ML23255A239

Long Term Effects and Numerical Simulation of Radiolytic Gas, Non-Condensable Gas and Boron Transport for Small Modular Light Water Reactors

ML23255A239
Person / Time
Issue date:	09/11/2023
From:	Stephen Bajorek, Antonio Barrett, Syed Haider, Shanlai Lu, Joseph Staudenmeier, Carl Thurston, Peter Yarsky NRC/NRR/DSS/SNRB
To:	Office of Nuclear Reactor Regulation
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Long Term Effects and Numerical Simulation of Radiolytic Gas, Non-Condensable Gas and Boron Transport For Small Modular Light Water Reactors By S. Lu, C. Thurston, S. I. Haider, A. Barrett, J. Staudenmeier, P. Yarsky, S. Bajorek U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 5th meeting of NEA Expert Group on SMR. 3-5 Ottawa, Canada. October 3-4, 2023 Abstract In the past decade, nuclear industries and governments worldwide have developed interests in designing and deploying small modular reactors (SMRs) as viable energy source options to reduce carbon dioxide emissions and help resolve climate change issues. According to the International Atomic Energy Agency (IAEA), there are approximately 70 small modular reactor designs under investigation in 17 countries. These new SMR designs are at different stages of research, development, licensing, and commercialization. The collaborations coordinated by IAEA and driven by major industrial countries have put several SMR designs as the front runners in this phase of technology commercialization.

Some of these SMR designs under development have evolved from large light water reactor (LWR) designs and use light water as both a coolant and neutron moderator with passive gravity driven systems for normal operation and accident mitigation. The design goals of these passive safety systems are to maintain reactor core cooling for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any on-site or off-site power supply or operator intervention for a broad range of hypothetical accident scenarios including loss-of-coolant accidents (LOCA) and station blackout. These SMR designs normally have significantly higher coolant inventory/reactor power ratios than that of conventional large LWRs and are expected to keep the reactor core covered under a two-phase water level, or, at a minimum, preclude the fuel from experiencing prolonged heat-up for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The reliance of such passive SMR designs on gravity driven buoyancy flow and natural circulation for their long term Emergency Core Cooling System (ECCS) operation makes it possible to eliminate the need for higher cost, active pumping systems, simplify the containment design, and reduce the initial capital investment. The gravity driven ECCS designs can be reliable because of their reliance on inherent features and natural phenomena. However, the gravity driven buoyancy flow and natural circulation change the system mass and energy distribution during the long term cooling period after the initial transient. Depending on the design, the systems may become sensitive to some physical phenomena, such as (1) radiolytic gas generation and migration, (2) non-condensable gas effects, and (3) boric acid transport if used in the primary circuit to control reactivity. These phenomena were not considered to be significant concerns for most current LWR designs utilizing active ECCS, so they were not explicitly modeled in detail historically. Detailed evaluation of the accumulated effects of these phenomena may become necessary for passive LWR SMRs designs as part of design basis analyses.

In this paper, the authors summarized the information on these three phenomena and identified their potential safety implications for passive LWR SMRs. The state-of-the-art computer simulation tools commonly used by both the industry and regulatory agencies are discussed for their applications to evaluate the accumulated effects of these phenomena, including the challenges, limitations of these computer codes and future development needs.

operations of at least some of these new SMRs by Introduction 2030, reactor vendors are accelerating the development of these new design concepts. Among Small modular reactors (SMRs) with power outputs these SMRs under development, several design between 10 megawatts electric (MWe) and 300 MWe concepts evolved from the large light water power have been gaining attention across the world since reactors (LWR) currently in operation. The most 2010 because of their potential simple and passive significant improvement by these LWR SMRs is the safety features, modularity, significantly reduced use of gravity driven natural circulation during normal construction cost per MWe capacity, and promising operation and passive Emergency Core Cooling fast deployment schedule (Ref. [1,Error! Reference System (ECCS) during accidents or transients.

source not found.]). Approximately 70 small modular Because of the use of a passive ECCS and the reactor (SMR) designs are under investigation in 17 relatively large amount of water inventory inside a countries. All of them are at different stages of SMR reactor vessel, the reactor is expected to be free research, development, regulatory review and of fuel cladding heat up during a loss of coolant commercialization. With goals of commercial accident (LOCA) for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without operator

intervention or safety grade power supplies. The designs, because of their passive nature, must be gravity driven ECCS designs can be reliable because analyzed for significantly longer periods of time, up to of their reliance on inherent features and natural 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or beyond. These extended analysis times phenomena. However, the gravity driven buoyancy were not considered nor needed for the traditional flow and natural circulation change the system mass pumped ECCS injection systems. Therefore, many and energy distribution during the long term cooling new phenomena related to the effects of long-term period, e.g, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the initial transient. reactor cooling and generation of radiolytic gases Depending on the design, these kinds of passive LWR become important and must now be considered.

