Primary Piping Research Articles

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.

Correction Techniques of Radiation Induced Errors for Raman Distributed Temperature Sensor and Experiment at the Experimental Fast Reactor: JOYO

Optical fibers have advantages like flexible configuration, intrinsic immunity for electromagnetic fields etc., and they have been used as optical fiber sensors. By some of these techniques, continuous or discrete distribution of physical parameters can be measured. Here, in order to apply Raman distributed temperature sensor (RDTS) to the monitoring of nuclear facilities: some correction techniques for radiation induced errors were investigated. It has been shown that, when uniform loss distribution can be assumed, simple correction technique with two thermocouples can be applied. Moreover, if loop type arrangement is applied, even when the loss distribution is not uniform, radiation induced errors can be canceled.For the demonstration of the feasibility of this technique, measurements using a commercial RDTS sys-tem were carried out along the primary piping system of the experimental fast reactor: JOYO. During the continuous measurements with the total dose of more than 107R, the radiation induced errors showed a saturating tendency. The correction technique with two thermocouples was applied and its feasibility has been demonstrated. Although the time response of the system should be improved, the RDTS can be expected as a noble temperature monitor in nuclear facilities.

A Study of Effects of Pipe Geometry on FAD Curves for Austenitic Stainless Steel and Ferritic Steel Piping Materials

A comprehensive study of failure assessment diagrams for circumferentially surface-cracked austenitic stainless and ferritic steel pipes was conducted with the use of the finite element method (FEM). While the majority of the analyses were conducted using the line-spring/shell finite element method, some three-dimensional finite element analyses, conducted independently, are also reported in this paper. Comparison of the predictions of the line-spring/shell and three-dimensional analyses reinforce the validity of the former approach for surface-cracked pipes. The results indicated that the ASME Code Case N-494-2 applicable for ferritic steel piping appears reasonably conservative even for pipes with mean radius-to-wall thickness ratios of 20, whereas the results showed that the newly adopted Code Case N-494-3 for austenitic stainless steel piping requires a limit for pipe with mean radius-to-wall thickness ratios larger than 15. For consistency, the limitation of Rm/t ≤ 15 was incorporated in the approved final version of Code Case N-494-3, and was incorporated in Code Case N-494-2 as well. Because these Code cases are applicable only to Class 1 primary nuclear piping, which typically has values of Rm/t ≤ 15, this is not a significant limitation. It was also shown that the choice of definitions of membrane and bending stresses as well as the choice of F1 function values in calculating the elastic part of the J integral have a profound effect on the resulting FAD curves.

Numerical diffusion and the tracking of solute fields in system codes: Part III. Application to a boron dilution transient analysis in the AP600

A study of a pump restart scenario in the AP600 with an unborated coolant plug in two of the four cold legs is presented. It has been performed with TRAC-PF1/MOD2 coupled with a 3-dimensional core neutronics model based on the nodal expansion method (NEM), and high order boron tracking algorithms. These are based on ULTIMATE-QUICKEST for 1-dimensional components and a flux corrected method developed by Smolarckievicz in the 3-dimensional vessel in order to reduce the numerical diffusion inherent to the upwind method used by most system codes to solve the transport equations. No turbulent diffusion model was included in the calculation to produce more conservative results. The results show that reduction of the numerical diffusion yields predictions with a significantly reduced margin in the size of the unborated plugs allowed to form in the primary side piping. In addition, two pump restart strategies have been suggested by the results, which could substantially decrease the size of an unborated plug injected into the core, in case it was suspected to have formed in a primary loop.

Vibration Analysis of Primary Inlet Pipeline of Pakistan Research Reactor-l during Steady State and Transient Conditions.

Pakistan Research Reactor-1 (PARR-1) has been converted to low enriched uranium (LEU) fuel and upgraded to a maximum power level of 9MW. During conversion and upgradation most of the reactor systems were modified and several additional facilities were provided. Major changes were carried out in the cooling system. These include installation of new primary pumps, set of heat exchanger assemblies, a cooling tower and piping system. Vibration tests on these systems were performed to check the integrity and detect any abnormality in normal operation. Results of the tests performed on the primary pumps and core support structure comply with the requirements for pre-operational and initial startup vibration testing of nuclear power plant (NPP) system. Attempt is now being made to analyze vibrations of primary piping. The Primary Water Inlet Pipeline (PW-IPL) is of stainless steel conveying demineralized water from holdup tank to the reactor pool. The section of the pipeline from heat exchangers to the valve pit...

