Emergency Core Cooling Research Articles

Integral Testing of the AP600 Passive Emergency Core Cooling Systems

In support of the development of AP600, Westinghouse is conducting two integral systems tests to examine the performance of the passive safety systems. A full-height, full-pressure test is being performed to simulate a small loss-of-coolant, steam generator tube rupture and large steam line break events. A one-quarter scale, low-pressure test is being performed to simulate transients with emphasis on the transition to the natural circulation post-accident, long-term cooling mode and to demonstrate the long-term cooling capability. Each of the tests will provide detailed experimental results for verification of the accident analysis computer codes.

Theoretical and Experimental Analysis of the Bistable Convection Loop for Liquid-Metal-Cooled Reactor Emergency Core Cooling

In this paper, an emergency core cooling system incorporating a bistable convection loop (BCL) for current passive liquid-metal-cooled reactors is proposed. The system has two stable operating modes. During the off mode, the system is in a pure conduction mode and transfers very little heat. In the on mode, the system switches to the low-resistance configuration of a closed natural convection loop and transfers significant amounts of heat. The switching occurs passively because of changes in the reactor temperature. Theoretical and experimental analysis shows that a BCL designed to remove 7% of peak reactor power in the on mode loses only 0.0007% in the off mode, yielding a ratio of 10,000:1.

Downcomer and tie plate countercurrent flow in the Upper Plenum Test Facility (UPTF)

The limitation of vertical steam-water countercurrent flow, called flooding, is important for the operation of Emergency Core Cooling (ECC) Systems in Nuclear Reactors. The ECC water injection flooding behavior is scale dependent and the reactor size behavior cannot be extrapolated from small scale data. A new flooding correlation is presented, based on the classic flooding equation where the effects of gravity, interphase momentum exchange, and instability of the gas/liquid interface are considered. Development of a new correlation was necessary in order to correlate the reactor scale downcomer and upper tie plate countercurrent flow data gained in the Upper Plenum Test Facility (UPTF). The new correlation is an extension of the well-known Wallis-type and Kutateladze-type correlations. Each of the three correlations is valid for the experimental facilities on a certain scale. The range of applicability for each of the three correlations is defined for the case of downcomer and tie plate countercurrent flow.

Loop seal clearing and refilling during a PWR small-break LOCA

During a cold-leg small-break loss-of-coolant accident (LOCA) in a Westinghouse-type pressurized water reactor (PWR), the core liquid level is depressed temporarily before steam clears liquid out of the primary-loop loop seals. The loop seal clearing and associated core level depression may occur repeatedly if the emergency core coolant (ECC) refills the loop seals after clearing. This paper presents an interpretative description of experimental results on loop seal clearing and refilling phenomena obtained at the ROSA-IV Large Scale Test Facility (LSTF).

Quantifying reactor safety margins Part 6: A physically based method of estimating pwr large break loss of coolant accident pct

The Nuclear Regulatory Commission (NRC) has issued a revised Emergency Core Cooling System (ECCS) rule allowing the use of best estimate computer codes for safety analysis. To support this revised rule the NRC and its consultants and contractors have developed and demonstrated the Code Scaling. Applicability and Uncertainty (CSAU) Evaluation Methodology. This effort lead to increased understanding of the phenomena, and their relative dominance, during Large Break Loss of Coolant Accidents (LBLOCAs) in Pressurized Water Reactors (PWRs). Consequently, it became possible, as is done in this paper, to develop a method for establishing clad temperature history by using physically based arguments and engineering correlations. The results from this method are compared with similar uncertainty estimates based on large computer codes. These comparisons provide a rationale, based on physical arguments, for evaluating the large computer code based estimates of uncertainty.

