Modification of Filtered Air Intake Flowrate to Improve Control Room Habitability

2018 ◽  
Author(s):  
Xinjian Liu ◽  
Weipeng Shu ◽  
Mengxi Wang
Keyword(s):  
Extended Period ◽  
Leakage Rate ◽  
Design Parameters ◽  
Source Terms ◽  
Control Room ◽  
Limit Operator ◽  
Radioactive Aerosols ◽  
Loss Of Coolant Accident ◽  
Operator Dose ◽  
Accident Conditions

Control room habitability (CRH) shall be maintained to provide adequate protection for control room operators, such that they can remain in the control room envelope (CRE) safely for an extended period and thus control the nuclear facility during normal and accident conditions. A critical objective of CRH systems is to limit operator doses and/or exposure to toxic gases. The CRH systems does this by the combination of the intake of filtered air, isolation of outside air, recirculation systems and etc. Among the parameters determining radioactivity in a control room (in proportion to radiation doses of operators), intake flowrate of filtered air is an important one. For different types of accident source terms, the evolution of operator doses in a control room versus intake flowrate were analyzed in this paper. It turns out that the increase of intake flowrate results in larger operator doses when inert radioactive gases are the dominant radioactive substances. On the contrary, increasing intake flowrate does good to lower the irradiation level of control room operators when radioactive aerosols dominate the source terms. The rationality behind this fact was interpreted in detail in this paper, with special attention paid to the unfiltered in-leakage rate. It can be inferred that an optimal intake flowrate probably exists leading to the minimum operator dose under an actual accident condition. This paper then performed a calculation analysis based on design parameters and source terms of design basis accident of LOCA (a large break loss of coolant accident) accident. The evolution of operator dose was found to be a U-curve versus increasing intake flowrate, which proved the existence of the abovementioned optimal intake flowrate of filtered air for CRH systems. Furthermore, the sensitivity analysis of intake flowrate was carried out to study the effects of unfiltered in-leakage rate and filtered recirculation. This study indicates that intake flowrate of filtered air can significantly influence the CRH. For different accidents, the intake flowrate should be properly modified rather than set as a fixed value. To optimize the radiological habitability of control rooms, the effects of unfiltered in-leakage must be taken into consideration. Besides, filtered recirculation is an effective way to control radiation exposure caused by iodine and radioactive aerosols.

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

An Experimental Study on Head Loss of Prototypical Fibrous Debris Beds During Loss-of-Coolant Accident Conditions

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Amir Ali ◽  
Edward D. Blandford
Keyword(s):  
Safety Issue ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Head Loss ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Debris Accumulation ◽  
Loss Of Coolant Accident ◽  
Accident Conditions ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (NRC) initiated a generic safety issue (GSI-191) assessing debris accumulation and resultant chemical effects on pressurized water reactor (PWR) sump performance. GSI-191 has been investigated using reduced-scale separate-effects testing and integral-effects testing facilities. These experiments focused on developing a procedure to generate prototypical debris beds that provide stable and reproducible conventional head loss (CHL). These beds also have the ability to filter out chemical precipitates resulting in chemical head loss. The newly developed procedure presented in this paper is used to generate debris beds with different particulate to fiber ratios (η). Results from this experimental investigation show that the prepared beds can provide reproducible CHL for different η in a single and multivertical loops facility within ±7% under the same flow conditions. The measured CHL values are consistent with the predicted values using the NUREG-6224 correlation. Also, the results showed that the prepared debris beds following the proposed procedure are capable of detecting standard aluminum and calcium precipitates, and the head loss increase (chemical head loss) was measured and reported in this paper.

Experimental Study of Moderator Circulation in CANDU6 Calandria Tank

2014 ◽  
Author(s):  
Hyoung Tae Kim ◽  
Han Seo ◽  
Sunghyuk Im ◽  
Bo Wook Rhee ◽  
Jae Eun Cha
Keyword(s):  
Heat Sink ◽  
Optical Measurement ◽  
Cooling System ◽  
Safety Concern ◽  
Measurement Data ◽  
Validation Data ◽  
3 Dimensional ◽  
Loss Of Coolant Accident ◽  
Piv Measurement ◽  
Accident Conditions

As a CANDU6 reactor has a high pressure primary cooling system and an independently cooled moderator system, the moderator in the calandria would act as a supplementary heat sink during a loss of coolant accident (LOCA) if the primary cooling and emergency coolant injection systems fail to remove the decay heat from the fuel. For the safety concern it is required to predict the 3-dimensional velocity and temperature distribution of moderator fluid to confirm the effectiveness of moderator heat sink. Korea Atomic Energy Research Institute (KAERI) is carrying out a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV). In the present work the PIV technique is used to measure the velocity distributions in the scaled moderator tank of MCT under iso-thermal test conditions. The preliminary PIV measurement data are obtained and compared with CFX code predictions.

