability in SA106 Gr.C piping steel. Also, the results were represented as a form of ‘LBB allowable load window'. In the DSA temperature region, the leakage-size crack length was smaller than that at other temperatures and it increased with increasing tensile strain rate. In the results of flaw stability analysis, the lowest instability load appeared at the temperature corresponding to the minimum J–R curve which was caused by DSA. The instability load depended on the loading rate of J–R data, and decreased with increasing tensile strain rate at the plant operating temperature. These are due to the strain hardening characteristic and strain rate sensitivity of DSA. In the ‘LBB allowable load window', the LBB allowable region at the temperature and loading conditions where DSA occurs was decreased by about 30% compared with that in other conditions.     

Observations on the effect of post-weld heat treatment on J-resistance curves of sa-106b seamless piping welds

Ontario Hydro has developed a leak-before-break (LBB) methodology for application to large diameter piping (21, 22 and 24 inch) Schedule 100 SA106B heat transport (HT) piping as a design alternative to pipe whip restraints and in recognition of the questionable benefits of providing such devices. Ontario Hydro's LBB approach uses elastic-plastic fracture mechanics (EPFM).In order to assess the stability of HT piping in the presence of hypothetical flaws, the value of the material J-integral associated with crack extension (J— R curve) must be known. In a material test program J-resistance curves were determined from various pipe heats and four different welding procedures that were developed by Ontario Hydro for nuclear Class 1 piping. The test program was designed to investigate and quantify the effect of various factors such as test temperature, crack plane orientation and welding effects which have an influence on fracture properties. An acceptable lower bound J-resistance curve for the piping steels and welds were obtained by machining maximum thickness specimens from the pipes and weldments and by testing side-grooved compact tension specimens. This paper addresses the effect of test temperature and post-weld heat treatment on the J-resistance curves from the welds.The fracture toughness of all the welds at 250°C was lower than that at 20°C. Welds that were post-weld heat treated showed high crack initiation toughness, J_lc, rising J-resistance curves and stable and ductible crack extension. Non post-weld heat treated welds, while remaining tough and ductile, showed comparatively lower J_Ic, and J-resistance curves at 250°C. This drop in toughness is possibly due to a dynamic strain aging mechanism evidenced by serrated load-displacement curves. The fracture toughness of non post-weld heat treated welds increased significantly after a comparable post-weld heat treatment.The test procedure was validated by comparing three test results against independent tests conducted by Materials Engineering Associates (MEA) of Lanham, Maryland. The J_Ic and J-resistance curves obtained by Ontario Hydro and MEA were comparable.     

A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

In modern CANDU nuclear generating stations, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the fuel bundles and the heavy water (D₂O) coolant. The pressure tubes operate at an internal pressure of 10 MPa and temperatures ranging from 250°C at the inlet to 310°C at the outlet. Over the expected 30 year lifetime of these tubes, they would be subjected to a total fluence of 3×10²⁶ n m⁻². In addition, these tubes gradually pick up deuterium as a result of a slow corrosion process. When the hydrogen plus deuterium concentration in the tubes exceeds the hydrogen/deuterium solvus, the tubes are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. The service life of the pressure tubes is determined, in part, by changes in the probability for the rupture of a tube. This probability is made up of the probability for crack initiation by DHC multiplied by the sum of the probabilities of break-before-leak and leak-before-break (LBB). A probabilistic model, BLOOM, is described which makes it possible to estimate the cumulative probabilities of break-before-leak and LBB. The probability of break-before-leak depends on the crack length at first leak detection and the critical crack length. The probability of a LBB depends on the shut-down scenario used. The probabilistic approach is described in relation to an example of a possible shut-down scenario. Key physical input parameters into this analysis are pressure tube mechanical properties, such as the crack length at first coolant leakage, the DHC velocity and the critical crack length. Since none of these parameters are known precisely, either because they depend on material properties, which vary within and between pressure tubes, and/or because of measurement errors, they are given in terms of their means and standard deviations at the different temperatures and pressures defined by the shut-down scenario.     

LBB和BP在核电站管道设计中的异同性分析

对LBB（破前漏）和BP（破裂排除）两种设计理念进行了比较，从材料质量要求、制造要求、应用范围、泄漏监测、裂纹分析及局部效应、全局效应等方面，对BP和LBB管道设计中的异同性进行了分析。分析结果表明：BP的“纵深防御”体系能包络LBB的主体内容；相对于LBB，BP对管道设计的要求更复杂；主冷却剂管道采用LBB和BP后的局部效应和全局效应大体相同。     