Correction Techniques of Radiation Induced Errors for Raman Distributed Temperature Sensor and Experiment at the Experimental Fast Reactor: JOYO.

Optical fibers have advantages like flexible configuration, intrinsic immunity for electromagnetic fields etc., and they have been used as optical fiber sensors. By some of these techniques, continuous or discrete distribution of physical parameters can be measured. Here, in order to apply Raman distributed temperature sensor (RDTS) to the monitoring of nuclear facilities: some correction techniques for radiation induced errors were investigated. It has been shown that, when uniform loss distribution can be assumed, simple correction technique with two thermocouples can be applied. Moreover, if loop type arrangement is applied, even when the loss distribution is not uniform, radiation induced errors can be canceled. For the demonstration of the feasibility of this technique, measurements using a commercial RDTS sys-tem were carried out along the primary piping system of the experimental fast reactor: JOYO. During the continuous measurements with the total dose of more than 107R, the radiation induced erro...

Application of Leak-Before-Break Concept in VVER 440-230: Results and consequences for TACIS-program 91/1.2

The project TACIS 91/1.2 `Primary Circuit Integrity, Application of Leak-Before-Break (LBB) Concept' was performed within the scope of the TACIS-program (Technical Assistance for the Community of Independent States) of the European Community. The application of the LBB concept was carried out for the primary piping (main coolant lines (MCL) and surgelines (SL)) of the first generation of the Russian VVER NPP's of the type VVER 440/230 of the nuclear power plants Kola 1 and 2 (K1, K2) and Novovoronesh 3 and 4 (NV3, NV4). The work was performed by the Consultant (Siemens, EdF, Framatome) and Russian subcontractors (headed by Vniiaes). The paper describes the comparison of the relevant safety concept, the scope of the LBB-program, the applied LBB-procedures (German and NUREG 1061 procedure), the results obtained and the recommendations (analysis of seismicity, if necessary LBB analysis taking into account additional optimized seismic supports and their installation for NV3 and 4 adaption of inservice inspection (ISI) and leak monitoring systems (LMS)). Proposals have been made for future work to improve the safety of these plants. This project was financed by the European Community's TACIS Programme, which provides grant finance for know-how to foster the development of marked economies and democratic societies in the New Independent States and Mongolia.

Demonstration of piping integrity with SMA technology

The safe function of a new pipe whip restraint device has been demonstrated in a full scale test. The restraint is based on using a shape memory alloy to protect a pipe and its environment in the event of a double-ended-guillotine-break. The evaluation test has been performed at boiling water reactor (BWR) operating pressure and temperature using a pipe representing BWR primary piping.

A study on probabilistic fracture mechanics for nuclear pressure vessels and piping

This paper describes some recent research activities on probabilistic fracture mechanics (PFM) for nuclear pressure vessels and piping (PV&P) performed by the RC111 research committee of the Japan Society of Mechanical Engineers (JSME) under a subcontract of the Japan Atomic Energy Research Institute (JAERI). To establish standard procedures for evaluating failure probabilities of nuclear PV&P, we have set up the following three kinds of PFM round-robin problems on: (a) primary piping under normal operating conditions, (b) aged reactor pressure vessel (RPV) under normal and upset operating conditions, and (c) aged RPV under pressurised thermal shock (PTS) events. The basic problems of the last one are chosen from some US benchmark problems such as EPRI (Electric Power Research Institute) and US NRC (Nuclear Regulatory Commission) joint PTS benchmark problems. For the above round-robin problems, various sensitivity analyses are performed to quantify effects of uncertainty of data on failure probabilities. This paper summarizes some sensitivity studies on the above three kinds of problems mainly varying material properties such as flow stress, fracture toughness, fatigue crack growth rate, Cu content. Employed in this study are the PFM computer codes developed in Japan and USA. Failure probabilities of nuclear PV&P are quantitatively discussed in detail.

Flaw evaluation of theramally aged cast stainless steel in light-water reactor applications

Cast stainless steel may be used in the fabrication of the primary loop piping, fittings, valve bodies, and pump casings in light-water reactors. However, this material is subject to embrittlement due to thermal aging at the reactor temperature, that is, 290°C (550°F). The Argonne National Laboratory (ANL) recently completed a research program and the results indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs). Thus, the US Nuclear Regulatory Commission (NRC) staff has accepted the use of SAW flaw evaluation procedures in IWB-3640 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to evaluate flaws in thermally aged cast stainless steel for a license renewal evaluation. Alternatively, utilities may estimate component-specific fracture toughness of thermally aged cast stainless steel using procedures developed at ANL for a case-by-case flaw evaluation.