Severe Fuel Damage Experiments With Advanced Cladding Materials to be Performed in the QUENCH Facility (QUENCH-ACM)

2008 ◽  
Author(s):  
L. Sepold ◽  
M. Große ◽  
M. Steinbru¨ck ◽  
J. Stuckert
Keyword(s):  
Nuclear Power ◽  
Cooling Water ◽  
Test Series ◽  
Experimental Program ◽  
Limited Data ◽  
Standard Type ◽  
Operational Conditions ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Accident Conditions

The QUENCH out-of-pile experiments are part of the Severe Fuel Damage (SFD) program at the Karlsruhe Research Center. They are to investigate the hydrogen source term that results from reflooding an uncovered core of a Light-Water Reactor (LWR) with emergency cooling water. In the QUENCH experimental program Zircaloy-4 was used as standard-type material for rod cladding and grid spacer. Up to the end of 2007, 12 QUENCH experiments have been performed with this type of cladding; two test bundles contained B4C and one AgInCd absorber. One experiment (QUENCH-12) was conducted with Zr1%Nb cladding (VVER-type). Due to the niobium-bearing cladding, the VVER-type test QUENCH-12 could be regarded as a precursor for the upcoming program “QUENCH-ACM” with advanced cladding materials, i.e. M5, Duplex, ZIRLO, to be tested under SFD or BDBA (beyond design basis accident) conditions. These materials were developed for longer operation times in nuclear power reactors and extended burnup. They are optimized regarding their corrosion behavior under operational conditions and were also tested for LOCA (loss of coolant accident) and RIA (reactivity-initiated accident) conditions by the manufacturers. However, there are only very limited data available on the behavior of the new alloys in the SFD/BDBA temperature range, i.e. above 1500 K. The QUENCH-ACM test series has been defined with three experiments, i.e. QUENCH-14 through QUENCH-16. As in the Zircaloy-4 experiments, fuel is represented by ZrO2 pellets. Also, the test section instrumentation will be as usual with thermocouples attached to the cladding, shroud, and cooling jacket at elevations between −50 mm and 1350 mm. The QUENCH-ACM test series is scheduled to be performed in the period of 2008–2010. Test matrix and test bundle arrangements are presented in this paper.

Simulation of Gas Stratification Build-Up in the Containment Under Severe Accident Conditions

2014 ◽  
Author(s):  
Michele Andreani
Keyword(s):  
Structural Integrity ◽  
Computer Code ◽  
Severe Accident ◽  
Test Sequence ◽  
Small Scale ◽  
High Concentration ◽  
Initial Transient ◽  
Loss Of Coolant Accident ◽  
Computer Codes ◽  
Accident Conditions

The presence of hydrogen stratification in a NPP containment in the case of a severe accident is a source of concern, as pockets of the gas in high concentration could lead to a deflagration or detonation risk, which might challenge the containment structural integrity. These issues, as well as the capability of various computer codes to predict the evolution of a representative accident, are addressed in the coordinated projects ERCOSAM of the 7th EURATOM FWP and the project SAMARA sponsored by ROSATOM. The projects aim to establish whether in a test sequence representative of a severe accident in a LWR hydrogen stratification can be established during the initial transient following a loss of coolant accident (LOCA) and whether and how this stratification can be broken down by the operation of Severe Accident Management systems (SAMs): sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Experiments with helium (as simulant of hydrogen) have been performed at “small scale” in TOSQAN (IRSN, Saclay), and “medium scale” in the MISTRA (CEA, Saclay), PANDA (PSI, Villigen) and SPOT ((JSC “Afrikantov OKBM”, Nizhny Novgorod) facilities. The present paper presents the analysis of the initial transient of some tests in the PANDA, TOSQAN and SPOT facilities using the GOTHIC computer code. The work therefore addresses the capability of the code and a relatively coarse mesh to simulate the pressurisation and build-up of steam and helium stratification for conditions representative of a postulated severe accident scenario, properly scaled to the various facilities. The prediction of the pressurisation is excellent, and the position of the gas concentration stratification front at the end of the steam and helium releases is generally well captured.

STANDARDIZING BOOM TEST PROCEDURES1

1985 ◽  
Vol 1985 (1) ◽  
pp. 41-45
Author(s):  
M. Borst ◽  
H. W. Lichte
Keyword(s):  
Standardized Test ◽  
Extended Period ◽  
Open Water ◽  
Fold Increase ◽  
Weather Conditions ◽  
Technical Committee ◽  
Design Parameters ◽  
Sea State ◽  
Water Testing ◽  
Wide Range

ABSTRACT The Interagency Technical Committee (OITC) of the U.S. Environmental Protection Agency's (EPA) Oil and Hazardous Materials Simulated Environmental Test Tank (OHMSETT) facility sponsored a combined series of in-tank and open-water tests on five booms. The booms selected cover the wide range of sizes and design parameters often considered appropriate for spill control. The tests were conducted at the OHMSETT facility and in nearby Sandy Hook Bay between November 1983 and May 1984. The in-tank tests measured the oil-holding performance of the boom; the open-water tests demonstrated wave conformance and survivability. The objective was to correlate the two sets of data. The five in-tank tests for each boom used a standardized test matrix for oil-holding ability. Wave conformance and endurance were observed under controlled conditions. The results of this testing compared favorably with historical tests performed in the test task. The five booms were deployed in the bay over an extended period. The length of deployment varied from 14 to 27 days depending on the local weather conditions. The booms were observed and videotaped to document approximate sea-state limits of the booms. The booms deployed in the bay were approximately 500 ft long. The tank tests used 100 ft lengths. Wave conformance in the two tests was similar. The five-fold increase in boom length made any lack of conformance more obvious, however. The data derived from the open-water testing were based primarily on visual observations and sea-state estimates. The in-tank endurance tests did not correlate as well as expected with the open-water testing. Determining the deterioration of a boom during long-term deployment by in-tank testing was discounted by this program. This paper documents the results of the initial in-tank and open-water tests, emphasizing the techniques of testing, and outlines plans for the future tests.

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