Quantifying reactor safety margins part 4: Uncertainty evaluation of lbloca analysis based on trac-pf1/mod 1

The Nuclear Regulatory Commission (NRC) has issued a revised Emergency Core Cooling System (ECCS) rule allowing the use of best estimate computer codes for safety analysis. The rule also requires an estimation of the uncertainty in the calculated system response and a comparison of the resulting bound with the acceptance limits of 10CFR Part 50. To support this revised rule the NRC and its consultants and contractors have developed and demonstrated the Code Scaling, Applicability and Uncertainty Methodology (CSAU). The last of the three elements of the methodology - Uncertainty Evaluation - is described in this paper.

Quantifying reactor safety margins part 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology

In August 1988, the Nuclear Regulatory Commission (NRC) approved the final version of a revised rule on the acceptance of ECCS entitled “Emergency Core Cooling System; Revision to Acceptance Criteria.” The revised rule states an alternate ECCS performance analysis, based on best-estimate methods, may be used to provide more realistic estimates of plant safety margins, provided the licensee quantifies the uncertainty of the estimates and includes that uncertainty when comparing the calculated results with prescribed acceptance limits. To support the revised ECCS rule, the NRC and its contractors and consultants have developed and demonstrated a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology. The CSAU methodology and an example application, described in this set of six papers, demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. It addresses in a comprehensive and systematic manner questions concerned with: (1) code capability to scale-up processes from test facility to full-scale nuclear power plant (NPP). (2) code applicability to safety studies of postulated accident scenario in a specified NPP, and (3) quantifying uncertainties of calculated results. The methodology combines a “top-down” approach to define the dominant phenomena with a “bottom-up” approach to quantify uncertainty. The methedology is able to address both: (1) uncertainties for which bias and distribution are quantifiable, and (2) uncertainties for which only a bounding value is quantifiable. The methodology is general, and therefore applicable to a variety of scenarios, plants, and codes. This paper provides an overview of the CSAU evaluation methodology and its application to a postulated cold-leg, large-break loss-of-coolant accident in a Westinghouse four-loop pressurized water reactor with 17 x 17 fuel. The code selected for this demonstration of the CSAU methodology was TRAC-PF1/MOD1, Version 14.3.

Quantifying reactor safety margins part 2: Characterization of important contributors to uncertainty

The NRC has issued a revised Emergency Core Coolant System (ECCS) rule which allows the use of best estimate computer codes for safety analysis, providing the uncertainty of the calculations are quantified and compared with acceptance limits contained in 10 CFR Part 50. To support the revised rule, the NRC and its contractors and consultants have developed and demonstrated a methodology to quantify uncertainty called Code Scaling, Applicability and Uncertainty (CSAU). The methodology consists of three primary elements containing 14 steps. The first element, “Requirements and Capabilities”, which contains the first six steps, is described and demonstrated in this paper. The objective of this element is to characterize the important contributors to uncertainty. The objective is accomplished by determining the applicability of a code to analysis of a transient in a Nuclear Power Plant (NPP) through comparison of the scenario- and plant-dictated requirements with the simulation capabilities of the code.

Assessment of TRAC-PF1 Condensation Heat Transfer Model for Analysis of ECC Water Injection Transients

Direct contact condensation of vapor occurs after injection of emergency core cooling (ECC) water during a loss-of-coolant accident (LOCA) in a light water reactor (LWR). Despite that vapor condensation may have large influences on system thermal-hydraulic responses, it is difficult to predict the condensation rate accurately during ECC water injection. This difficulty sometimes results in considerable discrepancy between predicted and measured system overall transients. For instance, analyses of PWR reflood tests with the TRAC-PF1/MOD1 code predicted non-physical large flow oscillations because the code largely overpredicted the condensation rate. In an attempt to resolve this shortcoming of TRAC, a sensitivity study was carried out on the condensation heat transfer Stanton number, that is assumed to be a constant in TRAC, by analyzing the broken-loop hot leg responses to ECC injection in a combined injection test conducted at the Cylindrical Core Test Facility (CCTF). Based on this sensitivity study, it is recommended to use a Stanton number of 0.0035, which is about six times smaller than the value used in the current version of TRAC, to predict LWR system responses during an ECC water injection transient.