Shell response of core barrel for tributary pipe break

Work on fracture mechanics has provided a technical basis for elimination of main coolant loop double-ended guillotine breaks from the structural design basis of a reactor coolant system. Without main coolant loop pipe breaks, the tributary pipe breaks must be considered as design bases until further fracture mechanics work can eliminate some of these breaks from design consideration. This paper determines the core barrel shell responses for the 3 inch pressurizer spray line nozzle break which is expected to be the only inlet break remaining in the primary side after leak-before-break evaluation is extended to smaller pipes in the near future. The responses are compared with those due to a 14 inch safety injection nozzle break and main coolant loop pipe break. The results show that, when the leak-before-break concept is applied to the primary side piping systems with a diameter of 10 inches or over, the core barrel shell responses due to pipe breaks in the primary side are negligible for the faulted condition design.

Residual stresses in austenitic stainless steel primary coolant pipes and welds of pressurized water reactors

Surface and through thickness residual stress measurements were performed on an aged cast austenitic-ferritic stainless steel pipe and on an orbital TIG weld representative of those of primary coolant pipes in pressurized water reactors. An abrasive-jet hole drilling method and a block removal and layering method were used. Surface stresses and through thickness stress profiles are strongly dependent upon heat treatments, machining and welding operations. In the aged cast stainless steel pipe, stresses ranged between −250 and +175 MPa. On and near the orbital TIG weld, the outside surface of the weld was in tension both in the axial and hoop directions, with maximum values reaching 420 MPa in the weld. On the inside surface, the hoop stresses were compressive, reaching −300 MPa. However, the stresses in the axial direction at the root of the weld were tensile within 4 mm depth from the inside surface, locally reaching 280 MPa.

The advanced neutron source: Designing for science

The Advanced Neutron Source (ANS) is a new experimental facility planned to meet the need for an intense steady-state source of neutrons. The facility will be build around a new research reactor with a fission power of about 330 MW, producing an unprecedented flux that will provide the most intense beams of steady-state neutrons in the world. This paper outlines the overall scientific mission and technical objectives of this project, defines the regulatory approach that is likely to be applied, and presents a summary of the facility layout, the major plant systems, and the reactor and experiment systems. Detailed discussions are then given related to the design of the reactor core and various options evaluated, and the approach taken on safety and severe accident analyses. A summary is also included on the topic of environmental compliance and several key aspects that must be addressed to ensure compliance in that area. A brief introduction here serves to define the fundamental aspects of the ANS facility. The project is planned around a four majore building array with the central structure a 60-m-diam cylindrical, domed reactor containment building. This building houses the reactor itself, with the first floor dedicated to beam and irradiation experiments, the second floor divided between experimental facilities and reactor operations, and a high bay top floor dedicated to reactor operations. The other structures surround this central complex and house the supporting systems as well as other experimental facilities. The reactor core is comprised of two right circular cylindrical elements that are coaxially aligned, separated on the vertical axis, and offset in radius such that unheated coolant enters each element. The core has a volume of approximately 67 L and is positioned in a replaceable core pressure boundary tube that constitutes a section of the primary coolant piping. Surrounding the core pressure boundary tube is a reflector tank holding heavy water, approximately 3-m in diameter, which serves as a neutron reflector and moderator for the reactor and experiment systems. Materials irradiation and transuranic production tartets are located inside the core pressure boundary tube near the core while beam tubes and other irradiation facilities are located in the reflector tank. For further moderation of neutrons to very low energies, the reflector tank will contain two cold sources. Each cold source is a helium-cooled cryostat containing a liquid deuterium moderator and serves as a source of cold neutrons.

Properties of cast austenitic materials for light water reactors

Primary piping elbows of ligh water reactors are made of cast austeno-ferritic materials (grade CF8 or CF8M) which are susceptible to embrittlement by thermal ageing in service. Ageing kinetics depends on chromium content and on the temperature. The effect of ageing on mechanical properties is determined by the decreasing of Charpy-U energy. It is shown that the rupture of such aged material comes from a combination of cleavage of the ferrite and shearing or ductile tearing of austenite ligaments. The fracture toughness is determined by J-R curves, which exhibit a large scatter related to the solidification macrostructure of the steel. It is shown that J-R curves are independent of the geometry of the specimens.

