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Huang, Z. & Ma, W. (2019). On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes. Nuclear Engineering and Design, 351, 189-202
Open this publication in new window or tab >>On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes
2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 351, p. 189-202Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
Severe accident, Debris bed coolability, CRGT cooling, Quench, Coupled simulation
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-255547 (URN)10.1016/j.nucengdes.2019.06.001 (DOI)000475396200017 ()2-s2.0-85067063084 (Scopus ID)
Note

QC 20190806

Available from: 2019-08-06 Created: 2019-08-06 Last updated: 2019-08-06Bibliographically approved
Huang, Z. & Ma, W. (2019). Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis. Annals of Nuclear Energy, 128, 330-340
Open this publication in new window or tab >>Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 128, p. 330-340Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
Coupling analysis, Severe accident, Spent fuel pool, Thermosiphon
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-246416 (URN)10.1016/j.anucene.2019.01.024 (DOI)000465054700035 ()2-s2.0-85060329248 (Scopus ID)
Note

QC 20190329

Available from: 2019-03-29 Created: 2019-03-29 Last updated: 2019-05-14Bibliographically approved
Huang, Z. & Ma, W. (2018). Numerical investigation on quench of an ex-vessel debris bed at prototypical scale. Annals of Nuclear Energy, 122, 47-61
Open this publication in new window or tab >>Numerical investigation on quench of an ex-vessel debris bed at prototypical scale
2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, p. 47-61Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Elsevier Ltd, 2018
Keywords
Coolability, Debris bed, Oxidation, Quench, Steam starvation, Boiling water reactors, Floods, Quenching, Two phase flow, Boiling water reactor (BWR), Local accumulations, Mitigation measures, Numerical investigations, Temperature feedback, Debris
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-236557 (URN)10.1016/j.anucene.2018.08.018 (DOI)000447107700005 ()2-s2.0-85051762018 (Scopus ID)
Note

Funding details: ENSI, Eidgenössisches Nuklearsicherheitsinspektorat; Funding details: 201409110105, CSC, China Scholarship Council; Funding details: CSC, China Scholarship Council; Funding text: This study was made possible by the support from the research programs of APRI, ENSI and NKS. The authors are grateful to the support of the China Scholarship Council (CSC, scholarship No. 201409110105). The authors acknowledge IKE of Stuttgart University for making the MEWA code available to us and providing technical supports in utilization of the codes. Special gratitude goes to Dr. Michael Buck for patient and knowledgeable guidance. QC 20181127

Available from: 2018-11-27 Created: 2018-11-27 Last updated: 2019-05-03Bibliographically approved
Huang, Z. & Ma, W. (2018). Validation and application of the MEWA code to analysis of debris bed coolability. Nuclear Engineering and Design, 327, 22-37
Open this publication in new window or tab >>Validation and application of the MEWA code to analysis of debris bed coolability
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 327, p. 22-37Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Elsevier, 2018
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-220367 (URN)10.1016/j.nucengdes.2017.11.038 (DOI)000425302400003 ()2-s2.0-85037543831 (Scopus ID)
Note

QC 20171219

Available from: 2017-12-19 Created: 2017-12-19 Last updated: 2019-05-03Bibliographically approved
Huang, Z. & Ma, W. (2016). Performance evaluation of passive containment cooling system of an advanced PWR using coupled RELAP5/GOTHIC simulation. NUCLEAR ENGINEERING AND DESIGN, 310, 83-92
Open this publication in new window or tab >>Performance evaluation of passive containment cooling system of an advanced PWR using coupled RELAP5/GOTHIC simulation
2016 (English)In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 83-92Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Elsevier, 2016
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-200214 (URN)10.1016/j.nucengdes.2016.10.004 (DOI)000390736400008 ()2-s2.0-84993967049 (Scopus ID)
Note

QC 20170202

Available from: 2017-02-02 Created: 2017-01-23 Last updated: 2017-02-02Bibliographically approved
Huang, Z. & Ma, W.On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes.
Open this publication in new window or tab >>On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes
(English)Manuscript (preprint) (Other academic)
Keywords
severe accident; debris bed coolability; CRGT cooling; quench; coupled simulation
National Category
Engineering and Technology
Research subject
Physics
Identifiers
urn:nbn:se:kth:diva-250708 (URN)
Note

QC 20190503

Available from: 2019-05-03 Created: 2019-05-03 Last updated: 2019-05-03Bibliographically approved
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Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0002-1179-2256

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