Main Steam Line Research Articles

RETRAN-02 MOD5 and RETRAN-03 code benchmarks for Beaver Valley plant transients and FSAR analyses

A RETRAN-02 MOD5 model and a RETRAN-03 model for the Beaver Valley Unit 1 plant were developed from an existing, but untested RETRAN deck. This study uses FSAR transient analyses, actual plant event data, and similar plant analyses performed with RETRAN and TRAC-BF1 MOD1 to benchmark the RETRAN-02 MOD5 and RETRAN-3 plant model. Eight transients were chosen to qualify the RETRAN models: (1) a standard reactor trip, (2) a loss of one reactor coolant pump, (3) a loss of all reactor coolant pumps, (4) an inadvertent opening of a pressurizer relief valve, (5) a loss of normal feedwater with auxiliary feedwater available, (6) a turbine trip, (7) a main steam line isolation valve closure, and (8) a main steam line break. These design basis transients were each chosen to isolate specific phenomena or plant design aspects of the RETRAN models that are exhibited in depressurization events, over-pressurization and over-temperature transients, and over-cooling accidents. Both the RETRAN-02 MOD5 and RETRAN-03 Beaver Valley models were successfully benchmarked. The RETRAN results compared well with the FSAR LOFTRAN analyses, an actual Beaver Valley single pump trip event, and Ringhals 2 and 4 analyses. The RETRAN-03 model performed well and provided a better best-estimate of plant phenomena than the RETRAN-02 MOD5 results; moreover, both the RETRAN models predicted more realistic transient behavior than the FSAR LOFTRAN results.

Dynamic response of reactor internals to pipe breaks

This paper investigates the horizontal responses of the reactor internals to a 14 inch safety injection nozzle break, which is expected to cause the largest loads of the branch line pipe breaks defined for a nuclear power plant. It examines the effects of two forcing terms, reactor vessel motions and internal hydraulic loads, and suggests a new procedure which can be used for tributary pipe break analysis. The analysis result confirms the applicability of the suggested procedure to small size tributary pipe break analysis. Also, this paper calculates the horizontal responses of the reactor internals to a 3 inch pressurizer spray line nozzle break, which is the only one remaining on the primary side after leak-before-break evaluation, and secondary side pipe breaks such as main steam line and economizer feedwater line. The responses are compared with those of safe shutdown earthquake to show that seismic loads with a conservative margin may be used for the pipe break loads in the preliminary design.

Investigation of fluid mixing with direct vessel injection

The object of this work is to investigate fluid mixing phenomena as they related to pressurized thermal shock (PTS) in a pressurized water reactor vessel downcomer during transient cooldown with direct vessel injection (DVI), using test models. The test model designs were based on ABB Combustion Engineering (CE) System 80+ reactor geometry. A cold-leg, small-break loss-of-coolant accident (LOCA) and a main steam line break were selected as the potential PTS events for the ABB-CE System 80+. This work consists of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluids and existing coolant in the downcomer region, and the second part presents the results of thermal mixing tests with DVI in the other test model. Flow visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small-break LOCA. The measured transient temperature profiles compare well with the predictions from the REMIX code for a small-break LOCA, and with the calculations from the COMMIX-1B code for a stream line break event.

Re-evaluation of superheat conditions postulated in NRC Information Notice 84-90 [nuclear plant

Information Notice 84-90, Main Steam Line Break Effect on Environmental Qualification of Equipment, describes a potential problem regarding existing plant analyses and equipment qualification (EQ) related to a postulated main steam line break (MSLB) with releases of superheated steam. This notice states that certain methodologies for computing mass and energy releases for a postulated MSLB did not account for heat transfer from the steam generator tube bundles if they were uncovered. Due to this potential change in the original environmental analysis, the EQ of various components may not consider the thermal environment which could result from superheated steam. Subsequent technical assessments may determine that the existing qualification basis for equipment/components does not envelop the postulated superheat condition. Corrective actions need to be taken to demonstrate that the affected equipment is qualified. The purpose of this paper is to describe the technical re-assessment of the MSLB/superheat condition. >