高温ガス炉の１次冷却系主配管破断事故時の空気浸入挙動，（ＩＩ） 炉心の黒鉛流路管温度が非均一の場合

A primary pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas Cooled Reactor (HTGR). When the primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. During the accident a complicated multi-component gas flow is expected to occur in the reactor vessel. This paper deals with experimental study on the process of air ingress into an apparatus simulating the HTGR during the 1st stage of the accident. Present study focuses on investigating the density change of the gas mixture, concentration change of each gas species, and duration of the 1st stage of the accident when the temperature of the central flow path is higher than that of the peripheral ones and when the temperature of the reactor simulator decreases with time after the pipe rupture accident. It was found that the cooling rate of the reactor core temperature for preventing natural circulation of air existed. The results of the density and concentration changes in the non-uniform temperature case are almost the same as those in the uniform temperature case if natural circulation does not occur in the reactor core flow paths.

Smoking Cessation and the Risk of Stroke in Middle-aged Men

To examine the effects of giving up smoking, years since quitting smoking and the quantity of cigarettes smoked, and primary pipe or cigar smoking on the risk of stroke. A prospective study of cardiovascular disease and its risk factors in 7735 men aged 40 through 59 years drawn at random from the age-sex registers of one general practice in each of 24 British towns from 1978 through 1980 (the British Regional Heart Study). Incidence of fatal and nonfatal major stroke events (strokes) during an average follow-up period of 12.75 years. During the 12.75 years of follow-up, there were 167 major stroke events (43 fatal and 124 nonfatal) in the 7264 men with no recall of previous ischemic heart disease or stroke. After full adjustment for other risk factors, current smokers had a nearly fourfold relative risk (RR) of stroke compared with never smokers (RR, 3.7; 95% confidence interval [CI], 2.0 to 6.9). Ex-cigarette smokers showed lower risk than current smokers but showed excess risk compared with never smokers (RR, 1.7; 95% CI, 0.9 to 3.3; P = .11); those who switched to pipe or cigar smoking showed a significantly increased risk (RR, 3.3; 95% CI, 1.6 to 7.1) similar to that of current light smokers. Primary pipe or cigar smokers also showed increased risk (RR, 2.2; 95% CI, 0.6 to 8.0), but the number of subjects involved was small. The benefit of giving up smoking completely was seen within 5 years of quitting, with no further consistent decline in risk thereafter, but this was dependent on the amount of tobacco smoked. Light smokers (< 20 cigarettes/d) reverted to the risk level of those who had never smoked. Heavy smokers retained a more than twofold risk compared with never smokers (RR, 2.2; 95% CI, 1.1 to 4.3). The age-adjusted RR of stroke in those who quit smoking during the first 5 years of follow-up (recent quitters) was reduced compared with continuing smokers (RR, 1.8; 95% CI, 0.7 to 4.6 vs RR, 4.3; 95% CI, 2.1 to 8.8). The benefit of quitting smoking was observed in both normotensive and hypertensive men, but the absolute benefit was greater in hypertensive subjects. Smoking cessation is associated with a considerable and rapid benefit in decreasing the risk of stroke, particularly in light smokers (< 20 cigarettes/d); a complete loss of risk is not seen in heavy smokers. Switching to pipe or cigar smoking confers little benefit, emphasizing the need for complete cessation of smoking. The absolute benefit of quitting smoking on risk of stroke is most marked in hypertensive subjects.

