Primary Piping Research Articles

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.

Fracture mechanics and full scale pipe break testing for the Department of Energy's new production reactor-heavy water reactor

Oak Ridge National Laboratory (ORNL) has completed a major task for the US Department of Energy (DOE) in the demonstration that the primary piping of the proposed new production reactor-heavy water reactor (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 growth of possible small pre-existing defects in the piping wall was estimated over a plant life of 60 years. This worst-case flaw was then evaluated using fracture mechanics methods. It was calculated that this worst-case flaw would increase in size by at least 14 times before pipe instability during a safe shutdown earthquake (SSE) would even begin to be possible. The approach to showing the improbability of an instantaneous DEGB for HWR primary piping required a major facility (pipe impact test facility, PITF) to apply all possible design loads, including an equivalent major earthquake (called the SSE earthquake). The facility was designed and built at ORNL in 6 months. The test article was 6.1 m long, 406 mm diameter, 13 mm thick pipe of stainless steel 316LN material that was fabricated to exacting standards and inspections following the nuclear industry standard practices. A flaw was machined and fatigued into the pipe at a tungsten inert gas (TIG) butt weld (ER316L weld wire) as an initial condition. The flaw-crack was sized to be beyond the worst-case flaw that HWR piping could see in 60 years of service—if all leak detection systems and if all crack inspection systems failed to notice the flaw's existence. Starting October 1991, the first test article was subjected to considerable overloadings. The pipe was impacted 104 times at levels equal to and well beyond the SSE loadings. In addition, over 560 000 fatigue cycles and numerous purposeful static overloads were applied in order to extend the flaw to establish the data necessary to confirm fracture mechanics theories, and more importantly, to demonstrate simply that instantaneous DEGB is highly improbable for the relatively moderate energy system.

Application of leak-before-break concept to integrity and safety of PWR primary piping with WWER 1000

The general methodology and criteria used for the leak-before-break (LBB) concept for the PWR primary circuit with WWER 1000 are described. The application of the LBB concept requires the following additional protections. First, the candidate system should be highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that they are not subjected to any disqualifying failure mechanism. Therefore the piping service experience and design bases are screened. Next, the analysis of leak-detection capabilities is performed. The postulated crack must be of sufficient size that the resulting leakage will be detected by the leak-detection systems installed. Finally, fracture-mechanics analysis is performed to demonstrate that the crack will not cause pipe rupture even if seismic loads are applied before the flawed section is discovered and repaired. The calculated and experimental results obtained to date are very encouraging and show the applicability of the LBB concept for the PWR primary circuit with WWER 1000.

LBB technology application to the primary piping system of the NPP V1 Jaslovke Bohunice: Zdarek, J., Pecinka, L. and Joch, J. Nucl. Eng. Des. (Oct. 1993) 144 (1), 69–76

LBB technology application to the primary piping system of the NPP V1 Jaslovke Bohunice: Zdarek, J., Pecinka, L. and Joch, J. Nucl. Eng. Des. (Oct. 1993) 144 (1), 69–76

Overview of reability test programme on primary coolant piping of light water reactors: Shibata, K., Isozaki, T., Ueda, S., Kurihara, R., Onizawa, K. and Kosaka, A. Nihon-Genshiryoku-Gakkai Shi (J. Atomic Energy Society of Japan) (Oct. 1993) 35(10), 923–939 (in Japanese)

Overview of reability test programme on primary coolant piping of light water reactors: Shibata, K., Isozaki, T., Ueda, S., Kurihara, R., Onizawa, K. and Kosaka, A. Nihon-Genshiryoku-Gakkai Shi (J. Atomic Energy Society of Japan) (Oct. 1993) 35(10), 923–939 (in Japanese)

Two-phase flow phenomena in full-scale reactor geometry

For the investigation of two-phase flow phenomena in full scale reactor geometry, a series of experiments were carried out at the Upper Plenum Test Facility UPTF, which represents the primary system of a 1300 MWe Pressurized Water Reactor with upper plenum, downcomer and primary main coolant pipes in 1:1 reactor scale. UPTF was the German contribution to the international 2D/3D project established by the Japan Atomic Energy Research Institute (JAERI), the Nuclear Regulatory Commission (USNRC) of the United States of America, and the Federal Ministry for Research and Technology (BMFT) of the Federal Republic of Germany. Large scale findings of the UPTF tests, related to two-phase flow phenomena in the downcomer, in the upper plenum, at the upper core tie plate, and in the main coolant pipes, will be discussed. The application of the UPTF test results for the validation of analytical models will be demonstrated.

