Containment Sump Research Articles

Simulation and calibration of radiation monitoring of nuclear power plant containment sump waste liquid

Waste liquid stored in the containment sumps of nuclear power plants may contain a variety of radionuclides. Real-time monitoring of containment sump waste liquid can ensure that accidents, such as leakage of cooling water, can be avoided. This paper presents the design of a radioactive monitoring system for waste liquid in a containment sump. The detector and the lead-shield in the measurement unit are optimized through Monte Carlo simulations. Experimental verification showed that the background count rate of the measurement chamber in the system was 418.3 cps, and the detection limit of the detection system was 3.01 Bq/L. Distinct gamma-ray characteristic peaks were also observable, demonstrating the system's ability to identify radioactive nuclides in the waste.

Assessment of spray system capabilities in reduction of containment pressure and deposition of fission products during a severe accident in VVER-1000

In this study, we have investigated the spray system's capabilities to reduce fission products in the containment atmosphere and control the containment pressure in a VVER-1000 nuclear power plant. Large Break Loss of Coolant Accident (LB-LOCA) without access to the active part of the Emergency Core Cooling System (ECCS), as a worst-case severe accident, has been investigated as a case study. Only containment is simulated via the MELCOR code, while leakage of various materials from the nuclear power plant's core and the primary circuit is considered as the boundary conditions. Following the Fukushima accident in 2011, many nuclear safety regulatory organizations focused on using portable equipment that is essential for decay heat removal in nuclear power plants during severe accidents. Based on the literature studies, the accident management scenarios have changed recently since severe accidents inside nuclear power plants have worried many scientists and the public. Evaluating the current spray system in a severe condition based on which this system was not designed is vital and novel. We considered that the spray system would be fully restored and perform its safety function using the existing portable equipment to depressurize the containment and deposition of the fission products. To calculate the maximum allowed delay time of the spray system recovery, the accident was modeled without the operation of the spray system while a safety criterion was considered to avoid reaching the ultimate pressure of containment. For a more detailed evaluation of the spray system, two different operating modes including the activation of one spray train and the activation of both spray trains have been selected. According to the obtained results, if the spray system is revived before 4,000s from the beginning of the accident, this system would be able to reasonably mitigate the accident consequences. Moreover, without the need for any external water supply source for the spray system, this system can be permanently fed from the containment sump for at least 800,000s after the initiation of the accident, requiring the relevant heat exchanger of the spray system to be recovered for this period.

External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krško

The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krško applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant’s site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant’s post-Fukushima improvements and adequacy with plant’s procedures for severe accident mitigation. Conclusion is that NPP Krško could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance’s (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the lost core geometry.

Organic Telluride Formation from Paint Solvents Under Gamma Irradiation

The interactions between tellurium and organic material during a nuclear reactor accident are critical to source term estimations because of the possible formation of volatile species. Reactions taking place in the containment sump are of interest since these can lead to re-volatilization and increase the fission product source term. This paper presents results from experiments investigating the interaction of tellurium dioxide with three paint solvents—texanol ester, methyl isobutyl ketone, and toluene—under containment sump conditions. The experiments were performed by irradiating a mixed solution of tellurium dioxide and paint solvents at a dose rate of 4 kGy/h up to 300 kGy. The resulting samples were analyzed for tellurium concentration and speciation. Tellurium(IV) was found to reduce to metallic tellurium under irradiation when paint solvents were present. More importantly, several volatile organic tellurides were identified in the irradiated samples, which suggests that tellurium can form volatile species in sump conditions when in contact with dissolved paint solvents. This paper provides novel evidence of organic telluride formation in the sump and raises further interest in tellurium chemistry during a severe nuclear reactor accident.

