Demonstrating leak-before-break for CANDU heat transport piping

The leak-before-break approach is being used for Ontario Hydro's Darlington nuclear generating station as an alternative to the provision of pipe-whip restraints on large diameter primary heat transport system piping. These technological developments have been used by Ontario Hydro as the nucleus of an approach for demonstrating that CANDU Class 1 carbon steel piping will not break catastrophically; at worst it would leak at a detectable rate and corrective action would be taken well before catastrophic rupture could occur. The present paper describes how the need for a leak-before-break approach evolved, considerations given to developing an approach, and how it is being applied. Finally, the paper updates the status of this program and discusses future plans in this area.     

Ontario hydro's leak-before-break approach to Darlington NGS heat transport system piping

The primary objective in our Leak-Before-Break studies is to show how a rational and comprehensive approach can provide an adequate measure of confidence in the assessment of piping integrity such that provision of design features (viz. pipewhip restraints, jet impingement shields) to protect against the dynamic effects of pipe rupture is not necessary. This study is one component of the overall Leak-Before-Break approach adopted at Ontario Hydro.

The results of a review undertaken to evaluate the system transients or events sequences which may subject the piping to a potentially significant increase in loadings are reported. The focus in this paper is to show the approach used in deriving loadings for use in the elastic-plastic fracture mechanics (EPFM) analyses required to demonstrate crack stability.     

The J-resistance curve leak-before-break test program on material for the darlington nuclear generating station

The Darlington Leak-Before-Break (DLBB) approach has been developed for large diameter (21, 22, 24 inch) SA106B heat transport (HT) piping and SA105 fittings as a design alternative to pipewhip restraints and in recognition of the questionable benefits of providing such restraints. Ontario Hydro's DLBB approach is based on the elastic plastic fracture mechanics (EPFM) method. In this test program, J-resistance curves were determined from actual pipe heats that were used in the construction of the Darlington heat transport systems (Units 1 and 2). Test blocks were prepared using four different welding procedures for nuclear Class I piping.

The test program was designed to take into account the effect of various factors such as test temperature, crack plane orientation, welding effects, etc., which have influence on fracture properties. A total of 91 tests were conducted. An acceptable lower bound J-resistance curve for the piping steels was obtained by machining maximum thickness specimens from the pipes and by testing side grooved compact tension specimens.

Test results showed that all pipes, welds and heat-affected zone materials within the scope of the DLBB program exhibited uppershelf toughness behaviour. All specimens showed high crack initiation toughness J_Ic, rising J-resistance curve and stable and ductile crack extension. Toughness of product forms depended on the direction of crack extension (circumferential versus axial crack orientation). Toughness of DLBB welds and parent materials at 250°C was lower than that at 20°C.     

Ductile fracture properties for assessing leak-before-break issues in ferritic weldments

A Leak-Before-Break (LBB) approach is being used by Ontario Hydro's Darlington nuclear generating station as a design alternative to pipe rupture restraint hardware on the large diameter piping of the primary heat transport system. The J-resistance curves of four different ferritic weldments, fabricated by either the submerged arc weld (SAW) or shielded metal arc weld (SMAW) process, were determined as part of this program.

Results indicated that the as-welded and post-weld heat treated (PWHT) welds were susceptible to varying degrees of static or dynamic strain aging at 200 and 250°C. Dynamic strain aging effects were most significant for as-welded welds, as evidenced by sudden load drops on the load-displacement curves and ductile crack jumping. The effect of loading displacement rate and PWHT on toughness was assessed and related to the weld's tensile properties and susceptibility to dynamic strain aging. Implications of strain aging to LBB assessments are discussed.     

Analysis of the failure performance of internally pressurised piping with surface flaws

Because of frequent failures of the Atucha PWR moderator circuit branch piping, which is made of stainless steel type AISI 347 (DIN 1.4550), studies have been made, involving the application of several fracture mechanics criteria, in order to determine the conditions of leak-before-break (LBB) and the critical crack length of the piping.These studies lead to the conclusion that, for straight pipe of outside diameter 219 mm and 16 mm wall thickness, with a circumferential flaw and with the principal stress being in bending, the LBB criteria are satisfied, the critical crack length being of the order of 400 mm.A better mechanical finishing and heat treatment was suggested in order to improve the resistance to crack initiation.     

Leak-before-break application in US light water reactor balance-of-plant piping

This paper describes the criteria and methodology for a leak-before-break (LBB) program for high energy balance-of-plant (BOP) nuclear piping in the United States. LBB, the analytical demonstration that high toughness piping will leak detectably before catastrophic failure, can be applied to any operational or pre-operational light water reactor plant to minimize pipe rupture hardware and to discount pipe rupture dynamic effects.