LBB technology application to the primary piping system of the NPP V1 Jaslovske Bohunice

Due to several deficiences of the WWER Model 230 type reactor a leak before break demonstration of this reactor is of primary importance. The complex project for NPP V1 Jaslovske Bohunice includes a static and dynamic stress analysis of the primary piping, a fatique damage analysis, leak rate assessments and an analysis of the stability of the heavy components supports. The material database includes data on fracture mechanics, on assessment of corrosion properties, and on the influence of 100000 hr service exposure on base metal and welds including disimilar welds. The program was supported by large scale experiments on RPV safe-end, pressurizer safe-end, elbow welds with through-wall cracks and leak rate measurements. The results and applications are discussed.

Behaviour of nonlinear supports on PWR coolant system during a postulated LOCA: Part 2: Effects of stiffness variations and cavity pressurisation

Part 2 of this two part paper deals with the variations in stiffness values of the supports and whip restraints significantly affected by uncertainties in evaluating concrete embedment stiffnesses and the effect of cavity pressurisation in the case of a Reactor Pressure Vessel (RPV) inlet or outlet nozzle break. Part 1 of the paper discussed the methodology of the LOCA analysis of the Reactor Coolant Loop (RCL) and a sensitivity study on the method of modelling the columns supporting the Reactor Coolant Pump and the Steam Generator. They were part of a series of sensitivity analyses performed to evaluate the influence of variations in the variables utilised as input to the primary coolant pipe dynamic structural analyses. The stiffness variations cause variations in loads received at the supports or restraints, thus influencing the design and the degree of accuracy to which it is necessary to define the stiffnesses. Cavity pressurisation is asymmetric pressurisation of the cavity between the RPV external surface and the primary shield wall. This generate external upthrust and lateral forces on the RPV thus increasing RPV movements and RCL support loadings. Different blockage situations impeding fluid pressure relief routes through the primary shield wall have been postulated due to the spalling of thermal insulation material. The loadings on the RCL supports and restraints as a result of increased RPV movements were analysed.

Fracture Toughness of Aged Stainless Steel Primary Piping Welds Evaluated by Multiple and Single-Specimen Methods

Fracture toughness testing was conducted on compact tension specimens cut from the Fast Flux Test Facility primary piping materials of 16-8-2 and 308 stainless steel welds aged at 427 and 482°C for 20,000 and 50,000 hr. The ductile fracture behavior of the materials was characterized at 205, 427, and 482°C using multiple and electric-potential single-specimen techniques. Electric-potential data were used to caculate crack extensions via an electric-potential calibration equation for the construction of J-R curves. Results demonstrate that the critical fracture-toughness values are in good agreement with those from the multiple-specimen method. Results showed that 20,000-hr aging caused more than 35 percent degradation in fracture resistance, and 50,000-hr aging resulted in a slight increase in Jc for 16-8-2 stainless steel welds. It was found that the fracture toughness levels of the primary piping after long-term aging were high and adequate at the aging temperatures and that fuel handling temperature, nonductile fracture was not expected to occur in these materials.

Containment for the low temperature district nuclear-heating reactor

An integral arrangement is adopted for the Low Temperature District Nuclear-Heating Reactor. The primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with the reactor core. The primary coolant flows in natural circulation through the reactor core and the primary heat exchangers. The primary coolant pipes penetrating the wall of the reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of the pressure boundary of the primary coolant. Therefore a small sized metallic containment closed to the wall of the reactor vessel can be used for the reactor. Design principles and functions of the containment are the same as for the containment of a PWR. But the adoption of a small sized containment brings about some benefits such as a short period of manufacturing, relatively low cost, and ease for sealing. A loss of primary coolant accident would not be happened during a rupture accident of the primary coolant pressure boundary inside the containment owing to its intrinsic safety.

Primary Circuit Metal State With Regard to NPP Lifetime Extension

This article describes the assessment of the suitability for operation of soviet reactors. The VVER-440 (10 units), VVER-1000 (16 units) and RBMK-1000 (15 units) are the main types of soviet NPP reactors. Reactor design lifetime is 30 years. Primary circuit main equipment and piping have accordingly the same lifetime. The problem of lifetime extension arises in the case of design lifetime expiration or in the case of deviations from the reactor nominal characteristics. At present, only two reactors have expired their design lifetime and are decommissioned (1 unit of Beloyarsc NPP and 1 unit of Novovoronezh NPP). Deviations from the reactor nominal characteristics may appear for different reasons. The majority of them are as follows: Level of equipment and pipeline manufacture and assembly technology which at present does not provide for an absolutely defect-free metal state; level of flaw detection monitoring method development considering that defect detection is an accidental, probabilistic process; deviations from nominal operation conditions which, as a rule, are of an accidental nature; design deficiencies related as a rule to in essential equipment components: Deviation in their operation leads to an essential equipment component operational condition change; highly conservative level of quality assessment standards of operational flawmore » detection control stages are practically the same. Defects due to structural deficiencies occur rather rarely, but in the case of such defect occurrence at one unit, the same defects are to be expected at the other ones.« less