VVER-440/V213 long-term containment pressurization during severe accident

The contribution deals with the performance of a new containment heat removal system that is intended to be installed in near future at Dukovany NPP in Czech Republic. This NPP is equipped with 4 units of VVER-440/V213 reactors. During the last two decades these units were systematically backfitted with new safety systems in order to cope with consequences of severe accidents. These safety systems were designed especially for the 4th level of defence in depth and thus they are – as much as possible – independent from the original plant systems. Namely, installation of a new depressurisation line is currently under preparation in order to avoid high-pressure core melting scenarios. In-vessel corium retention via external reactor vessel cooling (“reactor cavity flooding”) was adopted as a cornerstone strategy of severe accident management for all VVER-440/V213 reactors operated in central Europe. Regarding the hydrogen issue, large capacity passive autocatalytic recombiners were installed in containment. The prevention of containment pressurisation in the case of loss of ultimate heat sink represents the most important point of remaining safety issues. At first, special features of VVER-440/V213 containment equipped with bubbler condenser pressure suppression system are briefly described in presented paper. Than the mechanism of containment pressurisation due to steaming as a consequence of in-vessel retention and possible solutions for heat removal from the containment are discussed. The final design chosen at Dukovany NPP is based on the heat exchanger and turbo pump located inside the containment that are cooled and driven by external cooling water. External cooling circuit will be connected during the accident via special inlet/outlet nozzles installed at containment boundary. Main advantage of such containment cooling system is full independence on other plant systems (including essential service water), the fact that radioactive coolant is not taken beyond the containment boundary and all active components of the system (mobile pump, isolation valves) are located outside the containment and, thus, are easily accessible. The performance of this containment heat removal system during severe accidents were analysed using the ASTEC code and the most important results are presented. From the obtained results it follows that there is sufficient time margin for assembling the external cooling loop. Thus, the containment cooling can be assured long before the containment ultimate pressure is reached or before the coolant level in containment sump drops below critical level, when the external reactor cooling can be lost. After start of the heat removal system, the system is capable to prevent further containment pressurisation, increase the coolant level in containment sump and decrease the pressure to atmospheric or even sub-atmospheric level.

Tellurium Behavior in the Containment Sump: Dissolution, Redox, and Radiolysis Effects

In the event of a severe nuclear accident, one major concern is the release of radioactive material into the environment causing potential exposure of the general public to radiation. Among the volatile radionuclides are a range of tellurium isotopes. Due to the radioactivity and the volatility of tellurium, it has to be taken into account when assessing the overall effects of an accident. The interest in tellurium is not limited only to its release but also to the fact that some tellurium isotopes decay to iodine, and thus affect the iodine release behavior. The release and transport behavior of tellurium has been investigated over the past decades, however, the aqueous chemistry of tellurium in the complex containment sump system is still unclear. This study presents the behavior of tellurium dioxide in simplified containment sump conditions in relation to dissolution, redox reactions, and interactions with water radiolysis products. The results indicate that radiolysis products have a significant effect on tellurium chemistry in both a reducing and oxidizing manner depending on the solution composition. The redox reactions also affect the solubility of tellurium. The results show that the current information used to assess tellurium source term needs to be reevaluated for both severe accident management and for code validation purposes.

Review of the potential sources of organic iodides in a NPP containment during a severe accident and remaining uncertainties

In case of a severe accident in a nuclear power plant, organic iodides (RI) are one of the volatile iodine species that can be released in the environment thus possibly contributing to the radiotoxicity by inhalation. The formation and decomposition reactions of organic iodides have been studied through various experimental programs whose objectives were to understand their behavior and develop phenomenological models able to predict their volatility. Four potential sources of RI formation have been reviewed: (1) the reactor coolant system (RCS), (2) the containment sump, (3) the paints in the containment and (4) the containment gaseous phase. A review of the knowledge, chemical mechanisms leading to RI formation and decomposition and the remaining uncertainties is performed in order to recommend remaining needed experimental investigations. It has been identified that significant uncertainties remain on RI formation in the gaseous phase through reaction of iodine with volatile organic compounds (VOC, like methane) and that additional data should be obtained to improve its modeling as it can be a significant RI source during days after the beginning of the accident.