SMRs may become sensitive to three well known physical phenomena (1) radiolytic gas generation and In a PWR closed system, due to recombination, the migration, (2) non-condensable gas effects on system eventually attains equilibrium with respect to condensation heat transfer and (3) boric acid transport radiolytic decomposition during normal full power if used in the primary circuit to control the reactivity. operations. The concentration of gaseous products at These phenomena and their safety consequences have equilibrium is a function of the reactor power, water been well handled by current operating LWR pH and temperature, and the concentration approaches technologies. However, the accumulative distribution zero, with net production of only very small quantities and transport of these species may cause certain new of hydrogen dissolved in borated water. In a BWR safety implications which need to be appropriately open water/gas system, the product species are being accounted for to ensure the safe operation of these continuously removed via steam through the steam LWR SMRs. line and then inventory is replenished by new feedwater. In conditions where water is boiling This paper provides a highlight of the available public vigorously, H2 and O2 could be produced in domain information regarding the radiolytic gas stoichiometric portions based on applicable G generation and migration, non-condensable gas values for pure water exposed to primarily gamma effects, and boric acid transport. Based on the existing radiation. As a result, operating BWR plants have public domain information, the potential safety extensive off gas systems to process and control implications to LWR SMRs are identified. The release of radioactive effluents. Additionally, these numerical analysis capabilities in support of systems have been used to measure volumetric evaluating these phenomena are briefly reviewed and radiolytic gas production rate of hydrogen and oxygen the future development needs are discussed. in a proportion that is found to correlate with reactor power level. However, it is not known how well these Radiolytic Gas Generation and Transport measurements taken with traditional jet-pumped forced recirculation plants and whether they are Radiolysis of water occurs during both normal applicable to the newer BWR plant designs that are operation and accidents and involves the based on reactor vessel chimney with natural decomposition of water molecules by ionizing circulation flow. Natural circulation designs with radiation causing a molecular break sequence into lower recirculation rates, require a larger percentage of hydrogen peroxide, hydrogen radicals, and other new feedwater, as compared to the forced recirculation assorted oxygen compounds (Ref. [3]). The rate of designs.

hydrogen and oxygen generation is controlled by three factors: (1) decay heat energy or fission power (Ref. Most of the proposed new passive LWR SMR ECCS

[4]) (2) fraction of energy absorbed by the water, and designs heavily depend on the efficiency of steam (3) effective rate of hydrogen oxygen production per condensation to cool the reactor and/or containment unit of energy absorbed by the water, generally for the evaluation of design basis events including expressed as the product "G" value. accidents and anticipated operational occurrences (AOOs). The new passive ECCS and containment Traditionally, major concerns regarding hydrogen designs, with consideration of extended periods of generation were related to static or dynamic pressure operation, rely on condensing the steam generated loads from combustion in the containment that could during events to cool and depressurize the reactor initiate a release of radioactivity to the environment system and containment. These designs must consider and potential damage to safety-related equipment the generation of radiolytic gases under steady state (Ref.[5]). In recent LWR SMR designs, passive normal operation and post-accident decay heat, and systems and configurations are being employed to (1) how they may propagate to the condensing surfaces limit LOCA mass and energy releases (2) cool and used and their relative effect on the overall depressurize the reactor and (3) limit and reduce the condensation rate. The initial presence and migration containment pressurization. However, these newer of any additional non-condensable gases in the

containment atmosphere should also be considered incapable of crossing the gas-condensate interface, with respect to impact on condensation heat removal and accumulates near the condensing film and, thus, capability during the event. suppresses the vapor condensation. Hereunder the Depending on the geometry selected for the relevant details of the physical phenomenon are condensation surfaces, the propagation and provided.

accumulation of radiolytic gases can potentially collect in tubes or cold surfaces, impeding the Considering the component gases of the mixture to be condensation of steam and potentially significantly independent, condensation of one component vapor degrading the heat transfer efficiency of the ECCS will occur if the condensing surface temperature is component. Radiolytic gases are generally transported below the saturation temperature of the pure vapor at with steam, and after condensation, are left behind and its partial pressure in the mixture, i.e., the dew point.

