Reactor Trip Research Articles

A Microprocessor Tester for the Treat Upgrade Reactor Trip System

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.

LOFTRAN/RETRAN Comparison Calculations for a Postulated Loss-of-Feedwater ATWS in the Sizewell ‘B’ PWR

The loss-of-main-feedwater transient without reactor trip (scram) has received particular attention in pressurized water reactor (PWR) anticipated transient without scram (ATWS) analysis primarily due to the potential for reactor coolant system over pressurization. To assist in the licensing of the U.K. PWR, Sizewell 'B', comparative calculations of a loss-of-feedwater ATWS have been performed using the Westinghouse-developed LOFTRAN loop analysis code and the Electric Power Research Institute/ Energy Incorporated-developed RETRAN-01 code. The calculations were performed with and without the emergency boration system (EBS), which is included in the Sizewell reference design. Initial results showed good agreement between the codes for the major features of the transient, but also a time shift in the transient profiles at the time of the pressurizer pressure peak. This was found to be due to differences in the steam generator modeling, which resulted in a difference in the onset of the very rapid degradation in heat transfer as the steam generators approach dryout. When the same model was used in both codes, very good agreement was obtained. Remaining differences in the results are attributed primarily to differences in the boron injection models, which resulted in an over-prediction of the core boron concentration in the RETRAN calculation.more » The results with an EBS indicate that the peak pressurizer pressure is relatively insensitive to variations in modeling.« less

Ultimate heat sink for PHWRs—evaluation of existing methods and alternative proposals

Two methods are proposed for removal of decay heat after a reactor trip under loss of class IV power. One is based on natural circulation (thermosyphon) cooling while the other depends on the direct introduction of the standby cooling system (SCS) heat exchanger after the main pump coast-down. The present analysis shows that under bottled condition thermosyphon cooling is adequate to remove up to 10% full power without boiling and up to 12% power with boiling in the primary coolant channels. However, the direct introduction of the SCS obviates the uncertainties of thermosyphon and ensures positive flow driven by a pump available on class III power supply. This flow is so chosen that there is no boiling on the secondary side of SCS within the operable pressure range. The analysis shows that such an operation does not induce undue stresses in the equipment.

Application of Semiconductor Logic Components in Reactor Safety and Regulating Systems

A prototype of reactor trip logic for a CANDU reactor shutdown system, implemented with field programmable logic arrays (FPLAs), was developed as a step towards the phasing out of relays in this application. The prototype uses continuous self-testing to detect and locate faults. A design for control rod interlock logic, based on FPLAs, was also produced. The main design features of both systems are explained. The experience gained in this application of FPLAs is discussed. The potential application of this technology is in complex logic systems where a relay design is too inflexible and time consuming to fabricate, and a programmable controller or computer is impractical or undesirable.