Extended applications of LBB piping evaluation diagrams for a new plant design

Plant specific data, such as pipe geometry, material properties and pipe loads, are required to apply leak-before-break (LBB) to a piping system. Thus, LBB evaluation cannot be done until piping design and routing are completed. A simple method for evaluating LBB for piping systems during design process considering the effects of nozzle and the change in local and global compliance in cracked piping system is developed in this paper. This method produces piping evaluation diagrams for intermediate pipe locations and pipe-nozzle interface locations which defines the LBB requirements to the piping designer for use during the design process and is independent of pipe routing. Piping evaluation diagrams can be used for the LBB evaluation for a new plant design.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

Leak-Before-Break analysis for Pickering ‘A’ Unit 1 and Unit 4 large diameter main steam line pipes

This paper presents a Leak-Before-Break (LBB) analysis of large diameter main steam line pipes (i.e. NPS 28″ and 30″) running from reactor building to main steam balance header in Pickering nuclear plant Unit 1 and Unit 4. Recent development in LBB technology summarized in U.S. Nuclear Regular Commission report NUREG/CR-6765 was adopted. Based on the tiered approach of LBB philosophy, this LBB analysis belongs to level 2 or level 3 LBB analysis. Detailed fracture tolerance analyses and leakage rate calculations were performed. EPFM (elastic plastic fracture mechanics) theory of J-integral, resistance curve versus ductile crack extension was adopted in carrying out all fracture tolerance analyses. Through-wall cracks in axial and circumferential directions on both straight pipes and elbows were postulated and analyzed. The loads applied on the postulated cracked pipes were obtained from detailed piping stress analysis under deadweight load, design pressure, thermal expansion, seismic design based earthquake (DBE) and thrust load due to the opening of relief valves. J-resistance data were derived from the lowest fracture toughness testing data obtained from Ontario Power Generation's PHT (primary heat transport) LBB material testing programs. A margin of 2 on crack size was chosen in establishing maximum allowable crack sizes. Leakage rates were calculated using SQUIRT Windows Version 1.1 program. The fluid inside the main steam line pipes was assumed single phase steam at 100% quality. One tenth of the calculated leakage rates was proposed as the requirement for minimum leakage detection capability. The paper concludes that the absence of through-wall crack larger than 91.16 mm in length should be maintained in order to ensure the structural integrity of large diameter main steam line pipes. In lieu of this crack size requirement, a reliable leakage detection capability which could quantify mass steam leakage rate of 0.01678 kg per second, or volume leakage rate of 1.01 l/min, should be in place. If both of the above two requirements are met, the Leak-Before-Break of these large diameter main steam line pipes is warranted.     

Technical note Leak detection in nuclear piping outside containment

The results of an experimental investigation performed at Wyle Laboratories to evaluate various methods for detecting small leaks in high energy piping system is described. These experiments were designed to support the leak-before-break methodology currently being employed by the United States nuclear industry. This methodology requires that:

1. (1) the flow rate through a hypothetical leak be accurately predicted, and

2. (2) the lower limits of detecting the flow rate through such a leak be established.

The research described in this work was designed to establish experimentally this limit of detectability.The experiments performed covered a range of leakage flow rates between 0.1 and 5 gpm (378–18925 cm³/min through small penetrations in a 6-in. diameter carbon steel pipe. Insulation, typical of the types found in nuclear power plants, covered the test pipe.The key observation made is that leak rates down to 0.1 gpm (378 cm³/min) are easily detectable.     

LBB管道裂纹泄漏率计算软件PICLES验证研究

管道裂纹泄漏率计算是破前漏（LBB）分析中的关键技术，采用与有效软件进行对比和与实验结果进行对比的方式，对国内自主研发的泄漏率计算软件PICLES进行验证研究。与已有成熟工程应用的国际同类软件（PICEP和SI-PICEP）对比，PICLES与其计算结果相差较小；与管道裂纹泄漏率实验结果对比，PICLES计算出的泄漏率与其相差?80.23%~?43.79%，PICLES计算的泄漏裂纹长度与实测裂纹长度相差21.84%~79.07%，说明将PICLES用于过冷水管道LBB分析具有较高的保守性。因此，PICLES可用于实际工程中的LBB分析。      

Low cycle fatigue and ductile fracture for Japanese carbon steel piping under dynamic loadings

Dynamic fracture behavior of circumferentially cracked pipe is important to evaluate the structural integrity of nuclear piping from the viewpoint of the LBB concept under seismic conditions. Fracture tests have been conducted for Japanese carbon steel (STS410) circumferentially through-wall cracked pipes that are subjected to monotonic or cyclic bending loads at room temperature. In the monotonic-loading tests, the maximum load to failure increases slightly with increasing loading rate. The failure cycles can be expressed simply by ratio of the load amplitude to the plastic collapse load. Fracture analysis has been also conducted to model the pipe tests. A new equation for calculating ΔJ for a circumferentially through-wall cracked pipe subjected to bending has been proposed. The failure cycles under cyclic loads are satisfactorily evaluated using an elastic-plastic fracture mechanics parameter ΔJ.