collected around the condensation surfaces. If the During film condensation, condensation occurs at the ECCS is unable to efficiently remove heat from the interface of a liquid film on the wall. Due to the reactor due to the accumulation of non-condensable condensation process at the gas-condensate interface, gases, inadequate core cooling, containment there is a bulk movement of the gaseous mixture pressurization and a host of other problems can toward the wall, as if there were suction at the quickly occur. interface. As only the vapor is condensed, the NCG concentration is higher at the condensing film Transient radiolytic gas accumulation leading to interface than its value in the ambient. This, in turn, combustion of hydrogen is also a major safety concern reduces the partial pressure and saturation temperature since high pressures generated could breach of the vapor at the interface below the ambient values.

containment or damage other important safety-related At equilibrium, the NCG concentration at the interface equipment resulting in release of radioactivity. The is dictated by the balance between the NCG mass requirements for combustible gas control are well transfer away from the interface due to diffusion established in U.S Code 10 CFR 50.44 as long as the and/or convection and the vapor mass transfer from source and propagation of radiolytic hydrogen and the bulk to the interface to sustain the condensation.

oxygen generation and transport in steam are The resulting accumulation of NCG and the adequately modeled. Adequate measures as prescribed depression of temperature at the interface reduces the should be taken to mitigate or account for the condensation heat transfer rate across the liquid film accumulation of non-condensable gases to avoid below what would result for pure vapor under the same combustible limits. The hazards of hydrogen buildup conditions. The experimental database summarized on leading to deflagration and detonation has been the following figure from Ref.[7] clearly demonstrates realized worldwide as results of the TMI-2 and the reduction in condensation heat transfer coefficient Fukushima accidents. The generation of hydrogen with the increase of NCG mass fraction in the gas from radiolysis can have the same detrimental effect if mixture.

not adequately considered and controlled.

Therefore, it is imperative that the designs of these new passive systems adequately account for operational and post-accident generation of radiolytic gases, their initial presences, and their transport during AOOs and accidents.

Non-Condensable Gas Source and Transport In a gaseous mixture, condensation of one component vapor may occur in the presence of other noncondensable gas components. Vapor condensation in the presence of noncondensable gases (NCG) is of great interest to several engineering, industrial, and environmental applications. It is well established that the presence of NCG leads to a reduction in vapor Ref. [7] Figure 1(b) Variation of the condensation heat condensation and heat transfer. Sparrow et al. (Ref.[6 transfer coefficient with non-condensable gas mass

]) explained the mechanism by which NCG reduces fraction.

condensation. They showed that the vapor drawn to the condensing surface entrains with it the NCG that is

In nuclear applications, vapor-NCG mixtures exist in analytical solution (Ref. [6,7]) that is simple to use in the reactor containment and condensers following an engineering applications and provides results not far accident and the reactor design may account for their from the boundary layer analysis. The stagnant film active or passive removal. The NCG mixtures in model and the diffusion layer model were originally nuclear applications typically include radiolytic gases, formulated on a molar basis. Liao and Vierow (Ref.

nitrogen, and air. Condensation with NCG occurring [8]) developed a generalized diffusion layer model on in the containment and reactor systems plays a key role a mass basis that accounts for the effect of variable in designing the heat removal systems in LWR SMRs. mixture molecular weight across the diffusion layer The adverse effects of NCG on condensation and and fog formation effects on sensible heat. When saturation temperature depression can be minimized compared with a wide-ranging experimental database, with good condenser design practices that include the generalized model outperforms the one developed proper management and venting of the NCG. The by Peterson et al. (Ref. [7]). Under certain limiting degree of adverse impact of NCG on heat transfer also conditions, the generalized model reduces to the one depends on the dominant flow regime, among other developed by Peterson et al..

factors. Free convection flows are less efficient in sweeping the NCG away from the gas-condensate Boric Acid Transport Phenomenon interface that leads to a significant reduction in condenser performance, while the NCG effect is less Some LWR SMR designs continue the use of boric severe in forced convection flows that dampen the acid as a chemical shim. Boric acid, like all other formation of large NCG concentrations at the water-soluble substances, can experience interface. Some condensers are designed to perform precipitation, dilution, volatilization, and deposition in the presence of NCG. processes. The transport of boric acid throughout the primary reactor system or containment during an AOO A theoretical analysis of condensation from vapor- or accident may potentially cause precipitation in the NCG mixtures using a stagnant film model was first reactor core region and degrade the heat transfer due presented by Colburn and Hougen (Ref.[8]). In that to the blockages caused by the precipitated boric acid.