Development of Software for the Microsimulator for the KO-RI Nuclear Power Plant Unit 2

A workstation-based real-time simulator for two-loop pressurized water reactor plants is developed for classroom training in support of a full-scale simulator, on-site transient analysis, and engineering studies. The present simulator consists of three functional modules: plant module, graphic module, and man-machine interaction module. The plant module includes models for the core kinetics, reactor coolant system, steam generator, main steam line, balance of plant, and control and protection system. Each of the models is optimized to obtain the capability of real-time simulation. The graphic module is designed to provide the user with more information at a glance by dynamically displaying schematic diagrams of the systems, symbols indicating the operating status of each component, trend curves, and the main control room. As tools for the man-machine interface, the man-machine interaction model uses a color cathode ray tube monitor, a standard keyboard, and the mouse. The interactive communication module receives parameters from the user via the keyboard and mouse, and transfers them to the plant module so as to enable the user to perform his specific actions. This module provides the user with various initiating events (malfunctions and manual controls) through SYSTEM, CONTROL ROOM, and ACCIDENTS menus, and thus a wide range ofmore » nuclear steam supply system transients can be easily simulated. The FISA-2/WS is verified through comparisons with analytical solutions, separated tests and integral tests, and predictions by RETRAN-2 and RELAP5/MOD3.« less

Simulation and analysis of a main steam line transient with isolation valves closure and subsequent pipe break

Simulation and analysis of a real main steam line break transient at the Thermal Power Plant Drmno are presented. The main events of the transient were the closure of isolation valves in front of a high pressure turbine, an opening of a by‐pass line, and subsequent pipe break in front of isolation valves. Intensive pressure waves were generated and they propagated through the pipe network of the steam line, causing high fluid dynamic forces on the structure. The transient has been simulated by the computer code TEA‐01, based on the Method Of Characteristics with three characteristic directions. Several main steam line boundary conditions have been modelled and verified. Numerical results are compared with plant data logger records. Simulation has been performed for various scenarios in order to investigate the plant behaviour sensitivity on the boundary conditions. The phenomenology of the pressure waves propagation and the influence of the boundary conditions on these processes are described in detail, as well as fluid dynamic forces during the closure of isolation valves and subsequent pipe break in a section of the steam line in the vicinity of the pipe break.

Coupled 3-D kinetics thermal-hydraulic analysis of Hot Zero Power main steam line breaks using RETRAN and STAR codes

The Main Steam Line Break (MSLB) at End of Life (EOL), Hot Zero Power (HZP) conditions accident was analyzed using a fully time-dependent coupled thermal-hydraulic (T/H) and neutronics method, and compared against conservative Final Safety Analysis Report (FSAR) results, which predict a return-to-power. The development and improvement of coupled neutronics/T/H analysis techniques provide many advantages including the capability to evaluate the impact of modeling assumptions made in previous reactor kinetics and T/H calculations. The coupled STAR kinetics and RETRAN T/H technique developed here provides a means to evaluate the quasi-static, point kinetics approximation against a fully time-dependent, three-dimensional approach. Using the state-of-the-art 3-D STAR reactor kinetics code with the RETRAN reactor coolant system (RCS) T/H code in a best-estimate approach, it is now possible to evaluate the impact on safety margins imposed by conservative FSAR MSLB assumptions. The method presented shows how the time-dependent 3-D STAR nodal code model was used directly with core inlet conditions determined by RETRAN for a Westinghouse PWR. The STAR/RETRAN results clearly demonstrate that a return-to-power is NOT predicted when a 3-D thermal-hydraulically coupled time-dependent kinetics approach is used. This study shows that: (a) quasi-static and point kinetics methods are not able to describe severe PWR asymmetric transient phenomena adequately; and (b) fully coupled, 3-D time-dependent analysis methods should be used for PWR reactor transients instead. By coupling the RCS response in terms of updated core inlet conditions with 3-D time-dependent core kinetics response, in a tandem manner, the core power and T/H RCS conditions are forced to be self-consistent during the entire event, when non-equilibrium conditions exist.

