Evaluation of an accident management strategy using an emergency water injection in a reference PWR SFP

Annals of Nuclear Energy 113 (2018) 353–379

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Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene

Evaluation of an accident management strategy using an emergency water injection in a reference PWR SFP Kwang-Il Ahn a,⇑, YongHun Jung a, Jae-Uk Shin b, Won-Tae Kim b a b

Korea Atomic Energy Research Institute, Daedeok-daero 989 Beon-gil, Yuseong-gu, Daejeon 34057, Republic of Korea RETech Co., Ltd., 40, Samsung 1-ro 1-gil, Hwaseong-si, Gyeonggi-do 18449, Republic of Korea

a r t i c l e

i n f o

Article history: Received 2 June 2017 Received in revised form 21 November 2017 Accepted 26 November 2017 Available online 1 December 2017 Keywords: SFP BDBA Station blackout (SBO) Severe accident management (SAM) Severe accident scenarios MELCOR simulation

a b s t r a c t The Fukushima accident on March 11, 2011 has shown the relevance of examinations of severe accident inside a spent fuel pool (SFP) during beyond-design-basis external events, and the necessity for provisions to cope effectively with such events through a relevant severe accident management (SAM) strategy. Although the low decay heat of fuel assemblies and the considerable water inventory in an SFP can slow the progress of an accident compared to an accident in the reactor core, the numerous number of fuel assemblies stored inside it and the fact that the SFP building is not leak-tight present the potential for the formation of a direct path for fission products to rise from the SFP into the environment (i.e., a much greater severe accident risk). The purpose of this paper is to assess the effectiveness and success conditions of an emergency makeup water injection strategy, which is being as a representative SFP SAM measure after the Fukushima accident. Two typical accident scenarios (loss-of-cooling and lossof-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) were considered in the analysis. For the foregoing SAM strategies, the analysis results and relevant insights are summarized in relation to two major aspects: (a) the key events of the progression of an accident (such as the exposure, heat-up and degradation of the fuel assemblies; the generation of combustible gases such as Hydrogen; and the over-pressurization of the SFP building) and (b) the release of radiological fission products (such as Cesium and Iodine) into the environment. A simulation tool for severe accidents, MELCOR1.8.6, was used in the present analysis. Ó 2017 Elsevier Ltd. All rights reserved.

1. Introduction In terms of design, most spent fuel pools (SFPs) at nuclear power plants (NPPs) are large, robust monolithic structures, and the pool walls are generally made of steel-lined concrete which is more than 1-m thick (OECD, 2015). Fuel assemblies (FAs) inside the SFP are stored in racks that provide spacing for the coolant flow, and in some cases for criticality control. The pools are filled with several additional meters of water above the spent fuel (SF) to provide biological shielding. An active cooling and purification system maintains optimal conditions for the stored fuel. The cooling systems have built-in redundancies and are designed such that they are connected to emergency backup power to maintain their function. Furthermore, the inlet and outlet pipes of the cooling systems are positioned to prevent draining of the pool. The large volume of water provides significant thermal inertia, generally slowing transients and thus allowing sufficient time for operator ⇑ Corresponding author. E-mail address: [email protected] (K.-I. Ahn). https://doi.org/10.1016/j.anucene.2017.11.051 0306-4549/Ó 2017 Elsevier Ltd. All rights reserved.

intervention during the progression of an accident. Therefore, in recent times, the main concerns related to SFPs have mainly focused on (a) controlling the configuration of the fuel assemblies in the pool with no loss of pool coolant, and (b) ensuring adequate pool storage space to prevent fuel criticality events and to keep the fuel cool to reduce the potential for a design basis accident (DBA). Several pre-Fukushima studies along these lines are listed below:
Impact of high-density storage racks, resulting in a larger inventory of fission products (e.g., Cs-137) in the pool and a greater load on the pool cooling system (USNRC, 1975, 2001)
The potential for fire propagation between the assemblies in an air-cooled environment (Benjamin et al., 1979; Sailer et al., 1987)
Safety analysis of fuel assemblies in a SFP, which may be severely damaged during a fuel crud removal operation (OECD, 2008) However, the Fukushima accident on March 11, 2011, highlighted the possibility of severe accidents affecting an SFP as a type