The general methodology described herein, encompasses applicable US NRC regulatory requirements and incorporates experience gained in the licensing process of actual LBB programs. First, candidate piping systems must be carefully screened to verify that they are not subject to failure by phenomena that would adversely affect the accurate evaluation of flaws. Next, pipe stresses, material properties, and leak detection capabilities are gathered for the fracture mechanics and fluid mechanics analyses. At the piping locations which have the least favorable combination of material properties and stress, a crack is postulated which is of sufficient size that the resulting leakage will be detected by installed leak detection systems. Finally, LBB is demonstrated if the postulated crack remains stable even if a seismic event takes place before the crack is discovered and repaired. An LBB example is presented in this paper for a generic pressurizer surge line, and reflects the consideration of flow stratification on LBB analyses.     

A sensitivity study in probabilistic fracture mechanics analysis of light water reactor carbon steel pipe

Through the efforts in leak-before-break research for light water reactor pipings based on deterministic fracture mechanics analysis, simple models which evaluate pipe fracture behavior are being established. Using these models it is also becoming possible to apply probabilistic fracture mechanics analysis. This paper describes an example of such an analysis, using these proposed models.

Since the authors' interests are in the range of uncertainty of the calculated failure probability, the effects of changes in the input parameters or the analytical conditions are also estimated by a sensitivity analysis. The results show that the calculated failure probability may be influenced significantly by changes in parameters concerning initial crack size distributions, and that effects due to a change in the leak detection model may appear after long operation of the plant.     

Fracture behaviour of large scale pressure vessels in the hydrotest

Within the Nordic Countries a four-year research programme in the area of elastic-plastic fracture mechanics was initiated in 1985. This programme aims to assess the leak-before-break (LBB) criteria for pressure vessels and piping. The main experimental effort of the programme is to rupture large size pressure vessels, one having dimensions resembling those of a nuclear reactor pressure vessel, under internal pressure. Artificial flaws were made on the inner wall of the vessels. The dimensions of the flaws were defined by calculations so that the LBB condition was just anticipated during the test. For the time being two tests have been performed. The first test with a large pressure vessel was pressurized by water at 60°C, which was the lowest acceptable temperature for the hydrotest. In this paper experimental details including flaw preparation, instrumentation and material characterization are described. The fracture behaviour as well as experimental results of the tests are reported and compared to the analytical solutions of the analyses.     

Leak-before-break: An integrated approach for high energy piping

An integrated approach that involves system design, thermal hydraulics, materials, and fracture mechanics analyses to assure that pipe failure is highly unlikely is described. This approach is based on a leak-before-break (LBB) premise and includes through-wall flaw detectable leakage, through-wall flaw stability, and part-through-wall flaw fatigue crack propagation calculations. A successful application of LBB can reduce the amount of excessive pipe rupture restraint hardware. Assuring LBB not only reduces initial construction, future maintenance, and radiation exposure costs, but the overall safety and integrity of the plant are improved. This last benefit comes about from gaining additional insight into the piping systems and their capabilities. Details of the LBB methodology are presented here with specific examples for two pressurized water reactor lines (one inside containment fabricated of stainless steel, and the other outside containment made from ferritic steel). The application of this approach at Beaver Valley Power Station—Unit 2 indicates that pipe rupture hardware is not necessary for stainless steel lines inside containment as small as 6-in (152 mm) nominal pipe size that have passed screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 3-in (76 mm) diameter (outside containment) can qualify for pipe rupture hardware elimination.     

A computer model for probabilistic leak-rate analysis of nuclear piping and piping welds

This paper describes the development of a computer code entitled PSQUIRT for probabilistic evaluations of leak rate in nuclear piping. It is based on (1) the Henry-Fauske model of two-phase flow for thermal-hydraulic analysis and (2) an estimation model for elastic-plastic fracture-mechanics analysis. In both analyses, uncertainties arise due to the incomplete knowledge of the crackmorphology variables and statistical scatter of the pipe material properties. The relevant parameters required to conduct these analyses were modeled as random variables. Henceforth, the above thermal-hydraulic and fracture-mechanics models were put in a probabilistic format to allow statistical variability of input and determination of their effects in pipe fracture and leak-before-break (LBB) evaluations. A standard Monte Carlo simulation technique was used to perform the probabilistic analysis. Numerical examples are presented to illustrate the capabilities of the PSQUIRT code. Probabilistic analyses were performed by PSQUIRT for a stainless steel and a carbon steel pipe. Histograms were developed for leakage rate and flaw size in these pipes for LBB applications. The results suggest that the variability of leak rate can be significant due to statistical scatter of crack-morphology parameters. Using these histograms, the subsequent fracture stability of a leaking crack, actural or hypothetical, can be evaluated by either a deterministic or a probabilistic method.     