Optimizing Feedwater Control System Performance with the Chinshan Plant Analyzer

To apply fast and accurate simulation techniques to Taiwanese nuclear power plants, the Chinshan plant analyzer was developed based on the Brookhaven National Laboratory boiling water reactor (BWR) plant analyzer. The Chinshan plant analyzer provides user-friendly, on-line, interactive simulation capability with graphics display and is suitable for control system analysis. During the generator load rejection (GLR) test at the Chinshan BWR power station located in northern Taiwan, the reactor feedwater pump (RFP) tripped because of a high downcomer level (level 8). Feedwater control was then lost because of the RFP trip. By the end of the transient, a huge amount of water had accumulated in the reactor pressure vessel. The margin to main steamline flooding was decreased. An optimization module was developed and added to the Chinshan plant analyzer. With the optimized feedwater controller settings, the maximum downcomer level is below level 8, and RFP does not trip during the GLR transient. The margin to main steamline flooding is increased. These techniques will be applied for improving plant performance in the near future.

Improved Pressurized Water Reactor Steamline Break Analysis Using RETRAN-02, ARROTTA, and VIPRE-02

The steamline break accident is one of several specified severe transients addressed in the final safety analysis report for any pressurized water reactor plant as part of the licensing procedure. A rupture in a main steamline in the secondary system causes a sudden cooling of the water in the corresponding primary loop. The cold water flowing into part of the core represents a positive reactivity insertion that must be contained by control rods, which are scrammed into the core almost immediately. Later in the scenario, soluble boron reaches the core from the emergency core cooling system. When simulating a steamline break accident during the licensing procedure, many conservative assumptions are added to the transient description. Historically, a steamline break analysis is performed with a system analysis code like RETRAN, using a rather simplified (point kinetics) description of the core. The three-dimensionality of the event within the core is accounted for by constant blending factors, which are used to calculate the evolving point kinetics parameters based on a simplistic cold and hot partition of the core. In this paper the ARROTTA-01 and VIPRE-02 computer codes are coupled to allow a detailed three-dimensional simulation of the reactor core during a steamline breakmore » event.« less

Transient Analysis of the MSIV-ATWS in a 1000-MW(thermal) BWR-4

The integrity of MARK-1 containments is severely challenged by the consequences of a postulated main steam isolation valve-anticipated transient without scram (MSIV-ATWS) at full power. To elucidate the nuclear plant's response to this postulated beyond-design-basis accident, considerable research has been conducted, primarily in the United States. Paul Scherrer Institute is currently engaged in a sizable effort to develop the simulation models for the Transient Analysis of Reactors in Switzerland (STARS project). The current study as part of STARS aims primarily at the physical phenomena occurring during an MSIV-ATWS in a domestic boiling water reactor (BWR)-4. This plant was selected because it has the following features different from the ones widely analyzed: The capacity of the torus (primary containment) is increased by roughly a factor of 2 compared with other nuclear power plants of similar design. The feedwater pumps are motor driven and therefore do not automatically stop operation after a main steamline isolation event. The difference between the setpoints for opening and closing the safety relief valves (SRVs) is 12% of the opening pressure. This is considerably larger than the comparable values of the plant analyzed abroad. This paper reports that core power and core size are only about one-thirdmore » of the corresponding values for the boiling water reactor analyzed so far.« less

Dynamic characteristics of reactor core for pipe break and seismic excitations

This paper investigates the dynamic responses of the reactor core for main steam line and economizer feedwater line breaks. The tributary pipe breaks in lieu of main coolant loop breaks are considered because leak-before-break evaluation has provided a technical basis for the elimination of double ended guillotine breaks. This paper also calculates the response of the reactor core due to the motions induced from safe shutdown earthquake and operating basis earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, reported in this paper are the different response characteristics of each pipe break and seismic excitation.