The GO-FLOW Methodology: A Reliability Analysis of the Emergency Core Cooling System of a Marine Reactor Under Accident Conditions

A reliability analysis using the GO-FLOW methodology is given for the emergency core cooling system (ECCS) of a marine reactor experiencing either a collision or a grounding accident. The analysis ...

Assessment of TRAC-PF1 condensation heat transfer model for analysis of ECC water injection transients.

Direct contact condensation of vapor occurs after injection of emergency core cooling (ECC) water during a loss-of-coolant accident (LOCA) in a light water reactor (LWR). Despite that vapor condensation may have large influences on system thermal-hydraulic responses, it is difficult to predict the condensation rate accurately during ECC water injection. This difficulty sometimes results in considerable discrepancy between predicted and measured system overall transients. For instance, analyses of PWR reflood tests with the TRAC-PF1/MOD1 code predicted non-physical large flow oscillations because the code largely overpredicted the condensation rate. In an attempt to resolve this shortcoming of TRAC, a sensitivity study was carried out on the condensation heat transfer Stanton number, that is assumed to be a constant in TRAC, by analyzing the broken-loop hot leg responses to ECC injection in a combined injection test conducted at the Cylindrical Core Test Facility (CCTF). Based on this sensitivity study, it is recommended to use a Stanton number of 0.0035, which is about six times smaller than the value used in the current version of TRAC, to predict LWR system responses during an ECC water injection transient.

Reproducing Cultural Identity in Negotiating Nuclear Power: The Union of Concerned Scientists and Emergency Core Cooling

This paper advances the concept of `cultural identity' to account for the nexus between structure and practice in technological negotiations. It describes how the formation of the Union of Concerned Scientists (UCS), and that group's subsequent discourse and nonverbal actions, both reproduced the established identities of group members and contributed to negotiations that reconstituted those identities. In particular, UCS claims about emergency core-cooling systems in nuclear plants were congruent with the combination of a shared ideology, the social interests of MIT faculty, and established principles of engineering design. The cultural analysis of identity reproduction shows the opposition between cognitive and social phenomena to be a significant distinction framing action in Western culture. The analysis also suggests that new attention be given to the relationship between the constitutive and reproductive functions of discourse and nonverbal action.

Analysis of Instrument Tube Ruptures in the Zion-1 Pressurized Water Reactor

In Westinghouse four-loop pressurized water reactors (PWRs), many long small-diameter instrument tubes are employed to route flux monitoring instrumentation lines from the lower plenum of the reactor vessel to a flux mapping seal table. A recent safety concern is that a seismic event could hypothetically rupture instrument tubes at the seal table, effectively causing a lower plenum small-break loss-of-coolant accident (SBLOCA). Continued cooling of the reactor core during a SBLOCA requires depressurization of the primary coolant system such that emergency core cooling (ECC) injection flow balances the break flow. For a lower plenum SBLOCA, the break remains liquid-covered, thus retarding primary coolant system depressurization. As a result, for continued cooling of the core, the break must be sufficiently small such that ECC flow balances break flow at elevated pressures. This study investigates instrument tube ruptures at the seal table location. Separate effects analyses investigate instrument tube pressure and heat loss, instrument lines remaining within the tubes, and tube nodalization effects. Systems effects analyses evaluate the significance of the safety concern through a best-estimate, single-failure analysis for the Zion-1 PWR.

BWR Loss-of-Coolant Accident Tests at ROSA-III with High Temperature Emergency Core Coolant Injection

The effects of emergency core coolant (ECC) temperature on the performance of the emergency core cooling system (ECCS) during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) were investigated experimentally using the Rig of Safety Assessment (ROSA)-III integral test facility. The ECC temperature had no direct influence on the ECCS core cooling performance, since ECC became nearly saturated before reaching the core irrespective of the initial temperature, however, had indirect effects by changing the vessel pressure response. The ECCS injection timings and flow rates, and the core inlet flooding behavior were affected. The measured peak cladding temperature (PCT) was not affected by the ECC temperature for both large (200%) and small (5%) break tests.