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of beyond-design-basis accident (BDBA) and has stressed the necessity of provisions to cope effectively with such events through relevant severe accident management (SAM) strategies (USNRC, 2014; Song and Kim, 2014). Related to this, the low decay heat of fuel assemblies and the large water inventory in a SFP can slow the progress of an accident compared to an accident which occurs in the reactor core. However, numerous fuel assemblies are stored inside the SFP, which is typically not located in a leaktight building. This enables a direct path for radiological fission products such as Cesium and Iodine into the environment and thus a much greater risk of a severe accident. While in-reactor fueldegradation and the fission-product-release phenomenology are well known from previous severe accident studies, SFP-accident conditions may differ, and estimating SFP source terms may also be an additional challenge. Under such circumstances, implementing and strengthening provisions and countermeasures to prevent and mitigate BDBAs, including those causing severe damage to the SFP, and assessing their success and effectiveness can be considered as among the most crucial activities for the SAM of NPPs. Reflecting the foregoing circumstance, additional studies have been carried out after the Fukushima accident as follows:
Full-scale experiment and analytical studies of the possibility of fire propagation between the assemblies in an air-cooled environment (Durbin and Lindgren, 2012; Durbin et al., 2013; Adorni et al., 2016)
Identification of areas that need additional knowledge and identification of potential improvements in relevant computational models and tools, as well as on the status of SFP accidents and their associated mitigation strategies, as carried out for the purpose for better predictions of SFP accident progression and margin times related to significant releases of radiological source terms (OECD, 2015)
Plant-level analyses of SFP severe accident progressions to provide more practical insight to support SFP-related safety and accident management efforts (Ahn et al., 2016) The purpose of this paper is to assess the effectiveness and success conditions of an emergency water injection, a strategy under consideration as a representative SFP SAM strategy. Two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) were considered in the analysis. For the foregoing SAM strategies, the analysis results and relevant insights are summarized for (a) key chronological events (such as uncovery, the heat-up and degradation of fuel assemblies, and the generation of combustible gases including hydrogen, and over-pressurization of the SFP building), and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. A simulation tool for severe accidents, MELCOR1.8.6, was used in the present analysis. As a reference, the aforementioned injection flow paths to provide emergency cooling water from external sources to the steam generators (SGs) or reactor cooling systems (RCSs) and also the SFP are under consideration as typical measures to increase the mitigation capability during a prolonged station blackout (SBO) accident (Park and Ahn, 2015). The emergency cooling water system consists of a fixed pipe linking the SFP to the outside of the containment with a standby valve installed on the pipe. Following the occurrence of an accident, movable equipment (e.g., a fire truck hose) can be connected to the pipe outlet upon the opening of an isolation valve. While the inside of the containment cannot be made accessible or manageable in cases of accidents leading to very hazardous work conditions, an emergency cooling water system has the advantages of easy accessibility and maintenance during an accident because it can be operated from outside of the containment. In such cases, on-site intervention is not limited even

when fission products are released and ambient dose rates are already high before the fuel assembly starts to become uncovered. 2. Analysis methodology 2.1. Reference plant The reference plant selected for the present study was, the OPR1000 (Optimized Power Reactor, Korean Standardized NPP) (KEPCO, 1996), which is a two-loop pressurized water reactor (PWR) with a capacity of 1000 MWe (rated thermal power of 2815 MWth) and a large-dry containment. The reference SFP shares many common design features with typical PWR SFPs, including a large water-retaining structure monolithically consisting of thick reinforced concrete with an inner stainless steel lining to prevent leakage and maintain the quality of the water. The reference SFP is located adjacent to (but outside of) the primary containment, and its bottom is positioned above plant grade. Approximately 1,498 fuel assemblies can be stored inside the reference SFP. Spent fuel storage racks made of stainless steel with neutron absorbers are arranged on the floor of the SFP in square arrays. The racks are composed of spent fuel cell assemblies, base plates, and base support assemblies. The base plates provide a seating surface for the spent fuel assemblies. The racks have support assemblies that elevate them above the bottom pool. The lower gap region located below the base plate allows water to flow freely under the racks and circulate through the fuel assemblies. The spent fuel assemblies and racks are submerged in borated water, which is used as both a neutron absorber and coolant, and there is no cross flow between the rack cells. Fig. 1 shows the actual layouts of the reference SFP in two and three dimensions. The width and depth of the reference SFP are 10.7 m and 8.53 m (35 ft and 28 ft), respectively. The height of the reference SFP is 12.2 m (40 ft) and the minimum water level of the reference SFP allowed by technical specifications is 7 m above the top of the spent fuel assemblies and racks. On the other hand, Table 1 shows the maximum number of FAs that can be withdrawn from the reactor core to the SFP per fuel cycle (18 months for the reference plant) and the decay heat load estimated from those FAs (Ahn et al., 2016). In both normal and refueling modes, the last withdrawal of FAs from the reactor core is assumed to be 100 hours after a reactor trip, which is the maximum time allowed to move the spent fuels to the SFP in the case of the reference plant. Fig. 2 shows the corresponding progression of decay heat load with time after offload for both normal and refueling modes. 2.2. Cases analyzed In a previous study by the authors (Ahn et al., 2016), a plantspecific accident analysis served to examine six typical accident scenarios that could occur in the reference plant and each analysis was conducted using MELCOR1.8.6 SFP version (Gauntt et al., 2005). The six accident cases were six different combinations of three representative initiating events (i.e., LOCA, LOPI, and CLPI) and two representative reactor operation modes with their corresponding decay heat load values (i.e., normal and refueling). None of the six accident cases have any means of accident mitigation. The LOPI type of event, which leads to a loss of cooling water from the SFP, can have a range of causes, such as a (seismicinduced) break of the SFP wall, the failure of SFP penetration pipes, a loss of integrity of the SFP gate, heavy load drops (during dry cask movements), reactor-related phenomena causing structural failures, and a reactor pressure vessel (RPV) drain-down event (i.e., when an SFP is hydraulically connected to the RPV cavity). A LOPI