Smoking cessation and the risk of stroke in middle-aged men

<h3>Objective.</h3> —To examine the effects of giving up smoking, years since quitting smoking and the quantity of cigarettes smoked, and primary pipe or cigar smoking on the risk of stroke. <h3>Design, Subjects, and Setting.</h3> —A prospective study of cardiovascular disease and its risk factors in 7735 men aged 40 through 59 years drawn at random from the age-sex registers of one general practice in each of 24 British towns from 1978 through 1980 (the British Regional Heart Study). <h3>Main Outcome Measure.</h3> —Incidence of fatal and nonfatal major stroke events (strokes) during an average follow-up period of 12.75 years. <h3>Results.</h3> —During the 12.75 years of follow-up, there were 167 major stroke events (43 fatal and 124 nonfatal) in the 7264 men with no recall of previous ischemic heart disease or stroke. After full adjustment for other risk factors, current smokers had a nearly fourfold relative risk (RR) of stroke compared with never smokers (RR, 3.7; 95% confidence interval [CI], 2.0 to 6.9). Ex—cigarette smokers showed lower risk than current smokers but showed excess risk compared with never smokers (RR, 1.7; 95% CI, 0.9 to 3.3;<i>P</i>=.11); those who switched to pipe or cigar smoking showed a significantly increased risk (RR, 3.3; 95% CI, 1.6 to 7.1) similar to that of current light smokers. Primary pipe or cigar smokers also showed increased risk (RR, 2.2; 95% CI, 0.6 to 8.0), but the number of subjects involved was small. The benefit of giving up smoking completely was seen within 5 years of quitting, with no further consistent decline in risk thereafter, but this was dependent on the amount of tobacco smoked. Light smokers (&lt;20 cigarettes/d) reverted to the risk level of those who had never smoked. Heavy smokers retained a more than twofold risk compared with never smokers (RR, 2.2; 95% CI, 1.1 to 4.3). The age-adjusted RR of stroke in those who quit smoking during the first 5 years of follow-up (recent quitters) was reduced compared with continuing smokers (RR, 1.8; 95% CI, 0.7 to 4.6 vs RR, 4.3; 95% CI, 2.1 to 8.8). The benefit of quitting smoking was observed in both normotensive and hypertensive men, but the absolute benefit was greater in hypertensive subjects. <h3>Conclusion.</h3> —Smoking cessation is associated with a considerable and rapid benefit in decreasing the risk of stroke, particularly in light smokers (&lt;20 cigarettes/d); a complete loss of risk is not seen in heavy smokers. Switching to pipe or cigar smoking confers little benefit, emphasizing the need for complete cessation of smoking. The absolute benefit of quitting smoking on risk of stroke is most marked in hypertensive subjects. (<i>JAMA</i>. 1995;274:155-160)

Lifetime evaluation of Bohunice VI NPP primary loop components

Safety and reliability measures are based on fatigue damage accumulation analyses using the actual load history such that life-time estimates could be made for the critical components. This required a knowledge of the corresponding stresses and deformations. Finite element computations have been made for analyzing the operational conditions of the primary loop components including pipings of WWER (moderately cooled water reactor) V-230 reactor in 440 Nuclear Power Plant (NPP). The results showed that damage due to fatigue in all of the components at the critical locations is low and is not of a concern up to now. This is the basis of support for operating the Bohunice VI NPP primary piping components where the program makes up of the LBB (leak-before-bread) concept.

Pipe break testing of primary loop piping similar to the Department of Energy's new production reactor-heavy water reactor

This paper provides information about and results from specific tests completed at Oak Ridge National Laboratory (ORNL) in support of the Department of Energy's (DOE's) new production reactor-heavy water reactor (NPR-HWR) program. The paper also provides detailed analytical studies completed by Battelle Columbus Laboratory. All of this information is presented in a demonstration that the primary piping of the NPR-HWR, with its relatively moderate temperature and pressure should not suffer an instantaneous double-ended guillotine break (DEGB) under design basis loadings and conditions. The first ORNL test series obtained pipe crack extension data representative of large dynamic loads applied to a flawed stainless steel 316LN pipe weldment. The results of that testing were reviewed by a special Piping Integrity Review Group (PIRG) established by DOE. Members of this group are identified in this paper. The PIRG review comments were such as to necessitate a second series of tests. This paper provides results of that second series of pipe tests. An aged section of 304 stainless steel pipe, obtained from the C-reactor at Savannah River, was subjected to large fully reversed cyclic bending loads. An initial flaw size was based on leakage flow testing at prototypic NPR-HWR pressure and temperature. The detectable leakage flaw was cycled in bending at large loads for 40 cycles. The members of the PIRG assessed these results together with analytic work by Battelle and stated that, provided the caveats cited in this paper are in force, the following conclusion is applicable: • For DOE low-pressure (≤1·72 MPa), low-temperature (100°C) reactors with austenitic stainless steel piping, the DEGB should not be a design basis condition.

高温ガス炉の１次冷却系主配管破断事故時の空気浸入挙動 （Ｉ） 炉心の黒鉛流路管温度が均一の場合

A primary pipe rupture accident is one of the design-basis accidents of the High-Temperature Engineering Test Reactor (HTTR), which is now under construction at Japan Atomic Energy Research Institute. When the primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. During the accident a complicated multi-component gas flow is expected to occur in the reactor vessel. This report deals with experimental study on air ingress process in the 1st stage of the accident. The experiments are carried out with a test model simulating the HTGR. Temperature of the simulated reactor core is varied from about 400-1, 050°C.