analysis, the overall thermal resistance between the The resulting dilution of boric acid concentration in condenser tube wall and the bulk vapor-NCG mixture the reactor core region may also introduce excessive is the sum of the thermal resistances of the condensate reactivity, which, if not controlled properly, could film and the NCG boundary layer. The heat transfer cause the reactor returning to power or experiencing through the gas boundary layer includes the sensible uncontrollable power excursions.

and latent parts. The latent heat transfer is evaluated by using a stagnant film model combined with the heat Boric acid is volatile during the boiling process of its and mass transfer analogy. These calculations involve water solution (Ref. [17]). Although the volatilization the convergence of the unknown temperature and rate is relatively small and the boric acid concentration NCG mole fraction at the interface through matching in the steam is of the order of 10-3 ~ 10-2 of the liquid the condensate mass flux and the heat transfer through solution concentration, the accumulative removal of the condensate film. Although Colburn and Hougen the boric acid from the water solution over a long did not derive an expression for the NCG boundary period of time could significantly affect the reactor layer conductance, they suggested that the mass core boric acid concentration distribution. The transfer coefficient in the stagnant film model could be volatized boric acid is carried away from the boiling converted to a heat transfer coefficient by using the boric acid solution. It then either deposits on the solid Clausius-Clapeyron equation, for saturated mixtures. surface along the steam flowing path or redissolves Using this principle of heat and mass transfer analogy, into the solution during steam condensation. The boric Peterson (Ref. [9] ) proposed the diffusion layer model acid deposition process can remove the available mass to calculate the latent heat transfer. The theoretical of boric acid from the primary circulation system, basis underlying the heat and mass transfer analogy is which would affect the long term boric acid that the conservation equations for mass, momentum, distribution in the active core region. The boric acid and energy have similar mathematical forms. deposits in the primary system could impact the operability of the reactor vessel internals, including, Both the stagnant film model (Ref.[7]) and diffusion but not limited to control rod drive systems, layer model (Ref. [9]) neglect the longitudinal flow instrumentation, and measurement devices.

acceleration that can be accounted for by numerically solving the boundary layer equations for the gas and After the Three Mile Island Unit 2 accident in 1979, liquid phases (Ref.[9]). Neglecting the longitudinal extensive efforts have been made world wide by flow acceleration makes it possible to obtain an nuclear industries, research institutes, and universities

to study the boric acid transport phenomena which trapped in the primary system pump loop seals into the have significant safety implications to PWR designs. core. The sudden core boron dilution could cause Some of the latest studies on boric acid precipitation recriticality and a return to power. G. Jimenez et al.

can be found in Ref. [11]. The crystallization and analyzed a PWR boron dilution transient using the deposition of boric acid were clearly visible during the coupled neutronics and thermal-hydraulics codes experiment as shown in the circulated areas in the DYN3D/FLOCAL. Fig.2 of Ref. [15] showed the following figure. reactor core fuel and control rod loading.

Ref. [11] Fig. 9. Experimental results (end of the test).

The boric acid solubility limit is dependent on the temperature and pressure of the aqueous solution. The lower the temperature, the less the solubility. A literature review of the operating and transfer conditions examined by Ref. [12] include temperatures between 13 C (McLeskey, 2008) and 45 C (Fondeur, 2007); and concentrations from 0 to 3000 ppm in nitric acid as well as exposure of small amounts Ref. [15] Figure. 2. Core configuration and of entrained boric acid in the organic phase to the distribution of control rod groups.

sodium hydroxide caustic wash stream.

Assuming 18 m3 of diluted water slug being injected During the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of post LOCA long term cooling, into the reactor core, the coupled computer codes the core region of a SMR PWR may experience a predicted a power increase by 1010 times within 5.75 reduction in solubility due to the continuous decline of seconds as shown in Figure 4 of Ref. [15]. The power the system temperature and pressure. Once the boric peak was numerically predicted to be stabilized by the acid solution concentration exceeds the solubility negative reactivity feedback due to the Doppler effect.

limit, white boric acid crystal precipitate in the In reality, if a PWR core experiences a 1010 power solution and could accumulate in the core region and increase starting from the remaining subcritical fission cause fuel flow path blockages. Therefore, the core power due to the decay of delay neutron precursors, region of the boron concentration of a SMR PWR the reactor core fuel enthalpy and thermal limits could should be evaluated to show no precipitation is exceed the allowable values and the reactor core could possible during the long term cooling phase of the post experience severe damage in extreme circumstances.

