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Title: Status Report on Ex-Vessel Coolability and Water Management

Abstract

Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell vent path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.

Authors:
 [1];  [2]
  1. Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division
  2. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE), Nuclear Reactor Technologies (NE-7). Light Water Reactor Sustainability (LWRS) Program; Electric Power Research Inst. (EPRI) (United States)
Contributing Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
OSTI Identifier:
1324704
Report Number(s):
ANL/NE-16/18
130282; TRN: US1700062
DOE Contract Number:
AC02-06CH11357
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ACCIDENT MANAGEMENT; REACTOR CORE DISRUPTION; WATER; BWR TYPE REACTORS; PRESSURE VESSELS; VENTS; FLOORS; DEPTH; M CODES; C CODES

Citation Formats

Farmer, M. T., and Robb, K. R. Status Report on Ex-Vessel Coolability and Water Management. United States: N. p., 2016. Web. doi:10.2172/1324704.
Farmer, M. T., & Robb, K. R. Status Report on Ex-Vessel Coolability and Water Management. United States. doi:10.2172/1324704.
Farmer, M. T., and Robb, K. R. Thu . "Status Report on Ex-Vessel Coolability and Water Management". United States. doi:10.2172/1324704. https://www.osti.gov/servlets/purl/1324704.
@article{osti_1324704,
title = {Status Report on Ex-Vessel Coolability and Water Management},
author = {Farmer, M. T. and Robb, K. R.},
abstractNote = {Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell vent path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.},
doi = {10.2172/1324704},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Sep 15 00:00:00 EDT 2016},
month = {Thu Sep 15 00:00:00 EDT 2016}
}

Technical Report:

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  • The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the requiredmore » reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.« less
  • The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the requiredmore » reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.« less
  • Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergencemore » of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.« less
  • Progress in the EXOSAT data analysis program is reported. EXOSAT observations for four white dwarfs (WD1031-115, WD0004+330, WD1615-154, and WD0109-264) were obtained. Counting rates were unexpectedly low, indicating that these objects have a substantial amount of x-ray absorbing matter in their photosheres. In addition, soft x-ray pulsations characterized by a 9.25 minute cycle were discovered in the DA white dwarf V471 Tauri. A residual x-ray flux from the K dwarf companion can be seen during the white dwarf eclipse at orbital phase 0.0. Pronounced dips in the soft x-ray light curve occur at orbital phases 0.15, 0.18, and 0.85. Themore » dips may be correlated with the triangular Lagrangian points of the binary orbit. Smaller dips at phases near the eclipse may be associated with cool loops in the K star corona. Data for the white dwarf H1504+65 was also analyzed. This object is particularly unusual in that its photoshere is devoid of hydrogen and helium. Finally, existing data on the white dwarf Sirius B were analyzed to see what constraints from other data can be placed on the properties of this star. Interrelationships between radius, rotational velocity, and effective temperature were derived.« less
  • External reactor vessel cooling (ERVC) is a new severe accident management strategy that involves flooding the reactor cavity to submerge the reactor vessel in an attempt to cool core debris that has relocated to the vessel lower head. Advanced and existing light water reactors (LWRs) are considering ERVC as an accident management strategy for in-vessel retention (IVR) of relocated debris. In the probabilistic risk assessment (PRA) for the AP600 design, Westinghouse credits ERVC for preventing vessel failure during postulated severe accidents with successful reactor coolant system (RCS) depressurization and reactor cavity flooding. To support the Westinghouse position on IVR, DOEmore » contracted the University of California--Santa Barbara (UCSB) to produce the peer-reviewed report. To assist in the NRC`s evaluation of IVR of core melt by ex-vessel flooding of the AP6OO, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform: An in-depth critical review of the UCSB study and the model that UCSB used to assess ERVC effectiveness; An in-depth review of the UCSB study peer review comments and of UCSB`s resolution method to identify areas where technical concerns weren`t addressed; and An independent analysis effort to investigate the impact of residual concerns on the margins to failure and conclusions presented in the UCSB study. This report summarizes results from these tasks. As discussed in Sections 1.1 and 1.2, INEEL`s review of the UCSB study and peer reviewer comments suggested that additional analysis was needed to assess: (1) the integral impact of peer reviewer-suggested changes to input assumptions and uncertainties and (2) the challenge present by other credible debris configurations. Section 1.3 summarized the corresponding analysis approach developed by INEEL. The remainder of this report provides more detailed descriptions of analysis methodology, input assumptions, and results.« less