MELCOR and GOTHIC Analyses of a Large Dry PWR Containment to Support Resolution of GSI-191

The thermal-hydraulic response of large dry pressurized water reactor containments under loss-of-coolant-accident conditions “particularly with respect to containment pressure and sump pool temperature” is crucial for risk-informed decision making about Generic Safety Issue 191. Texas A&M University has developed models with several computer codes including MELCOR and GOTHIC to model such scenarios.MELCOR is a best-estimate thermal-hydraulic and severe accident code created and actively maintained by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. GOTHIC is a thermal-hydraulic software package meant for design, licensing, and safety calculations for, among other systems, nuclear power plant containments. It was developed and is maintained by Numerical Applications Inc. for the Electronic Power Research Institute.The overarching goal of the analyses presented here is twofold: (1) produce best-estimate time profiles of sump pool temperature under double-ended guillotine-break conditions with MELCOR and GOTHIC and (2) investigate differences between the MELCOR and GOTHIC code results via a sensitivity study. The sump pool temperature was selected as a key parameter to compare because it has direct implications for sump pool chemistry, residual heat removal during recirculation, and pressure drop across sump screens.Aspects of the MELCOR and GOTHIC modeling strategies are discussed, and best estimates of the containment thermal-hydraulic response are presented. There are significant disagreements between code predictions. Hypotheses to explain the differences are tested through a comparative code sensitivity study. In this context, “sensitivity” refers to how containment thermal hydraulics respond to differences in code inputs or code phenomenological models. Sensitivity calculations are performed to exclude, individually, the model effects on comparative thermal-hydraulic responses of containment fan coolers, containment sprays, thermal surface condensation/films, and break source definition. Calculations are also performed with multiple models excluded. Using containment sump pool temperature as an indicator, the most impactful physics in terms of code agreement are those of thermal surfaces (condensation, film phenomena) whereas fan cooler models have a minimal effect. Containment spray exclusion results in disagreement in parts of the event sequence, while break source definition and/or break effluent flashing models lead to disagreement.

Iodine Benchmarks in the SARNET Network of Excellence

Accurate calculation of iodine behavior in the containment is very important in determining the potential radioactive release to the environment in light water reactor severe accidents (SAs). Of particular significance is the behavior of gas phase iodine, particularly organic iodine, which is difficult to remove by filtration, e.g., in containment venting systems. Iodine behavior is closely linked with the containment thermal hydraulics, which have a major influence on the distribution of iodine throughout the containment atmosphere and sump. In the European 7th Framework SARNET project, European Commission (EC) cofunded from 2007 to 2013, SA modeling code capability was assessed through two integral benchmarks. In the first, the basis was the German THAI Iod-11/12 tests, where molecular iodine transport with atmospheric flows and iodine interactions with steel surfaces were emphasized. In the second, data from the international Phébus FPT3 test were used, where all aspects of SAs were studied from core degradation, fission product (FP) release, circuit transport/deposition, and containment behavior using realistic FP sources. Thermal hydraulics in the containment were simpler, being well-mixed, and radiolytic interactions of iodine, e.g., with painted surfaces, were studied. These interactions may be an important source of organic iodine in the containment atmosphere. The two benchmarks are thus complementary. In the FPT3 exercise, the calculations could predict the containment thermal hydraulic conditions fairly well. For the more detailed data from THAI, differences were noted for atmospheric flows and relative humidities, outside experimental uncertainties, affecting iodine behavior. The FPT3 iodine results themselves showed a spread in calculated results outside data uncertainties, indicating the need for model improvements in this area, e.g., for radiolytic interaction of iodine with paint. Experimental programs to generate the necessary data needed for code improvement have been recently completed, e.g., in the OECD/THAI, THAI2, BIP, and BIP2 projects, or are in progress, in OECD/STEM and EC/PASSAM. When model improvements have been made, repeat benchmarks are planned to check progress toward code convergence with experimental data, e.g., under the aegis of the new NUGENIA association of which SARNET now forms a part.

Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly

Under post loss-of-coolant accident (LOCA) conditions, it is postulated that debris can be generated and transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, in-vessel downstream effect tests for the advanced power reactor (APR) 1400 were performed. Fibrous debris is the most crucial material in terms of causing pressure drops, and was prepared in this study to satisfy the fiber length distribution obtained through a strainer bypass test. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of water chemistry and fiber length distribution. The pressure drops with debris laden pure water were substantially less than those with debris laden ordinary tap water. The experiment with fiber length distribution suggested by WCAP-16793 showed lower pressure drops than those with the APR1400 specific fiber length distribution. All the experimental results showed that the pressure drops in the mock-up fuel assembly were less than the available driving head at each LOCA scenario.

원전의 내환경기기검증 화학환경 및 핵분열생성물 제거능력 평가

원자력발전소는 냉각재상실사고(LOCA)와 같은 과도상태시 pH 조절을 통해 격납건물의 핵분열생성물(요오드) 제거 능력을 유지한다. 이와 더불어 격납건물 내부의 스테인레스강 기기들의 응력부식균열(Stress Corrosion Cracking)을 방지하고 알루미늄 또는 아연 부식에 의한 수소생성을 최소화할 수 있기 때문에 살수 및 집수조냉각수의 화학조건(pH) 조절능력이 요구된다. 현재 원전은 LOCA시 능동형 살수첨가제인 NaOH를 사용하여 격납건물 살수 및 집수조냉각수의 pH를 조절하도록 설계되어있다. 본 논문에서는 LOCA시 집수조냉각수의 pH를 분석하고, 살수화학조건 pH 관련 최신규제요건인 표준심사지침(SRP) 6.5.2에 따라 핵분열생성물제거상수 및 제염계수를 계산하였다. 분석결과, 격납건물집수조 pH는 8.09~9.67로서 설계기준을 만족한다. 그리고 격납건물살수계통에 의한 핵분열생성물 제거상수 및 제염계수는 원전 내환경기기검증을 위한 방사선환경 평가의 입력으로 제공된다. Nuclear Power Plants require the control ability of chemical condition (pH) because pH control during transient accident such as LOCA makes an able the fission product removal capability to be maintained, stress corrosion cracking of stainless steel equipment to be prevented and the production of hydrogen by aluminum and zinc to be minimized. An NPP is designed to control the pH of containment spray and sump coolant using the spray additives 30% NaOH in the event of loss of coolant accident. In this paper, the pH of sump coolant of an NPP during LOCA was analyzed and the fission products removal constant and decontamination factor were calculated according to Standard Review Plan 6.5.2 related to spray chemical conditions of pH. The calculated pH value of recirculation mode using the computer code corresponds to 8.09~9.67, which meets the chemical environment regulation requirements. The fission product removal capability caused by containment spray system is performed to provide input to radiation analysis.

The predicted effects of selected Phebus FPT1 sump constituents on iodine volatility