Long-term integrity of main pressure boundary components in the first generation of NPPs in Czechoslovakia

The determination of extremely low probability of the reactor pressure vessel and the primary main coolant pipe fracture is one of the prerequisites for further operation of the first generation NPPs in Czechoslovakia. The project for this task includes experimental and analytical work on the reactor pressure vessel and on the main piping components. Review of the project and preliminary results are presented. The results show extremely low probability of failure. Future work is focussed on annealing studies and tests for the pressure vessel and the application of the Leak-Before-Break technology on primary piping circuit.

Studies on diffusion and natural convection of two-component gases

A primary-pipe rupture accident is one of the design-based accidents of a high-temperature engineering test reactor (HTTR), which is being developed at JAERI. When the primary pipe ruptures, air is expected to enter into the reactor core from the breach by molecular diffusion and natural convection. In order to investigate the process of air ingress during the early stage of the primary-pipe rupture accident, experimental and analytical studies are performed on the conjugate phenomenon of the transient molecular diffusion and natural convection of a two-component gas mixture in two test sections, a reverse-U-shaped tube and a test model simulating simply the reactor. One-dimensional basic equations for continuity and momentum conservation are numerically solved to obtain a concentration change of gas species and an initiation time of a natural circulation of pure nitrogen in the reverse-U-shaped tube. Moreover, a modified numerical solution is proposed to reduce the computing time. A one-dimensional flow net work model is employed to calculate the transport process of air in the test model simulating the reactor. The calculated results agree well with the experimental ones on the concentration change of gas species and the initiation time of the natural circulation of pure nitrogen or pure air.

Vibration monitoring of the primary piping system during the hot functional tests of the Mülheim-Kärlich PWR

During the hot functional tests of the Mülheim-Kärlich plant, which was the first plant of its type, vibration measurements were made on the reactor pressure vessel and its internal parts and on the primary piping system and the main coolant pumps. This paper contains the results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement program is to confirm that the components, which are of new structural design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. A good correspondence was found. In the course of these comparisons, information about the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained. The vibration of the main coolant pumps was also continuously monitored. The pump surveillance system for each pump includes two non-contacting displacement sensors for measuring the kinetic shaft orbit, as well as velocity sensors for recording the vibrational velocity of the pump motor housing. During the continuous monitoring, it was checked whether the signal amplitudes remained within the allowable limits. In addition the frequency content of the signals was determined periodically. In this way deviations could be detected immediately and be explained by means of subsequent correlation analysis. Thus amplitude changes resulting from resonance effects were identified.

Evaluation of LBB in Piping Considering Multiple Fatigue Crack Growth.

Through the recent development of fracture mechanics methodology in piping analysis, the LBB concept has been recognized as applicable to the safety design of LWR primary circuit piping. An important subject still remaining in the LBB analysis is the consideration of multiple fatigue crack growth and development in crack geometry in piping. In the evaluation procedure currently employed, a simple crack geometry is assumed to judge whether the crack is leak-detectable or not. The paper describes a LBB evaluation procedure in which the growth of multiple fatigue cracks is considered. Two criteria, i.e., critical condition of crack coalescence preceding the penetration of wall thickness and evaluation of crack configuration at the penetration, are introduced to account for cracks that coalesce and develop to the largest possible through-wall single crack at the onset of penetration. The paper also presents the results of a case study of PLR piping of BWR using the above procedure. It is shown that LBB can be justified in PLR piping of 4 inches or greater in diameter.