Core temperature transient in the early phase of core uncovering at TMI-2

This paper describes a best-estimate analysis of the initial core boil-down and heat-up transient at Three Mile Island Unit (2) on 28 March 1979. This transient began shortly after all reactor coolant pumps were secured (100 min after reactor trip) and was terminated by a period of sustained high pressure injection of emergency cooling water, starting at 202 min. The analysis is primarily directed to understanding the progression of core damage, rather than providing a detailed characterization of the core end-state condition. The latter objective can be achieved only after vessel head removal and visual examination. The thrust of the present effort has been to: (1) develop a core coolant mixture level (dry-out level) calculation which satisfies the boundary conditions implied by various instrument responses and system operational characteristics; (2) couple the level calculation with a core heat-up modelto simulate the accumulation of thermal damage in the exposed, upper regions of the core; (3) compare calculated gross damage to the core with measurements of hydrogen and fission product releases subsequent to the accident. Results indicate that: 1. (i) Observed containment hydrogen levels were due to Zircaloy/stainless steel corrosion that occurred during the period of core uncovering between the de-activation of the loop A reactor coolant pump (100 min after trip) and sustained operation of the high pressure injection system 100 min later. Appreciable zircaloy oxidation probably commenced at 150 min after trip, and continued at a high rate until the sustained high pressure injection at 202 min caused a major core quench. 2. (ii) There was some potential for fuel liquefaction. Calculations imply that peak fuel temperatures did not exceed the UO 2 pellet melting temperature, but 30% of the fuel was exposed to temperatures where liquid UZrO alloys could have formed. 3. (iii) A substantial fission product release was obtained from fuel over-heating; however, an apparent disparity between the expected fission product release by calculation and the high range of fission product estimates obtained from plant measurements suggests that a significant release fraction may have originated from powdered or rubbilized fuel during cooldown. Additional gas releases may have developed from hot spots which persisted after core quench. 4. (iv) Steam temperatures in the upper plenum, at the outlet nozzle elevation, were generally below 900°C (1650°F) although this value was probably exceeded for a few min during the partial fuel quench caused by activation of the loop 2B reactor coolant pump, at 174 min after trip. The metal-work in the upper plenum, above the upper tieplate did not experience appreciable heating. Thermal damage to the fuel and consequential weakening and mechanical disruption of the core was essentially complete 230 min after turbine trip.

Improved Testability Provisions for Boiling Water Reactors

Safety Systems (Emergency Core Cooling and Reactor trip) circuits at Peach Bottom were provided with complete, overlapping zone, on-line testability. This was accomplished by adding to and greatly expanding the coverage of the manufacturer installed test plugs. The general philosophy is that with proper insertion of test plugs, full logic initiation can be tested including valve operation and pump starts. No jumpers or lifted leads are required and injection into the core is defeated during test. These provisions simplified the testing program for Technical Specification (Tech Spec) requirements and eased licensing considerations. Although applied to a BWR, the principles outlined are applicable to any system or plant using digital inputs.

Dynamic Compensation of Rhodium Self Powered Neutron Detectors

A mathematical model of a rhodium self-powered neutron detector has been developed to investigate the detector transfer function and to apply this transfer function as compensation for the inherent time delay in the rhodium-neutron interaction. The various reactions that comprise the detector signal have been evaluated experimentally and the resultant transfer function compensation has been tested in the field against a prompt responding ion chamber during a reactor trip.

The effect of grid frequency decay transients on pressurized water reactors

This paper describes the effects of loss of reactor coolant flow as caused by power system frequency decay on a Westinghouse pressurized water reactor. A protection scheme employing measurement of coolant flow, and reactor coolant pump power supply voltage and frequency is described. Analyses are presented which provide a basis for a low frequency reactor trip setpoint which adequately protects the nuclear plant, and which does not unduly compromise plant availability.

A dynamic simulation of a nuclear power station for control analysis

Nonlinear simulation has been applied to the control study of an organic-cooled Canadian CANDU reactor. The study was carried out as an integral part of an overall design study. The control characteristics of the power plant were studied for reactor start-up, pump failures, reactor trips and various rates of power change. The response of the power plant to these disturbances is discussed.

The PWR Protection System - Present and Future

In a nuclear power generating station, the system used to protect the reactor and its containment is the plant protection sytem. The pressurized-water-reactor (PWR) protection systems encompass all electrical and mechanical devices and circuitry (from sensors to actuated device input terminals) involved in generating those signals associated with the protective function. These signals include actuation signals for the Reactor Protection System (RPS) and Engineered Safety Features System (ESFS) such as Safety Injection Actuation System, Containment Spray Actuation System, Main Steam Isolation System, Auxiliary Feedwater Actuation System, and Recirculation Actuation System. This paper deals with the present status and future development of the Plant Protection Systems utilized by Combustion Engineering, Inc. (C-E) and Kraftwerk Union (KWU) for their Pressurized Water Reactors. The latest techniques of signal processing, logic design, reactor trip and engineered safety features actuation circuitry, automatic testing, isolation and interfacing, are discussed. Current system design criteria, which as a result of the dynamic national and international licensing atmospheres influence system design philosophies, are highlighted.