The analytical results of containment sump samples in the Phebus FP1Test show the presence of Cd, Sn, U, Ag, Re (destroyed thermocouples) as well as Ag at certain concentrations. An exploratory study was conducted at PSI to study the potential effect of these constituents by modelling different reactions and by conducting specific experimental programmes. Respective reactions are modelled in the PSI's iodine behaviour prediction code, PSIodine (Cripps et al., 2011a), by supplementing the aqueous phase iodine reactions of the code with reaction sets for Cd, Sn, U, Ag and Re. Their reaction rate constants with iodine species were either taken from the available literature or estimated using PSI's experimental data or made by ‘engineered guesses driven from similar reactions or diffusion controlled reactions.We show in four test cases, the effects of these measured FPT1 sump constituents on the fractional I2 yield released from the aqueous solution. They were studied in a comparative approach by including or excluding the reactions of iodine with these constituents. In Test Case 1 (without considering additional models but with the model for AgI dissolution and radiolytic decomposition), only 0.2% I2 is predicted to release. In Test Case 2 (which considers Case 1 reactions and silver species reactions). Excess and isolated Ag atoms have significantly increased I2 yields to 22% by shifting AgI and AgCl dissolution equilibria. In Test Case 3, considered are Case 2 reactions and reactions for Cd, Sn, UO2+ and ReO4− in estimated speciation have further increased AgI oxidation to ∼70% I2. Test Case 4 considered Case 3 reactions and reduction reactions of I2 and IO3− with metal species. The additional reduction reactions of metal species and ReO42− ions with I2 have suppressed the I2 release to only 0.2%. PSI experiments have confirmed the model prediction that irradiated and sparged CsI solutions at pH 4.6 produce lower %I2 yields with increasing perrhenate ion concentration.In conclusion, these results show that predicted iodine chemistry in the aqueous phase can be significantly affected by including additional models to simulate reactions between iodine and metallic species, which were reported in the Phebus FPT1 post-test sump. These models are first attempts and although most of the corresponding kinetic data were literature-sourced, some were “informed estimates”. Thus no claim is made for direct insertion of these models into established iodine prediction codes. Nevertheless, the authors hope that the results highlight the need for in-depth experimental and modelling studies of these and possibly other metallic sump constituents or suppression pools, that is, in research areas, which were hitherto relatively unexplored.’

Debris filtering efficiency and its effect on long term cooling capability

A cutting-edge debris-filter designs, Protective Grid (P-grid) and Guardian Grid (G-grid) attached to the upper part of bottom nozzle, have been employed for the PWRs in Korea since 2000s to protect the fuel from debris-induced fuel failures. The debris-filter efficiency of the P-grid and G-grid designs is improved by relatively smaller flow areas formed by the grid straps and dimples. The debris-filter efficiency of the P-grid design is further improved by the relatively smaller flow-hole bottom nozzle. The debris-filter flow tests employing eighteen debris types showed that the debris-filter efficiencies of the P-grid and G-grid designs are 91 and 96%, respectively, while that of the SYS80 fuel design having only the standard flow-hole bottom nozzle is 26%. The slightly better debris-filter efficiency of the G-grid design against the P-grid design may be explained by relatively smaller flow areas at the G-grid dimple region as well as by the relatively longer solid end plug and the higher G-grid strap. The P-grid design may capture circular shapes of debris larger than 3.44mm in diameter at the flow holes formed by the P-grid dimples, whereas the G-grid design may capture circular shapes of debris larger than 2.54mm in diameter at the flow holes formed by the G-grid dimples. The aforementioned difference in the debris-filter efficiency between the P-grid and G-grid designs may be predicted by the solid modeling technique generating three-dimensional flow paths. Using the minimum flow-hole areas generated by the P-grid and G-grid designs, on the other hand, the effect of debris injected from the containment sump into the reactor core on the long term cooling (LTC) capability after a loss of coolant accident (LOCA) was evaluated, which indicates that the debris-filter capability of the P-grid and G-grid designs may not have a detrimental effect on the LTC capability after a LOCA only if the sump mesh size is smaller than 2.54mm in diameter.

Applications of results from the Phébus-FP programme in the US regulatory process

Results of the Phébus-FP experiments are being used by the U.S. Nuclear Regulatory Commission to validate, update, and revise its mechanistic models of severe reactor accident phenomena. These updated models are being used to revise accident analysis source terms used in a variety of regulatory processes including plant siting, equipment qualification and control room habitability assessments. Important revisions include assumption of a less volatile chemical form for radioactive cesium, larger release of radioactive molybdenum and prolonged duration of the reactor core degradation process. The treatment of iodine release to the reactor containment remains problematical and requires additional research. This research should address the need to assure containment sump water pH be kept high throughout a reactor accident.

Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the insulation material during reverse flow. This will certainly further improve the coolability of the core.The spacer grids were modeled as a strainer, which completely retains all the insulation material reaching the uppermost spacer level. There, the accumulation of the insulation material gives rise to the formation of a compressible fibrous cake, the permeability of which to the coolant flow is calculated in terms of the local amount of deposited material and the local value of the superficial liquid velocity.Before the switch over of the ECC injection from the flooding mode to the sump mode, the coolant circulates in an inner convection loop in the core extending from the lower plenum to the upper plenum. The CFD simulations have shown that after starting the sump mode, the ECC water injected through the hot legs flows down into the core at the so-called breakthrough channels located at the outer core region where the downward leg of the convection roll had established. The hotter, lighter coolant rises in the center of the core. As a consequence, the insulation material is preferably deposited at the uppermost spacer grids positioned in the breakthrough zones. This means that the fibers are not uniformly deposited over the core cross section.When the inner recirculation stops later in the transient, insulation material can also be collected in other regions of the core. Nevertheless, with a total of 2.7kg fiber material deposited at the uppermost spacer level, the pressure drop over the fiber cake is not higher than 8kPa and all the ECC water could still enter the core.

Mechanical Analysis of Conflux Channel of Containment Sump Strainer

In this paper, the conflux channel of the containment sump strainer in nuclear power station was adopted as the research object. The modal analysis and the stress calculations were conducted in LOCA (Loss-of-Coolant Accident) condition to test the validity of the conflux channel. The research results could provide the basis for designing the double filter cylinder of the compact containment sump strainers.

Verification and Validation of Numerical Models of the Transport of Insulation Debris

A combination of the two-fluid and drift flux models have been used to model the transport of fibrous debris. This debris is generated during loss of coolant accidents in the primary circuit of pressurized or boiling water nuclear reactors, as high pressure steam or water jets can damage adjacent insulation materials including mineral wool blankets. Fibre agglomerates released from the mineral wools may reach the containment sump strainers, where they can accumulate and compromise the long-term operation of the emergency core cooling system. Single-effect experiments of sedimentation in a quiescent rectangular column and sedimentation in a horizontal flow are used to verify and validate this particular application of the multiphase numerical models. The utilization of both modeling approaches allows a number of pseudo-continuous dispersed phases of spherical wetted agglomerates to be modeled simultaneously. Key effects on the transport of the fibre agglomerates are particle size, density and turbulent dispersion, as well as the relative viscosity of the fluid-fibre mixture.

Mechanical Analysis of Filter Cylinder of Containment Sump Strainer

In this paper, the double filter cylinder of the compact containment sump strainer innuclear power station was adopted as the research object. The impact and transmission mechanicalexperiments were carried out to investigate the material and structural properties in the event ofearthquake and LOCA (Loss-of-Coolant Accident). Furthermore, the modal analysis and the stresscalculations were conducted in LOCA condition to test the validity of the filter cylinder. The researchresults could provide the basis for designing the double filter cylinder of the compact containmentsump strainers.

On the radiolysis of iodide, nitrate and nitrite ions in aqueous solution: An experimental and modelling study