核电站高中能管道断管影响探讨

在设计核电站时,须对安全壳内外的高能与中能管道进行发生假想性断管事件及其后续效应的设计分析。这些管道一旦发生破裂,泄漏的高能量流体将对管道施加很大的横向力,使管道产生高速运动,即管道甩动。这种高速运动的管道可能会对周围结构造成严重破坏,因而引起世界上各主要发展核电国家的重视,并开展了大量的研究工作。详细叙述常规岛侧高中能管道断管位置的判定准则、假想破口的类型、断管的后续影响及其防护等。还进一步介绍有关管道断裂甩动问题的各种计算方法,如力矩平衡法、能量平衡法、有限元法等。     

The difference between the fracture initiation and maximum load net-section stresses for a cracked structure

Experimental results for 304 stainless steel centre-cracked panels containing through-thickness cracks have been interpreted as indicating critical net-section stresses both for the initiation of fracture and for maximum load attained after a limited amount of stable crack growth, the fractional difference between these stresses being $\text{～ 10\%}$. By applying results from recent developments in general yield fracture mechanics, this paper shows that this small difference depends on the ligament width. This dependency has important implications when the net-section stress approach is applied to the development of leak-before-break criteria for structures containing part-through cracks.     

Leak-before-break qualification of primary heat transport piping of 500 MWe Tarapur atomic power plant

The advent of Leak-Before-Break (LBB) concept has now replaced the traditional design basis event of the Double-Ended-Guillotine-Break (DEGB) to design the Primary Heat Transport (PHT) system piping of the Pressurised Heavy Water Reactor (PHWR) and Pressurised Water Reactor (PWR). This approach is being adopted to design the PHT system piping of 500 MWe Indian PHWR to be built at Tarapur (Tarapur Atomic Power Plant 3 and 4). The LBB concept basically demonstrates through fracture mechanics analysis that there is negligible chance of any catastrophic break of PHT pipes without prior indication of leakage. There are several steps in this work of LBB qualification, namely, evaluation of loads on the piping components, generation of tensile and fracture properties of PHT pipe base and weld material, determination of leakage size crack (LSC) and the elastic–plastic fracture mechanics (EPFM) and limit load analysis of the piping components with postulated LSC to evaluate the critical load at unstable ductile tearing and the limit load, respectively. The paper deals with the fracture analysis of the straight pipes and elbows of three pipe lines in the PHT system of TAPP 3 and 4. Three crack configurations are considered in the analysis. These are throughwall circumferential crack at the weld location of straight pipe and extrados of the elbow and throughwall axial crack at the elbow crown. In all the cases, necessary factor of safety with respect to the anticipated safe shutdown earthquake (SSE) load and LSC are shown to be more than the minimum required values for LBB qualification.     

Experience in the replacement of safety related piping in German boiling water reactors

In the German boiling water reactors (BWR) of the 69 series and their forerunner plant, high-strength low-alloy ferritic materials were used for a large number of pipings both inside and outside the pressure boundary (PB). The choice of this type of material led to comparatively thin-walled piping which, at that time, had been designed and manufactured in accordance with the codes and standards applying in the Federal Republic of Germany.

Due to material properties resulting from production in a conventional manner, design features which did not sufficiently meet the requirements for nondestructive testability, and defects caused during processing, mainly in the area of circumferential welds, the ferritic pipings inside the PB were replaced in the course of a plant upgrading by new piping designed and manufactured according to the basis safety concept.

The improvements and experience gained during backfitting of five German BWR plants are part of the German safety strategy and can be summarized as follows:

1. (1) Exclusion of large fractures on the basis of an optimized quality level for the piping.

2. (2) Elimination of need for the pipe whip restraints which existed in the former piping.

3. (3) Limited reduction of the former scope of inservice inspections, mainly as a consequence of improved weld quality and optimized weld performance.

4. (4) Reduction of personnel radiation exposure, e.g. by reduced number of welds and by manufacture of welds using automatic equipment, as well as by improved nondestructive testing.

5. (5) Availability values for backfitted BWR comparable to German PWR values.

The pipings made of stabilized austenitic materials, which are arranged inside and outside the containment of the BWR plants, were not replaced since their quality level has been proved to be sufficient even on the basis of the present standards.     

Leak-before-break considerations for LMFBR structures

Design and safety aspects of LMFBR's are discussed. The specific conditions of LMFBR's and in particular the pool-type reactor now favoured in Europe for commercial use are outlined. The leak-before-break criterion is evaluated with respect to the pool design and the safety concept. A procedure is proposed to demonstrate the leak-before-break behaviour by fracture mechanics analyses. The corresponding boundary conditions are discussed. The uncertainties which remain and the efforts to resolve them are indicated.     