LOCA transient.

Boron dilution could happen in a PWR core due to boric acid loss from the fluid region, potentially caused by entrainment, volatilization, and mixing with in-coming pure or highly diluted water. As studied by Argonne National Laboratory in 1995 Ref. [14], the most significant boron dilution event for operating PWRs was found to be the restart of reactor coolant pumps which quickly transport the diluted condensate

temperature and void fraction. The autoclave data collected by S. Bohlke et al showed much lower volatility than they measured from their test rig with a void fraction greater than 60%. Other autoclave data showed higher volatility for the given temperature Ref. [17]. These different research papers showed that the volatility is normally small within the range of 0.25 ~ 2% and is sensitive to the local fluid conditions at the test rigs. The temperature, local void fraction, and flow regime affect the volatility value. Even at the same test rig, the measured volatility values vary when local void fraction or flow velocity changes.

Ref. [15] Fig. 4. Slug 1 fission power evolution. Therefore, if the accumulative loss of boric acid from A natural circulation PWR, without a pump, however, the core fluid region due to volatility needs to be does not have a mechanism for such a sudden transport evaluated for certain reactor design configurations, of a fresh water slug into the core. If the coolant in the considerations should be given to the actual core reactor lower plenum or downcomer is diluted, e.g, as configuration and system parameters.

a result of evaporation and condensation, this could lead to a slow dilution of the coolant in the core region, depending on mixing at the core inlet. The reactor could reach re-criticality if there is a semi-continuous flow of diluted coolant from the downcomer and lower plenum to the core. The additional fission power could escalate the vaporization and dilution causing either flow instability or sudden level changes due to depressurization, which could lead to a sudden in-flux of diluted coolant into the core and cause the similar power surge as shown in Figure 4 of Reference [15].

Therefore, reactor vendors and regulatory agencies should pay attention to the transport mechanism of boric acid during reactor normal operation, AOOs, design basis accidents, and post accidents mitigation phases. Proper boric acid mixing in the reactor lower plenum and downcomer regions can help to avoid boron dilution induced re-criticality and the potential Ref. [13] Fig. 4. Volatility of boron out of boiling for power excursions. pentaborate solution with a defined boron content (cboron = 1.3 g/L) at various void fractions-The volatilization of boric acid can slowly remove the measurements with BORAN (BWR ATWS Tests) boric acid from the reactor core region. The accumulated removal over a long period of time can Following volatilization, boric acid mixes with the be significant. The volatilization happens at the steam and may either deposit on the solid surface or interface between the vapor and the fluid. The redissolve in the water during the steam condensation volatilization rate is highly dependent on the vapor process. The gradual accumulation of deposits on the generation process in the core region. The flow solid surface may affect the operability of some regimes, the local liquid boric acid concentrations, the reactor vessel internals, valves or measurement vapor bubble rising history, and the system parameters devices. These specific situations need to be evaluated such as pressure and temperature influence the on a case-by-case basis to determine the safety volatilization process. Therefore, there has not been an implications if there are any.

universal correlation developed to bound all scenarios.

S. Bohlke et al., conducted a series of boron Numerical Simulations of Radiolytic Gas, Non-volatilization tests and data analysis using a test Condensable Gas, and Boric Acid Transport facility simulating the BWR core boiling processes during a BWR Anticipated Transients Without Scram Radiolytic gas, non-condensable gas, and boric acid in event with an injection from the standby liquid control the primary system of light water SMRs are normally system Ref. [13]. As shown in Figure 3 of Ref. [13], carried by either liquid water or vapor. Because of the measured boron volatility increases with higher their very low concentrations for most of the scenarios,

these species do not normally impact the transport of mixture. The static quality, X, is likewise defined as their carrying media except for the case inside a the mass fraction based on the mass of the vapor/gas containment full of non-condensable gas during phase.

normal operations. Therefore, the movement of these species is largely determined by their media transport, Similar to the typical non-condensable gas mass i.e., liquid and vapor flow. The numerical simulation conservation approach, an Eulerian boron tracking of liquid and vapor flow in a LWR has been the model is normally used in these system codes to primary focus of safety analysis code development simulate the transport of a dissolved component during the past 70 years. These thermal-hydraulic (solute) in the liquid phase (solvent). The solution is safety analysis codes are used to model the primary assumed to be sufficiently dilute that the following and often the secondary sides of nuclear plants and are assumptions are generally made:

called upon to simulate the liquid and vapor flow in a wide range of accident scenarios. The codes currently

• Liquid (solvent) properties are not altered by in use include but not limited to TRAC [18], RELAP the presence of the solute.