High radiation fields are predicted in the atmospheres in the reactor containment under postulated severe accident conditions. In particular, an experimental and modelling project (PARIS) has provided results, which show that nitrogen dioxide was the dominant measured species at high doses (>1 kGy) in air/steam mixtures. In addition, nitrite and nitrate ions were measured in the post-test containment sumps in some Phebus FP in-reactor integral experiments; but their separate effects on iodine volatility from irradiated aqueous solutions of iodide using appear unreported in open literature. To address this issue, this paper presents an overview of the results of an experimental and modelling study carried out at PSI, Switzerland. The experimental study consisted of small-scale irradiations of CsI, boric acid and tracer containing aqueous solutions. Tests were first carried out without nitrate or nitrite ions to obtain results to confirm expected iodine behaviour by correlation with predicted results, which were generated by concurrently developed code (PSIodine). The solutions were sparged with N 2O, argon and air to provide different net oxidation systems and to remove volatile iodine for measurement. Since the radiation chemistry of N 2O-saturated iodide solutions is well established and the measured I 2 yields and final pH correlate very well with the PSIodine code predictions, the effect of added nitrite ions to these solutions has provided results, which support the correct choice of relevant reactions for inclusion in the nitrate-model. Both experimental and predicted results clearly show that nitrate or nitrite ions in argon-sparged and irradiated iodide solutions (conc., 10 −4 mol dm −3) containing boric acid lower % I 2 yields up to an initial NO 3 − concentration of ∼5.0 × 10 −3 mol dm −3. Using a low CsI concentration (4.0 × 10 −5 mol dm −3) estimated in containment sump during a postulated severe-accident, an initial nitrate concentration (10 −3 mol dm −3) and pH 7.1, ∼3.6% I 2 was produced by argon sparging at a dose of 20.5 kGy. In contrast, 80% I 2 was formed at a dose of 12 kGy in the absence of nitrate ions. Irradiated and air-sparged CsI solutions, due to the lower net oxidation, gave the expected lower % I 2 yields than their argon-sparged counterparts. However, the reduction of volatile iodine species by radiation products of nitrate ions was enhanced. For example, using CsI solutions (conc., 10 −4 mol dm −3), initial pH 4.6 and a range of nitrate ion concentrations (0, 10 −4, 10 −3 and 10 −2 mol dm −3), the % I 2 yields at a dose of 2 kGy have lowered from 26.9% to 25.6%, 3.2% and 2.8%, respectively. In the latter experiment, only ∼6% I 2 at a dose of ∼20 kGy was produced. In the absence of nitrate ions, a yield of ∼93% I 2 at only 5 kGy dose was obtained. By comparison, the % I 2 yields from their Ar-sparged counterpart experiments were lowered from ∼88% to ∼72%, 11% and 22% in the same nitrate concentration range. Therefore, apart from the lower net oxidation in O 2-containing solutions, the results indicate that the reduction of volatile iodine species by products of nitrate ion irradiation is enhanced in the presence of dissolved O 2.

Modeling the vapor source term associated with the spill of LNG into a sump or impoundment area

The recent publication of evaluation protocols for vapor source term models and vapor dispersion models have influenced the modeling approaches that can be used for approval of new and expansion projects at LNG receiving terminals. In the past few years the scientific basis of integral vapor source term models has been questioned with growing concerns regarding their validity. In this paper, the shallow water equations (SWEs) were solved to study the characteristics of the evaporating LNG pool associated with a constant flow rate spill of LNG into a concrete sump. In the early stages of pool spreading, the leading edge thickness profile of the SWE model scales with the square root of the distance from the leading edge as the pool spreads. After the edge of the pool reaches the wall, the reflected wave forms a hydraulic jump that travels back towards the center of the pool at a speed that is considerably slower than the initial spreading of the pool. Once the hydraulic jump reaches the center, the pool assumes a nearly flat free surface for the rest of the spill. The pool spreading and the rate of evaporation from the SWEs were then compared to the solution provided by the integral model, PHAST. The two approaches were found to agree well with one another. The SWE model was also used to demonstrate the influence of an elevated spill source. With an elevated source, the LNG pool spreads faster, significantly increasing the initial rate of vaporization and peak vaporization rate. This increase in the initial rate of vaporization could lead to an increase in the vapor cloud hazard distance. The SWE model was also used to demonstrate the influence of an inclined sump floor in the shape of an inverted cone where the spilling LNG accumulates in the low vertex of the cone. Inclined sump floors can be used to significantly reduce the cumulative evaporation, making them attractive as a possible mitigation approach in cases where a containment sump is located close to a property boundary.