Analyses of fatigue crack growth and fracture in carbon steel piping

The objective of the present study is to evaluate the fatigue crack growth behavior and fracture conditions of Japanese carbon steel piping, which are relevant for the Leak-Before-Break behavior. A fatigue crack growth analysis was conducted for a circumferential inner-surface crack in the main feedwater line, the main steam line and the ECCS line under the design loading conditions of BWR and PWR plants. The fatigue analysis showed that a crack does not penetrate the pipe wall under the design loading conditions if the initial crack depth is equal to or smaller than 10–20% of the pipe wall thickness. Conditions for stable crack growth and pipe failure were analyzed by non-linear fracture mechanics. Good agreement between experiment and analysis was shown in the load-deformation relationship during stable crack growth. Fracture conditions were described with a good accuracy by the net-section collapse criterion for a STS42 6-inch diameter pipe.

Nuclear Power Plant Status Diagnostics Using an Artificial Neural Network

In this work, nuclear power plant operating status recognition is investigated using a self-optimizing stochastic learning algorithm artificial neural network (ANN) with dynamic node architecture learning. The objective is to train the ANN to classify selected nuclear power plant accident conditions and assess the potential for future success in this area. The network is trained on normal operating conditions as well as on potentially unsafe conditions based on nuclear power plant training simulator-generated accident scenarios. These scenarios include hot- and cold-leg loss of coolant, control rod ejection, total loss of off-site power, main steamline break, main feedwater line break, and steam generator tube leak accidents as well as the normal operating condition. Findings show that ANNs can be used to diagnose and classify nuclear power plant conditions with good results. Continued research work is indicated.

Development of Evaluation Code for Long-Term Cooling of Containment with Passive Safety Feature

Simplified BWRs are characterized as an adoption of a passive ECCS and a passive containment cooling system (PCCS). While a passive ECCS has a short term core cooling function, a PCCS has a long-term decay heat removal function. As a PCCS, several concepts, differing in cooling location and method employed, have been considered. From the containment thermal- hydraulic response analysis viewpoint, simplified BWRs are essentially different from the current BWRs. For evaluating and comparing the performance of several PCCSs over full break spectra, the new containment safety evaluation code TOSPAC was developed as a preliminary design tool for PCCS. This paper summarizes the thermal-hydraulic modelings of the TOSPAC code and the validity evaluation of the TOSPAC code, compared with TRAC-BF1 calculation. From the validity evaluation concerning a main steam line break (MSLB) accident analysis for an isolation condenser (I/C) as a PCCS, it was found that the TOSPAC calculation result shows reasonable agreement ...

Heat Removal Performance Evaluation of Several Passive Containment Cooling Systems during Loss of Coolant Accident.

As a passive containment cooling system (PCCS), which is adopted in simplified BWRs, several concepts, differing in cooling location and method, such as the suppression chamber water wall, the drywell water wall, the isolation condenser (I/C) and the drywell cooler, have been considered. This paper summarizes the characteristics of each PCCS concept, and the analysis results of the performance for several PCCSs during a main steam line break LOCA for a reference simplified BWR plant, obtained by the newly developed containment thermalhydraulic response analysis code TOSPAC. The performance comparison suggests that I/C and drywell cooler have good heat removal capability with regard to the smallest heat transfer area among PCCS concepts evaluated in the present analysis. I/C removes decay heat efficiently, since it absorbs steam directly from the reactor pressure vessel, which is the hottest portion inside the containment. The suppression chamber water wall is ineffective, mainly due to high non-condensabl...

