Anticipated Transient Without Scram Research Articles

The use of neural networks and a prototype expert system in BWR ATWS accidents diagnosis

Abstract Neural networks and a symbolic expert system are employed to form a prototype system in BWR anticipated transients without scram (ATWS) accidents diagnosis. Unsupervised learning based on discovery of cluster structures (Pao's approach) and back propagation (BP) neural networks are used to group and memorize different ATWS patterns. Multiple training data sets derived from output files of the SABRE computer code are used in training the BP networks to cope with oscillations of reactor power and other parameters. Tests of neural networks recall correctness are performed, given the presence of time shift of sample data, random noise and incomplete information. The expert system can simulate the ATWS strategy developed by Pennsylvannia Power & Light company. It undertakes diagnosis, using an event tree structure, and can execute the trained BP networks. The expert system is also able to deal with temporary loss of reactor water level information.

Best-Estimate Methodology for Analysis of Anticipated Transients Without Scram in Pressurized Water Reactors

Union Fenosa, a utility company in Spain, has performed research on pressurized water reactor (PWR) safety with respect to the development of a best-estimate methodology for the analysis of anticipated transients without scram (ATWS), i.e., those anticipated transients for which failure of the reactor protection system is postulated. A scientific and technical approach is adopted with respect to the ATWS phenomenon as it affects a PWR, specifically the Zorita nuclear power plant, a single-loop Westinghouse-designed PWR in Spain. In this respect, an ATWS sequence analysis methodology based on published codes that is generically applicable to any PWR is proposed, which covers all the anticipated phenomena and defines the applicable acceptance criteria. The areas contemplated are cell neutron analysis, core thermal hydraulics, and plant dynamics, which are developed, qualified, and plant dynamics, which are developed, qualified, and validated by comparison with reference calculations and measurements obtained from integral or separate-effects tests.

A Realistic Anticipated Transient Without Scram Evaluation of the Zorita Nuclear Power Plant

A best-estimate methodology for analysis of an anticipated transient without scram (ATWS) in a pressurized water reactor (PWR) is applied to the simulation of the passive response to postulated ATWS scenarios of the José Cabrera (Zorita) nuclear power plant (NPP) owned and operated by Union Fenosa, which is the only Westinghouse PWR with a single coolant loop. A justification of the calculation hypotheses is included. The results of the specific studies are evaluated, and the conclusion is that the intrinsic safety margins of the original design of the plant guarantees the integrity of the fuel, primary circuit, and containment, without the need to incorporate an automatic ATWS mitigation system. Finally, a suitable plant-specific prototype emergency operating procedure is designed that is substantially different from the previous Zorita NPP procedure and from the generic procedure applicable to multiloop plants. This procedure is validated by simulating the operator-plant interface by means of a validation matrix including the scenarios presenting the most adverse dynamic modes foreseeable.

A simulator-based evaluation of operator behavior in a Chinese nuclear power plant under ATWS condition

In this paper the behavior of nuclear power plant operators during anticipated transient without scram (ATWS) situation has been researched using the pressurized water reactor (PWR) simulator in Tsinghua University. Test data from six operating crews have been collected and analyzed. It is concluded that (1) a half-hour safety design rule should be applied; (2) education and training should be improved; (3) the collection of data on human errors should proceed; and (4) the function of the safety engineer should be established.

A Long-Term MAAP 3.0B Analysis of a Severe Anticipated Transient Without Scram Accident in a Boiling Water Reactor

This paper reports that the long-term responses of the reactor coolant system and the containment of a boiling water reactor (BWR) during an anticipated transient without scram (ATWS) initiated by an inadvertent closure of the main steam isolation valves (MSIVs) are analyzed with the MAAP 3.0B computer code. The plant used in the analysis is the Kuosheng nuclear power station, a General Electric BWR6 reactor with a Mark III containment. A comparison of near-term system responses to an MSIV closure ATWS event shows that the MAAP 3.0B code fails to reproduce the detailed thermal-hydraulic results of the sophisticated RETRAN-02 code. The MAAP code, however, makes a reasonable prediction of the timing of major events and core power in the ATWS accident analyzed. The results of a long-term MAAP ATWS analysis show that if the standby liquid control system fails in an MSIV closure ATWS event, the ATWS event might develop into a core melt accident with the containment failing before the core melts.

