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  • 1.
    Huang, Zheng
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Numerical Investigations on Debris Bed Coolability and Mitigation Measures in Nordic Boiling Water Reactors2019Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    This thesis is aiming at coolability assessment of particulate debris beds formed in hypothetical severe accidents of Nordic boiling water reactors (BWRs) which may employ either lower drywell flooding or control rod guide tubes (CRGT) cooling as severe accident management strategies. For this purpose, quench and cooling limit (dryout) of debris beds after their formation from fuel coolant interactions were investigated by numerical simulations using the MEWA code. The thesis works consist of four following tasks: (i) validations of the computational tool (the MEWA code) against latest experiments: POMECO-FL and Tutu tests for friction laws, POMECO-HT tests for dryout, and PEARL tests for quench; (ii) Assessment of the long-term coolability of prototypical debris beds with varied characteristics (e.g. height, shape); (iii) studies on quench of prototypical debris beds; (iv) effectiveness analysis of mitigation measures dedicated to Nordic BWRs in term of debris coolability enhancement.

    The comprehensive validations indicate that the MEWA code is a credible and computationally-efficient tool to simulate the two-phase thermal hydraulics of particulate debris beds under both thermal equilibrium and non-thermal equilibrium conditions. Comparisons of the predicted results with experimental data showed a satisfactory agreement, and key phenomena were reproduced.

    Simulation for the prototypical debris beds of the cylindrical, conical and truncated conical shapes showed that the beds’ heights were significantly affecting their coolability, and the values of their dryout power density were roughly inversely proportional to their heights regardless of shapes. Such a relationship was correlated based on the simulation results, which can be employed to guide design and operation of relevant experiments. The impacts of bed’s shape on coolability can be characterized by three factors: multidimensionality and contour surface area of debris bed, as well as uniformity of bed’s shape. An increase in uniformity can improve coolability, since it promotes uniform distributions of temperature and void fraction.

    Simulations for an ex-vessel heap-like debris bed for a reference Nordic BWR showed that the quench front propagated in a multi-dimensional manner. It was found that the upper region adjacent to the centerline of the bed was subject to a higher risk of remelting. It is also found that the oxidation of the residual Zr in the corium had a great impact on coolability of the debris bed due to the release of reaction heat and H2. Therefore, it is crucial to lower the temperature of the whole bed to avoid escalation of oxidation.

    Based on insights from previous studies, several coolability enhancement concepts were proposed as mitigation measures, including downcomer, bottom injection and CRGT cooling. Simulations demonstrated that all the three measures were effective to improve the debris bed coolability. An embedded downcomer increases the cooling capacity by inducing an extra downward flow of water and providing a preferential exit path for steam. Compared with top-flooding, water injection from bottom was predicted to be more efficient to quench a debris bed, since the water inflow was not hindered by the upward flow of steam and therefore could infiltrate the whole bed quickly. The CRGT cooling strategy, applying coolant injection to a particulate debris bed in the lower head of PRV, was proved to be practically feasible to quench the in-vessel debris bed. However, a special attention should be paid to the side effect of Zr oxidation, since it may deteriorate the quenching process and lead to an uncoolable state as a result of release of considerable heat and H2. Therefore, it is essential that the in-vessel debris bed is sufficiently cooled to such an extent during its formation that substantial oxidation would not occur when the CRGT cooling is applied.

  • 2.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Numerical investigation on quench of an ex-vessel debris bed at prototypical scale2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, p. 47-61Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with the coolability of the heap-like debris beds formed in the cavity of a Nordic-type boiling water reactor (BWR) during a postulated severe accident. A numerical simulation using the MEWA code was performed to investigate the quenching process of the ex-vessel debris bed at post-dryout condition upon its formation. To qualify the simulation tool, the MEWA code was first employed to calculate the quenching tests recently conducted on the PEARL facility. Comparisons of the simulation results with the experimental measurements show a satisfactory agreement. The simulation for the debris bed of the reactor scale shows that the heap-like debris bed flooded from the top is quenched in a multi-dimensional manner. The upper region adjacent to the centerline of the bed is the most difficult for water to reach under the top-flooding condition, and thus is subject to a higher risk of remelting. The oxidation of the residual Zr in the corium has a great impact on the coolability of the debris bed due to (i) large amount of reaction heat and the subsequent positive temperature feedback, (ii) the local accumulation of the produced H2 which may create a “steam starvation” condition and suppresses the oxidation. As possible mitigation measures of oxidation, the effects of bottom-flooding and bypass on quench were also investigated. It is predicted that the debris bed becomes more quenchable with water injected from the bottom, especially for the case with the floor partially flooded in the center. A bypass channel embedded in the center of the debris bed can also promote the quenching process by providing a preferential path for both steam escape and water inflow.