FISA-2/WS, A Compact Real-Time Simulator for Two-Loop Pressurized Water Reactor Plants

The FISA-2/WS simulator pack-age is divided into three modules: plant, on-line graphic display, and interactive communication. The plant module consists of models for core kinetics, reactor coolant system, steam generator, main steam line, and control and safety systems. Simulation results are displayed periodically at a user-specified time interval on a color monitor by the on-line graphic display module. The FISA-2/WS interactive communication module enables the user to initiate or mitigate accidents and to select one of the menu-driven graphic displays with the mouse and keyboard. The results presented here are obtained from simulations of steady state, turbine load change transient, and small-break loss-of-coolant accidents

Chemistry of Radioactive Nitrogen in BWR Primary System

Numerical model calculations and evaluation of plant experience with chemical behaviors of radioactive nitrogen (13N and 16N) in BWR primary systems have suggested the followings : (1) Calculated results showed that the major species of radioactive nitrogen released to the turbine system was nitrogen mono-oxide. (2) Based on 13N measurement in the main steam of a BWR, 85% of 13N was found to be anion. This result supported the prediction above. (3) Calculated results showed that chemical behaviors of 13N and 16N were not the same because of the difference in half-lives of 13N and 16N; 13N was easily accumulated in the core by-pass channels where the residence time of water was several times longer than in fuel channels, but the main source of 16N was the boiling channels in the reactor core. (4) Nitrogen-16 released to the main steam line could be decreased by lowering the hydrogen atom concentration in the water; typically to around 1/10 of the normal level by the addition of nitrous acid of ~1×10~6M at the reactor core inlet.

Chemistry of radioactive nitrogen in BWR primary system.

Numerical model calculations and evaluation of plant experience with chemical behaviors of radioactive nitrogen (13N and 16N) in BWR primary systems have suggested the followings : (1) Calculated results showed that the major species of radioactive nitrogen released to the turbine system was nitrogen mono-oxide. (2) Based on 13N measurement in the main steam of a BWR, 85% of 13N was found to be anion. This result supported the prediction above. (3) Calculated results showed that chemical behaviors of 13N and 16N were not the same because of the difference in half-lives of 13N and 16N; 13N was easily accumulated in the core by-pass channels where the residence time of water was several times longer than in fuel channels, but the main source of 16N was the boiling channels in the reactor core. (4) Nitrogen-16 released to the main steam line could be decreased by lowering the hydrogen atom concentration in the water; typically to around 1/10 of the normal level by the addition of nitrous acid of ~1×10~6M at ...

A Steam Generator Dynamic Mathematical Modelling and Its Using for Adaptive Control Systems Testing

The dynamic once-through subcritical steam generator model, based on the plant design data, is presented. This model allows the investigation of power plant system transients . Detailed description of the simulation results is documented on the example of once-through subcritical steam generator o f 200 MW power unit. For the purpose of both the similaritv with the physical process and intelligibilitv of the model for a design engineer, familiar with technology, the structural boiler model was chosen. This one is composed of the following parts: combustion system, evaporator, uncontrolled and controlled superheaters, main steam line, and a representation o f the boiler load (steam consumer-reheat turbine). The boiler model is applied both to investigate the dynamical behavior of uncontrolled plant and to verify controller s and control concepts too. As the example of the model used for control testing the cascade superheated steam temperature control is described. The superheater model taken from the whole large boiler model was completed with regulating unit, nonlinearities, etc. Various controllers, especially PI and LQ STC, were compared in the role of the main controller.

Waste heat recovery of dura (Iraq) oil refinery and alternative cogeneration energy plant

The first part of this paper presents a waste heat recovery scheme for the Dura (Baghdad, Iraq) oil refinery energy plant. Both the wasted heat of the process return condensate and the flue gases are utilized for low temperature feedwater and fuel heating. The steam saved, both from the main steam line and turbine extraction system, was found to increase the steam and plant overall efficiency by 18%. An alternative cogeneration energy plant is presented in the second part of this study. The proposed plant utilizes the gas turbine exhaust, in conjunction with a heat recovery boiler, to produce the process steam requirement. With this alternative plant, the overall efficiency increases by 31.6%, while the steam efficiency increases by 19%. The outstanding features and advantages of the proposed plants are highlighted.