Passive Safety Features of a Bottom-Supported Fast Breeder Reactor Vessel

A system dynamics analysis is applied to a pool-type fast breeder reactor to examine the influence of a bottom-supported reactor vessel (BSRV) design on anticipated transient without scram (ATWS) events such as an unprotected loss of flow (ULOF), an unprotected loss of heat sink (ULOHS), and an unprotected transient overpower (UTOP) by using the ARGO safety analysis code. The BSRV enhances negative feedback because of the differential displacement between the core and the control rod as compared with a top-supported reactor vessel. In particular, the BSRV has the potential, especially in a mixed-oxide-fueled core, to mitigate the design requirements to prevent boiling of the coolant during an ULOF and ULOHS through the elongation of the primary flow coastdown and enhancement of the axial expansion of the control rod drive line. In the metallic-fueled core, the effects of the BSRV on the ATWS events are diminished by the limitation of the sodium temperature increase.

The NUREG-1150 probabilistic risk assessment for the Grand Gulf Nuclear Station

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Grand Gulf Nuclear Station performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, consequence analyses, and risk results obtained when the results of these analyses are combined with the accident frequency analysis. The offsite risk from internal initiating events was found to be quite low, both with respect to the safety goals and to the other plants analyzed in NUREG-1150. The offsite risk is dominated by short-term station blackout plant damage states. The long-term station blackout group and the anticipated transients without scram (ATWS) group contribute considerably less to risk. Transients in which the power conversion system is unavailable are very minor contributors to risk. The low values for risk can be attributed to low core damage frequency, good emergency response, and plant features that reduce the potential source term.

On the human error probability for injecting boron during ATWS at a BWR

Abstract This note presents a calculation and discussion of the probability of operator failure to inject boron with the standby liquid control (SLC) system during an anticipated transient without scram (ATWS) at a boiling water reactor (BWR). Calculated results are compared to the value used in some past risk assessments, as well as with the estimated SLC hardware unavailability, to determine if the assertion that the SLC hardware unavailability dominates the human error probability is reasonable. The sensitivity of this calculation to the value used for the suppression pool heat capacity temperature limit is also examined.

Critical Heat Flux for Annulus under High-Pressure, Low-Flow and Mixed Inlet Conditions

Dryout experiments of water have been conducted in an annulus with inside heating (heat flux from inner wall only) under high-pressure, low-flow and mixed inlet conditions which are of importance in the core thermal-hydraulic behavior during a loss-of-coolant accident (LOCA) and also partially during an anticipated transient without scram (ATWS) of a nuclear reactor. The experimental conditions have covered ranges of pressure of 3 MPa, mass flux from 105 to 320 kg/m2·s and inlet quality from 0.15 to 0.90. The dryout data have been compared with several existing empirical critical heat flux (CHF) correlations and a new correlation. The Katto correlation predicts best the CHF among the existing correlations examined. However, even the Katto correlation overpredicts the CHF by factors up to 2 at about 1/6 data points of the present dryout data. The present dryout data are divided into two groups (regions) according to the value of a non-dimensional number l bo/d he, where l bo is the assumed boiling length a...

Containment venting as a mitigation technique for BWR MARK I plant ATWS

Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it. Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure. For both cases the operators are assumed to take action to intentionally restrict injected flow such that fuel in the upper part of the core would be steam cooled. Resulting fuel temperatures are estimated with an off-line calculation and found to be acceptable.

Loss-of-Coolant-Accident and Anticipated Transient Without Scram Calculations for Homogeneous and Heterogeneous Advanced Pressurized Water Reactors

Loss-of-coolant-accident (LOCA) and anticipated transient without scram (ATWS) calculations have been performed for the two Kernforschungszentrum Karlsruhe advanced pressurized water reactor refere...

Analysis of a Loft ATWS Event for PWR using the Autoregressive Model

Using a method of time-series analysis, the characteristics of a loss-of-fluid test (LOFr) PWR is analysed and a real-time simulation model for the IDFr reactor dynamics is also developed under the condition of an anticipated transient without scram (ATWS) event. Impulse response functions and step response functions are derived from the developed model. It is 'Shown that the result of the model is in good agreement with actual LOFT ATWS experimental measurements within the time which is from the initiation of ATWS event to its fully development stage. The validity of the developed model can also be analysed with theory. During the operation of reactor, the prediction of ATWS accident is of significant values from the views of both safety and economy. The other j ob done in this paper is that the model for ATWS accident has been developed using the forcasting theorem, and then, the on-line forcasting and diagnosis of the ATWS accident Mve been done using the developed model.