Satisfying flow criteria in the reactor vessel of an atomic power plant

In planning nuclear power plants, special attention is paid to their safe operation. Thus, starting in 1980, our country is completing a large scientific research program on reactor safety of nuclear power plants after loss of containment of any element. The chosen concept is based on the unconditional realization of the criteria "flow before failure." The choice became possible thanks to design and operation tests of various types of nuclear power plants, and also by achievements in the mechanics of failure. The destruction of the physical continuity of metals is related to the presence or birth and subsequent development of crack-like defects under a cyclic load. This means that leakage of the vessel and piping is a consequence of fatigue crack growth from the moment the crack appears to the onset of local instability. It is important that the construction features of nuclear power plant reactors prevent accidental breakage of any primary piping from leading to serious safety consequences: the requirements of the regulations [I] must be satisfied with a margin. Therefore the basic goal of this analytical and experimental research was to analyze possible seal failures and their expected size. This problem, generally speaking, has a stochastic character, because the load capacity Of a body depends substantially on the size and number of initial defects. However, considering the consequences of such accidents, a deterministic approach was used to solve this problem. Experiments on large-scale models of vessels made from reactor-grade steel using standard technology were used to establish possible seal failure mechanisms for fatigue loads as a function of initial crack diameters and the acting stresses. Moreover, analytical formulas were derived which can be used for quantitative evaluation of the extent of seal loss after penetrating cracks are formed [2, 3]. The engineering methodology, developed from experimental data, was used to analyze possible seal loss in the core of a nuclear power plant and to determine the maximum flow rates. As a result, it was shown that the structural integrity of the metal in the vessel is not destroyed, even if a defect through a quarter of the wall thickness occurs anywhere on the vessel. Problems Investigated. The problems posed were designed to study both hazardous conditions, which result from the presence and development of cracklike defects, and also methods to prevent them. In order to solve these problems, a scientific research program was completed, which allowed the development of an engineering methodology for qualitative analysis of crack development in reactor vessels in order to evaluate the probability and scale of seal loss under nuclear power plant conditions [4].

APFIM and AEM investigation of CF8 and CF8M primary coolant pipe steels

Atom probe field ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF8 and CF8M primary coolant pipe steels after long term thermal aging. These cast steels have a duplex microstructure consisting of austenite with approximately 15 vol. -% ferrite. In material aged at 300 or 400°C for up to 70 000 h, the ferrite had spinodally decomposed into a modulated fine scale interconnected network consisting of an iron rich a phase and a chromium enriched α′ phase with a periodicity of between 2 and 9 nm. Roughly spherical G phase precipitates 2 to 10 nm in diameter were also observed at concentrations of more than 1021 m−3. The degradation in mechanical properties of these materials is a consequence of the spinodal decomposition and G phase precipitation in the ferrite.MST/1184

APFIM and AEM investigation of CF8 and CF8M primary coolant pipe steels

Atom probe field ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF8 and CF8M primary coolant pipe steels after long term thermal aging. These cast steels have a duplex microstructure consisting of austenite with approximately 15 vol. -% ferrite. In material aged at 300 or 400°C for up to 70 000 h, the ferrite had spinodally decomposed into a modulated fine scale interconnected network consisting of an iron rich a phase and a chromium enriched α′ phase with a periodicity of between 2 and 9 nm. Roughly spherical G phase precipitates 2 to 10 nm in diameter were also observed at concentrations of more than 1021 m−3. The degradation in mechanical properties of these materials is a consequence of the spinodal decomposition and G phase precipitation in the ferrite.MST/1184

An Alkaline Prefilming Process for the Primary Piping in a New Japanese Boiling Water Reactor

A prefilming method called the alkaline process was successfully applied to a new Japanese boiling water reactor plant during the nuclear heatup phase during the plant preoperation phase. This method was developed to suppress deposition of radioactive corrosion products such as 60Co on ex-core components, which are primarily made of stainless steel. When stainless steel is exposed to actual reactor water that contains oxygen at pH of 8.0 to 8.5, at 288°C, for 410 h, corrosion-resistant oxide films can be formed. The results of the exposure tests indicate that an ∼30% reduction in the deposition of 60Co and 58Co was obtained by this process compared to ordinary operation procedures.

Experimental Study on PORV Break LOCA in PWR Plants

Small break loss-of-coolant accident (LOCA) tests simulating a pressurizer power operated relief valve (PORV) break LOCA were performed using the EOS (Emergency of System) test facility. The break sizes were 0.25 and 0.88% of a guillotine break of a primary piping. The following major conclusions were obtained and the useful data and information for the verification of a computer code were obtained: (1) The pressurizer was almost full of water due to flooding limitation in the surge line of the pressurizer, when no water was in the hot leg piping. (2) The core was kept completely covered with two-phase mixture during the small break LOCA. (3) The core was sufficiently cooled down by reflux condensation in the steam generator even after the primary system natural circulation stopped. (4) After the depressurization of the primary system was stopped or when the depressurization rate of the primary system was small, the operator action to bleed the steam in the secondary system could steadily depressurize the primary system. (5) This operator action could resume the natural circulation in the primary system.