  • 3.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubesManuscript (preprint) (Other academic)
  • 4.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 351, p. 189-202Article in journal (Refereed)
    Abstract [en]

    Since the reactor pressure vessel (RPV) of a typical BWR features a lower head that is penetrated by a forest of control rod guide tubes (CRGTs), coolability of the debris bed formed in the lower head during a severe accident can be realized by coolant injection through the CRGTs (so-called "CRGT cooling"). This paper is concerned with performance assessment of such CRGT cooling system, whose heat removal capacity is determined by two mechanisms: (i) heat-up and boiling of coolant inside the CRGTs; and (ii) evaporation of coolant which reached the top of the debris bed from CRGTs (top flooding). For this purpose, analyses were accomplished by coupling the COCOMO and RELAP5 codes, which simulate the quenching process of the debris bed and the coolant flow inside the CRGTs, respectively. An analysis was first carried out for a unit cell with a single CRGT, whose decay heat removal was limited by heat conduction from debris to the CRGT wall. The simulation indicated that without top flooding, though the temperature of the unit cell was eventually stabilized by the cooling of the CRGT wall, remelting of metallic debris (Zr) in the peripheral region was unavoidable due to low conductivity of corium. Boiling in the CRGT was not only beneficial to heat transfer, but also contributing to a flat axial temperature profile. Given the nominal flowrate of the CRGT cooling, the coolant was not completely boiled off in the CRGT, and therefore the remaining liquid water at the outlet of the CRGT was available for top flooding of the debris bed. The subsequent simulation including the top flooding showed that the debris bed was rapidly quenched without any remelting. However, the top flooding may have a side effect which was Zr oxidation risk at high temperature, leading to production of reaction heat and H-2. Finally analyses were performed for prototypical cases for a reference Nordic BWR, and the results implied that the CRGT cooling could be used as a promising strategy for severe accident mitigation. It is critical that the debris bed is sufficiently cooled down during its formation so that the oxidation risk is eliminated when the CRGT cooling is applied.

  • 5.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Performance evaluation of passive containment cooling system of an advanced PWR using coupled RELAP5/GOTHIC simulation2016In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 83-92Article in journal (Refereed)
    Abstract [en]

    Motivated to investigate the thermal hydraulic characteristics and performance of a passive containment cooling system (PCS) for a Generation III pressurized water reactor (PWR), a coupled RELAP5/GOTHIC model was developed, which was then employed to simultaneously simulate the transient responses of the PCS and the containment during a large break loss of coolant accident of the reactor. The results show that the PCS is capable of lowering the containment pressure to an acceptable level for a long period (up to 3 days). In a separate-effect study, it was found that the height of the PCS loop plays an important role in determining the flow characteristics and heat removal performance of the PCS. Within the range of the considered loop heights, phase change occurs in the riser of the loop after the height exceeds a specific value (between 13 m and 15 m), below which only single-phase flow takes place. With increasing height of the loop, the heat removal capability increases monotonically at first; however, it is no longer sensitive to the height after two-phase flow appears. Finally, a feed-and-bleed operation for the cooling tank of the PCS was proposed as an enhancement measure of the heat removal capacity, and the simulation results show it further mitigates the accident. Moreover, a simplified analytical model is developed to predict the impact of the feed-and-bleed flowrate on the PCS performance, which can be used in engineering design.