Pressurized Water Reactor Inherent Core Protection by Primary System Thermohydraulics

Current light water reactors (LWRs) depend for the protection of core integrity on a multitude of active systems and components, such as instrumentation, cables, electronic logics, relays, actuators, etc., and on human judgment. This approach to safety has led to a complex and expensive plant design in which all parts of the plant where these systems are present must be protected against damage due to, e.g., earthquake. It has also failed to persuade the public about the safety of the reactors because of the existing (but very small) probability of multiple failures leading to core meltdown.With the process inherent ultimate safety (PIUS) approach, this dependence on active systems is eliminated. The safety is now no longer a result of their intervention but is built into the thermohydraulics of the primary system itself. The PIUS primary system response to a number of severe anticipated transients without scram (ATWS) is described, as studied by means of a specially developed computer simulation program.The method is shown by which the thermohydraulic self-protection properties of the primary system terminates these ATWS transients, which could have severe consequences in a conventional LWR, with neither the core nor the rest of the plant suffering any damage (beyond the initial failure assumed).This has important economic consequences. The surveillance and control systems used to run the plant and the buildings in which they are housed can be designed as for a fossil plant, since they no longer have the ultimate responsibility for nuclear safety. The ensuing design simplification pays for the more expensive pressure vessel and primary system. Inherent safety is obtained as a bonus.

RETRAN-02 Analysis of Loss-of-Fluid Test Experiments L9-3 and L9-4

Two simulations of pressurized water reactor (PWR) anticipated transient without scram (ATWS) sequences were performed in the loss-of-fluid test (LOFT) facility. These were designated as tests L9-3 and L9-4. Test L9-3 is a loss-of-feedwater transient, while L9-4 simulates an ATWS accompanied by loss of off-site power. In the latter case, the main steam valve and primary circulation pumps are tripped at the beginning of the experiment, along with the steam generator feedwater flow. The behavior of these experiments was analyzed with the RETRAN-02 computer code in order to evaluate the capability of the code to predict ATWS behavior against the experimental evidence exhibited by LOFT. The complex sequence of events, which occurs in L9-3, creates a difficult system modeling problem. The primary influence on the entire system response is the steam generator heat transfer rate (given core power as a boundary condition). Results of the analysis demonstrate how the course of the computer calculation is influenced by the steam generator model and its initial conditions. Test L9-4 does not contain the thermal-hydraulic complexity of test L9-3 due to the immediate isolation of the steam generator. Nonetheless, it reinforces the general conclusion that RETRAN-02 will produce an adequate simulation of PWR more » ATWS behavior if the initial and boundary conditions are completely defined. « less

An Investigation of Decreasing Reactor Coolant Inventory as a Mechanism to Reduce Power During a Boiling Water Reactor Anticipated Transient Without Scram

Under certain anticipated transient without scram (ATWS) sequences for a boiling water reactor, it would be desirable to reduce system power, particularly where the primary system has been isolated by closure of all main steam isolation valves and is discharging steam through its safety/relief valve system to the suppression pool. Reducing reactor power increases the time available to shut down the reactor by minimizing the heat dumped to the suppression pool and by helping to keep the suppression pool temperature within limits. Under proposed emergency procedure guidelines for the ATWS event, the reactor water level would be lowered to reduce reactor power. The analyses provide an assessment of the power level that would be attained, assuming the reactor operators were to reduce the the downcomer level down to the top of the active fuel.

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

Analysis of Pressurized Water Reactor Anticipated Transient Without Scram Transients with RETRAN-02

This paper describes the analysis of a loss of feedwater (LOFW) ATWS on a U-tube PWR (pressurized water reactor). The capabilities of the RETRAN-02 code have been extended to allow it to perform pressurized water reactor anticipated transient without scram (ATWS) calculations. A loss-of-feedwater ATWS was performed with RETRAN-02 and compared with vendor calculations with excellent agreement. 2 refs.

Safety implications of stochastic analysis of nuclear reactor transients

A ‘neutron-temperature random process’ is formulated by means of a probability generating function technique. The model accounts for the effect of the negative temperature feedback by treating the temperature as a continuous random variable, thereby making possible the computation of the expected values of the temperature, neutron and cumulative fission densities as well as their standard deviations. The special relationship of the stochastic behavior of the nuclear reactor to safety considerations is discussed. Typical numerical results are presented related to startup accidents and Anticipated Transient Without Scram (ATWS) in thermal reactors.

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