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Fig. 1. Two- and three-dimensional layouts of the reference SFP.

Table 1 Maximum heat loads according to the withdrawal history of the fuels. FA Withdrawal (offload) Histories

Plant Operation Mode

Number of FAs

Decay Heat Load (MWth)

100 hours after offload

Normal Refueling

68 177

4.1 9.8

Previously stored in SFP

–

860

1.1

event should be analyzed while taking into account the spectrum of the break sizes and the locations, for which different impacts are anticipated. As a reference case of this study, however, a flow path with a diameter of two inches (equivalent to the diameter of a nominally sized pipe) was modeled at the bottom center of the SFP. The foregoing break size induces a maximum leak rate of about 38 kg/sec immediately after it is triggered, and as the water level decreases, the leak rate decreases with small initial oscillations, but mostly drying out the SFP in about 11.3 hours for both normal and refueling operation modes. Table 2 lists the four base accident cases in this study, which were used to assess the effectiveness and success conditions of emergency makeup water injection, Here, two CLPI cases from

the original study are excluded. It is clear that the change in the SFP coolant level is closely related to the progression of the accident. In emergency water injection scenarios, the two key parameters, which govern how an accident progresses, in this case changes of the SFP coolant level and the temperatures of fuel assemblies, are therefore when and how much water to inject (i.e., the timing and rates of emergency water injections). The emergency water was injected at two different times: (1) when the spent fuel storage racks started to become uncovered (see MATRIX-A of Table 2) and (2) when half of the height of the spent fuel rods became uncovered (see MATRIX-B of Table 2). Starting with no injection, i.e., the base case, the injection was increased to 30%, 60%, or 100% of the maximum coolant evaporation rate in the LOCA scenario and the maximum leakage rate in the LOPI scenario. Consequently, a total of 28 simulation cases, consisting of four base cases and six injection cases for each base accident case (three for MATRIX-A and the other three for MATRIX-B), were analyzed as shown in Table 3. The maximum loss rates of the SFP coolant were derived from Eqs. (1) and (2) for the LOCA scenario.

_ v ap m

n

¼

Q_ n ¼ 2:5 kg=sec hfg  q

Fig. 2. Decay heat loads with time after offload from the reactor.

ð1Þ

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Table 2 Base accident cases for the present study. Base Cases

Initiating Event (I.E)

Operation Mode

Heat load [MWth]

Max. Coolant Loss Rate [kg/sec]

LOCA-N-00 LOCA-R-00

LOCA: Loss-of-Cooling Accident

Normal Refueling

5.2 10.9

2.5 (Evaporation) 5.2 (Evaporation)

LOPI-N-00 LOPI-R-00

LOPI: Loss of Pool Inventory

Normal Refueling

5.2 10.9

38 (Leakage) 38 (Leakage)

Table 3 Mitigation cases considering the different timings and rates of the emergency makeup water injection: MATRIX-A and MATRIX-B. Test Cases

(Max. Coolant Loss Rate [kg/sec])

Operating Mode (Heat Load [MWth])

LOCA-N-00 LOCA-N-03 LOCA-N-06 LOCA-N-10

LOCA (2.5)

Normal Mode (5.2)

LOCA-R-00 LOCA-R-03 LOCA-R-06 LOCA-R-10

LOCA (5.2)

Refueling Mode (10.9)

LOPI-N-00 LOPI-N-03 LOPI-N-06 LOPI-N-10

LOPI (38)

Normal Mode (5.2)

LOPI-R-00 LOPI-R-03 LOPI-R-06 LOPI-R-10

LOPI (38)

Refueling Mode (10.9)

_ v ap r ¼ m

Initiating Event (I.E)

Q_ r ¼ 5:2 kg=sec hfg  q

Water Injection Rate (kg/sec)

Water Injection Time (hrs)

(Percent. of Max. Coolant Loss Rate)

[MATRIX-A] Start of SF Racks uncovery (4.718 m)