  • 6.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 128, p. 330-340Article in journal (Refereed)
    Abstract [en]

    In the aftermath of the Fukushima Daiichi nuclear accident, a great concern has been raised about enhancing the inherent safety of a spent fuel pool (SFP). A passive cooling system using two-phase thermosiphon loops was concerned in this paper. A RELAP5/MELCOR coupling interface was developed, aiming at simultaneously simulating the transient behaviors of the SFP (by MELCOR) and the passive cooling system (by RELAP5). First the RELAP5 model of the thermosiphon loop was qualified against an experiment of a prototypical scale. Comparisons between the experiment and predictions show a good agreement. MELCOR standalone calculations for both station blackout (SBO) and loss of coolant accident (LOCA) without the passive cooling system demonstrate severe degradation of fuel rods. In contrast, for the SBO accident, the coupling simulation shows that the passive cooling system can effectively remove the decay heat, thus keeping fuel rods intact. As for the LOCA scenario, it is more challenging for the passive cooling system due to: (i) the heat transfer power is low during the drainage of water since the natural circulation of steam is blocked by the residual water at the bottom, leading to unavoidable heat-up and oxidation of fuel cladding; (ii) the heat transfer coefficient between steam and the evaporator is very small, which consequently may require a larger heat transfer surface area. Nevertheless, the heat transfer power substantially increases after the pool is emptied and natural circulation is established. The decay heat can be removed by steam convection, thus maintaining the mechanical integrity of fuel rods and stabilizing the fuel temperature eventually. It is also observed that H 2 production is undesirably promoted because the steam supply is enhanced. However such adverse effect can be diminished by increasing the thermosiphon loops number.

  • 7.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation and application of the MEWA code to analysis of debris bed coolability2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 327, p. 22-37Article in journal (Refereed)
    Abstract [en]

    This paper was first aimed at validating the MEWA code against experiments for two-phase flow and dryout in particulate beds, and then investigating the coolability of ex-vessel debris beds with cylindrical, conical and truncated conical shapes assumed to form under severe accident scenarios of a boiling water reactor. The validation was mainly performed against the POMECO-FL and POMECO-HT experiments carried out at KTH for investigating frictional laws and coolability limit (dryout) of particulate beds, respectively. The comparison of the experimental and numerical results shows that the MEWA code is capable of predicting both the pressure drop of two-phase flow through porous media and the dryout condition of various stratified beds. While the coolability of a one-dimensional homogeneous debris bed is bounded by counter-current flow limit (CCFL), the coolability of a heap-like debris bed can be improved due to lateral ingression of coolant in a multi-dimensional geometry. The simulations showed that the dryout power density of a prototypical debris bed was roughly inversely proportional to the bed's height regardless of the bed's shape. The impacts of a debris bed's features on coolability are manifested in three aspects: multidimensionality and contour surface area of the bed, as well as the uniformity of its shape. The contour surface area is defined as the interface between debris bed and water pool, and its effective value depends on the surface orientation that determines the amount of water ingress and vapor escape. The perfect uniformity in bed's shape as cylindrical bed results in even distributions of temperature and void fraction. The dryout power density was also predicted to be strongly correlated to the uniformity of bed's shape. The MEWA simulation also predicted that coolability was improved by an downcomer embedded in the center of debris bed. The efficiency of such enhancement was largely determined by the downcomer's length, whose optimal value was obtained in simulation.

  • 8.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thakre, Sachin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation of the MEWA code agsinst POMECO-HT experiments and cool ability analysis of stratified debris BEDS2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, 2015, Vol. 4, p. 3279-3291Conference paper (Refereed)
    Abstract [en]

    Motivated by qualification of the MEWA code for coolability analysis of debris beds formed during severe accidents of light water reactors, the present work presents a validation of the code against the experimental data obtained on the POMECO-HT facility for investigation of two-phase flow and heat transfer limits in particulate beds with various characteristics. The volumetrically heated particulate beds used in the POMECO-HT experiment are packed in various configurations, including homogeneous bed, radially stratification, triangular stratification, axial stratification, and multi-stratification. To investigate coolability enhancement by bottom-fed induced natural circulation, a downcomer is employed. Besides, the influence of the interfacial drag is also studied. The results show that simulation results of the MEWA code is overall comparable with the experimental data in term of dryout conditions of the particulate beds. For the 1-D top-flood case, the dryout heat flux is mainly determined by counter-current flow limit. While for certain cases the multidimensionality may help to break CCFL. Besides, the debris bed’s coolabiltiy can be significantly improved due to the natural circulation flow from the bottom induced by using downcomer. The interfacial drag affects the coolability by means of varying the pressure field inside the bed. For the top-flood case, the dryout condition deteriorates since the vapor and coolant flow reversely and thus the interfacial drag increases the flow resistance. Whereas for the bottom-fed case, the dryout heat flux rises remarkably when considering the interfacial drag, because the vapor and coolant flow in the same direction and the interfacial drag helps to pull coolant upward from the bottom.

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