BWR loss-of-coolant accident tests at ROSA-III with high temperature emergency core coolant injection.

BWR loss-of-coolant accident tests at ROSA-III with high temperature emergency core coolant injection.

LOFT experiment LP-02-6 analysis by RELAP5/MOD2 code with improved minimum film boiling temperature.

Groeneveld-Stewart's minimum film boiling temperature correlation was incorporated into the RELAP5/MOD2 code in order to explicitly define the minimum film boiling temperature. The transition boiling curve in the code was also modified. The Loss-of-Fluid Test (LOFT) experiment, Experiment LP-02-06 which was a cold-leg double-ended break LOCA experiment with minimum emergency core coolant injection, was analyzed with the modified RELAP5/MOD2 code. The modified RELAP5/MOD2 code well calculated system transients including the rod surface temperature transient. The temporary rewetting of rods in the early phase of blowdown, which had not been predicted by the original RELAP5/MOD2 and other codes, was predicted by the modified RELAP5/MOD2 code.

An experimental investigation of BWR/4 long-term cooling with either intact or broken jet pumps

Experiments simulating emergency core cooling (ECC) conditions after a hypothetical design basis loss-of-coolant-accident (LOCA) in a 218-BWR/4, for either intact or broken jet pump seals, were performed at RPI in a special parallel channel effects (PCE) test section in which Freon-114 was used as the working fluid. The experimental results show that, as expected, core uncovery occurs first in the cold channels. These data were shown to agree with the results of an analytical drift-flux model, and indicate that broken jet pump seals can lead to early core uncovery, and may have a determinal effect on the long-term cooling of a BWR/4. Moreover, it was found that breaks in the lower seal are far more detrimental than a breach of the throat-to-diffuser slip joint.

Considerations for Realistic Emergency Core Cooling System Evaluation Methodology for Light Water Reactors

Various phenomena governing the course of large- and small-break loss-of-coolant accidents in light water reactors and affecting the key parameters such as peak cladding temperature, and timing of the end of blowdown, beginning of reflood, and complete quench have been identified. The models and the correlations for these phenomena in the current literature, in advance codes, and as prescribed in the current emergency core cooling system methodology as outlined in Appendix K of CFR50 have been reviewed. It was found that the models and correlations in the present best-estimate codes such as TRAC or RELAP5 could be made more realistic by incorporating some of these models from the literature. However, an assessment program will be needed for the final selection of the models for the codes.

Fluid and pressure oscillations occuring at direct contact condensation of steam flow with cold water

Pressure and fluid oscillations at pressure suppression containment and cold leg flow oscillations at Emergency Core Cooling Water Injection in water-cooled nuclear power reactors are studied concerning flow oscillations occuring at direct contact condensation of steam flow with cold water. Classifications and mechanisms of these oscillation patterns are presented. Linear frequency analyses are developed in explaining the oscillation frequency and oscillation threshold of both condensation oscillation and plug oscillation. Oscillation frequencies at chugging and at ON-OFF oscillation are examined with analytical models developed for such large amplitude oscillations. Analytical results with these models are compared with the data by simulation experiments. Discussions are made focusing on the applicability of the analyses.

Use of load path dependent material fracture toughness values (WPS) in the safety analysis of RPV stade

The effect of warm prestressing (WPS) has been investigated for the specific material and loading conditions of the central circumferential weld of the reactor pressure vessel of KKS under emergency core cooling. Warm prestressing results in a significant rise of effective fracture toughness. The magnitude of the WPS-effect as a function of warm prestressing, path of unloading, amount of cooling (e.g. change of yield stress) can be predicted on the basis of a theoretical model of Chell et al.. Crack initiation can be excluded for emergency core cooling (ECC)-loading and for material conditions beyond end of life for the central circumferential weld of the pressure vessel of KKS.