Leak-before-break in French nuclear power plants

Practical applications of the leak-before-break concept at the present stage are quite limited in French nuclear power plants. However, discussions with safety authorities have included leak-before-break arguments for the different types of reactors: pressurized water reactors (PWRs), liquid metal fast-breeder reactors (LMFBRs) and gas graphite reactors (GGRs).

At present, the fracture mechanics part of the studies are complete for the following components:

• —pipes in GGRs;

• —primary and auxiliary lines in PWRs;

• —steam-generator tubes in PWRs;

• —pipes and main vessels in LMFBRs.

The different approaches are consistent but some specific problems have to be taken into account, depending on the plant, such as creep regime, thin shell components, in-service inspection or the issue of design safety.

A large research and development program, realized in different cooperative agreements (national or international), completes the general approaches. It comprises different topics, such as material properties, elastoplastic fracture mechanics, leak-area determination and leak-detection devices.

Although practical applications are limited at present, EDF, in conjunction with Framatome, Novatome and CEA, will define a complete validated methodology to be used on a variety of cases.     

Analytical studies of four-inch pipe whip tests under BWR LOCA conditions

The purpose of pipe rupture studies in JAERI is to perform model tests on pipe whip, restraint behavior, jet impingement and jet thrust force, and to establish a computational method for analyzing these phenomena. This report presents the analytical results of 4-inch pipe whip tests under BWR LOCA conditions.Dynamic response analyses were performed using the general-purpose finite element program ADINA. The test pipe was modelled by straight beam elements and the four restraints were modelled by a single truss element. The analytical results were compared with the experimental results. Impact time and maximum total restraint force showed good agreement with experimental results. On the other hand, pipe strain and pipe deflection could not be predicted so well. The reason for this is that the sliding of the restraint during the pipe whip movement cannot be considered in the analyses.     

Allowable sizes of axial flaws in pressurized pipes made of moderate toughness materials

Failure stresses for axially part-through flawed pipes made of moderately tough materials are predicted by several fracture mechanics. However, allowable flaw sizes using these fracture mechanics cannot be simply described because there are many effective parameters such as pipe diameter, wall thickness, material properties, etc. To establish codes and standards to evaluate flaws for piping of light water reactors, we determine unified allowable sizes for axial flaws in pipes subjected to internal pressure from J-integral based fracture mechanics. The allowable sizes are simply tabulated using a single parameter which consists of pipe geometry and material properties.     

Principles of the German approach to the analysis of flawed structures

The analysis of flawed structures in the linear elastic regime (LEFM) has been treated very extensively, especially in the framework of the 4th German nuclear programme. Investigations with large specimens have demonstrated the applicability of the LEFM for a variety of materials ranging from low to high toughness levels. In the elastic-plastic (EPFM) regime, however, especially the J-integral method has been used for quantifying the safety margin, because with a required value of upper shelf Charpy energy alone this quantification is not possible.

A correlation between the crack initiation value J_i and Charpy upper shelf energy was evaluated experimentally, justifying the generally upgraded upper shelf energy level adopted in the Code. Furthermore, this correlation can be employed when fracture toughness values are not available as is mostly the case for irradiated material from surveillance programmes.

To cope with the problems resulting from repeated transients in a complex component, incipient crack intitiation and propagation under cyclic thermal load have been investigated experimentally and theoretically on a RPV nozzle corner in the HDR plant and on a thick-walled (200 mm) hollow cylinder subjected to pressurized thermal shock (PTS). Since the OCA code covers only the linear elastic range, for this loading case elastic-plastic fracture mechanics calculations have been carried out by means of the Finite Element Method. The first test performed with high toughness material has shown good agreement with the J-integral approach. Additional validation tests are under way to demonstrate the lowest tolerable toughness level to withstand PTS without catastrophic failure.

With respect to critical flaw sizes in degraded piping and vessels, the ‘leak before break’ limit curve for different loading conditions has been established and experimentally validated using piping and model vessels of different sizes, crack configurations and toughness levels.

The still existing uncertainties in the detection, sizing and interpretation of signals from nondestructive examination are the background of NDE validation programmes for both medium size and full size reactor pressure vessels. Acoustic emission trials as well as extensive ultrasonic (US) examinations will be pursued in cooperation programmes, last but not least in the framework of PISC III.     

Limit loads and fracture mechanics parameters for thick-walled pipes

In this paper, information on plastic limit loads and both elastic and elastic-plastic fracture mechanics parameters is given for cracked thick-walled pipes with mean radius-to-thickness ratios ranging from two to five. It is found that existing limit load expressions for thin-walled pipes can be applied to thick-walled pipes, provided that they are normalized with respect to the corresponding un-cracked thick-walled pipe values. For elastic fracture mechanics parameters, FE values of the influence functions for the stress intensity factor and the crack opening displacement are tabulated. For elastic-plastic J, it is shown that existing reference stress based J estimates can be applied, provided that a proper limit load for thick-walled pipes is used.