[19,28], COBRA/TRAC [20], TRACE [21], TRACG

• Solute is transported only in the liquid phase

[24], CATHARE [25], ATHLET [23] and MARS [25]

for primary systems and GOTHIC [27] and MELCOR (solvent) and at the velocity of the liquid

[26] for containment analyses. Both GOTHIC and phase (solvent).

MELCOR codes have been developed to model

• Energy transported by the solute is multiphase, multicomponent fluid flow for performing negligible.

both containment design basis accident (DBA)

• Inertia of the solute is negligible.

analyses, severe accident analyses and equipment qualification analyses. Based on these assumptions, the typical mass conservation for boric acid in the liquid can be the The general approach for including the non- following:

condensable component consists of assuming that all the non-condensable component present in the vapor-gas mixture moves with the same velocity and has the same temperature as the vapor phase. Based on these (Equation 3.1-51 of Ref. [28])

assumptions, an additional mass conservation equation is typically used for the total non- where the spatial boron density, b, is defined as condensable component including radiolytic gas in the vapor/gas phase (Equation 3.1-52 of Ref. [28])

Cb is the concentration of boron, X is the static quality (Equation 3.1-39 of Ref. [28]) and A is the 1-D flow area, f is the liquid density.

Xn is the total non-condensable mass fraction in the The inclusions of non-condensable gas and boric acid vapor/gas phase in the mass conservation equations for the two-phase models are typical among the system codes. Finite-volume and finite difference schemes are then applied to solve these two-phase flow conservation equations.

Most system analysis codes model relatively large spatial regions and provide only averaged behavior of conditions within those regions. This is primarily what distinguishes nuclear systems codes from codes (Equation 3.1-40 of Ref. [28]) that perform Computational Fluid Dynamics (CFD, eg, FLUENT code Ref.[29]). A nuclear system code Mni is the mass of i-th non-condensable gas; Mn is the attempts to determine the distribution of mass and total mass of non-condensable gas in the vapor/gas energy throughout an entire hydraulic network using phase; Ms is the mass of vapor in the vapor/gas phase; large computational nodes, where the highly detailed N is the number of non-condensable. The thermal nodalization can be achieved by CFD to analyze the properties of the vapor/gas phase (subscript g) are localized turbulence, vortexes and thermal mixing.

mixture properties of the vapor/non-condensable

Systems thermal-hydraulic codes are subject to accumulated effects may become necessary for significant uncertainties because of their dependence passive LWR SMR designs. Although significant on large spatial regions and the use of semi-empirical studies have been done in the past 70 years to develop models and correlations to simulate the average system codes and computational fluid dynamic codes behavior of conditions within those large regions. which have some capabilities to simulate these Most of them have been developed to model fast and phenomena, improvements on multi-species tracking short system transients, such as a loss of coolant and multiphase flow simulation with finer accident (LOCA). Therefore, they have greater nodalization are necessary to avoid expensive design prediction uncertainties to deal with gravity or specific tests.

buoyancy driven flow conditions which are important to passive LWR SMR technology. The CFD codes, on the other hand, can simulate boundary-layer mixing and turbulence phenomena. Both CFD codes and Disclaimer system codes are subject to large uncertainties when modeling two-phase flow in a buoyancy force The views expressed herein are those of the authors dominant fluid field, where all models for flow regime and do not necessarily represent an official position of interfacial phenomena and flow pattern transitions are the U.S Nuclear Regulatory Commission. This empirical. Quite often, models and correlations for material is declared as a work of the U.S. Government two-phase flow are derived from experiments with and is not subject to copyright protection in the United small diameter pipes which do not readily scale to the States.

large flow region in a reactor vessel internal fluid field.

Thermal stratification in large pools depends on Acknowledgement turbulence and temperature gradients to mix fluids.

System codes do not generally have the capability to The confirmatory analysis work completed by U.S.

model the turbulent mixing and thermal stratification Nuclear Regulatory Commission subject experts, P.

due to the lack of detailed nodalization or local Lien, S. Campbell, A. Ireland and C. Boyd has greatly phenomenon modeling capability. shed light on the key view points of this paper.

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