[MATRIX-B] SF rods half uncovery (2.261 m)

0.75 (30%) 1.5 (60%) 2.5 (100%)

N/A 116.4 116.4 116.4

137.6 137.6 137.6

1.56 (30%) 3.12 (60%) 5.2 (100%)

N/A 52.8 52.8 52.8

63.2 63.2 63.2

11.4 (30%) 22.8 (60%) 38 (100%)

N/A 6.1 6.1 6.1

8.2 8.2 8.2

11.4 (30%) 22.8 (60%) 38 (100%)

N/A 6.4 6.4 6.4

9.1 9.1 9.1

ð2Þ

where _ v ap n = maximum evaporation rate of the coolant during norm mal operation _ v ap r = maximum evaporation rate of the coolant when m refueling Q_ n = heat load during normal operation = 5.20 MWth

Q_ r = heat load during refueling = 10.9 MWth hfg = latent heat of water vaporization at 0.1 MPa = 2117.06 kJ/ kg q = density of water at 333.15 K = 983.15 kg/m3 2.3. MELCOR modeling In this study, a computational tool for severe accidents, MELCOR1.8.6 (YV.3084.SFP version) (Gauntt et al., 2005), was used to assess the effectiveness and success conditions of an emergency makeup water injection for the reference SFP. The MELCOR inputs were developed in accordance with the plant-specific characteristics of the reference SFP and with various conditions associated with the accident progression and mitigation. The MELCOR code is based on a building block design (i.e., a user-specified arrangement of control volumes, flow paths, and heat conductors). The COR package of the code includes specialized models to represent the primary thermal radiation pathways for the SFP geometry, e.g., a rack component allowing the separate modeling of the SFP rack and the radiation heat transfer between the fuel and the rack. MELCOR1.8.6 also includes an empirical life-

time breakaway oxidation model that was developed by SNL (Sandia National Laboratories) to take into account the transition to the breakaway oxidation kinetics in air environments when the corresponding condition is met. The present analysis employs an exponential coefficient of 42.038 and a maximum lifetime parameter (LF) value of 1.2 as model parameters to trigger the breakaway oxidation, as applied in the OECD Sandia Fuel Project benchmark study (Durbin and Lindgren, 2012; Ahn, 2013). In contrast, the code currently does not include a thermal–mechanical model for rod ballooning but does have provisions for variable inertial and viscous resistance levels as a function of other calculated parameters. The rack module is composed of fuel cell assemblies, a base plate, and base support assemblies. The base plate has chamfered through holes centered at each storage location, providing a seating surface for the fuel assemblies. The racks have support assemblies that elevate them above the bottom pool. Below the rack base plate is the lower gap region which allows water to flow freely under the racks and circulate through the fuel assemblies. The SFP racks and the lower gap region below the SFP racks can be modeled using existing core components. The MELCOR1.8.6 COR (Core) package is designed with a two-dimensional cylindrical geometry and nodalization of the SFP must fit within this framework. 2.3.1. COR packages Fig. 3 shows the modeling of the fuel assemblies and racks through the MELCOR1.8.6 COR package, reflecting the plantspecific SFP structure and spatial distribution of the FAs and the racks. The spent fuel assemblies and racks including the lower gap region, were modeled as three rings in a two-dimensional cylindrical geometry using the COR package. This was done in a manner that reflected the difference between the two reactor

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Fig. 3. Modeling of fuel assemblies in the COR package.

operating modes, as well as the plant-specific characteristics, in this case the spatial distribution of the spent fuel assemblies and racks. In this three-ring model, racks with newly withdrawn spent fuel assemblies (68 assemblies with a heat load of 4.1 MWth for the normal operation mode or 177 assemblies with a heat load of 9.8 MWth for the refueling mode) and with previously stored spent fuel assemblies (860 assemblies with a heat load of 1.1 MWth for both reactor operating modes) were modeled as Ring 1 and Ring 2, respectively. Empty storage racks (with no spent fuel assembly or heat load) were modeled as Ring 3. Each ring was

composed of 17 nodes, 10 nodes of which (from the fifth to the fourteenth from the bottom up) represented the active fuel region. 2.3.2. CVH, FL, HS, and CAV packages Fig. 4 shows the nodalization scheme of the reference SFP, which is composed of 18 control volumes (CVs) and 21 flow paths (FLs). These control volumes and flow paths were modeled using the CVH (Control Volume Hydrodynamics) and FL (Flow) packages of the MELCOR code, respectively. As mentioned in the previous section, any cross flow between SF racks (or among COR rings)

Fig. 4. Nodalization scheme of the reference SFP.

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except for a flow path through the bottom of SFP racks is not allowed in the present SFP. The walls around the pool (its thickness is 2.36 m), the upper volume of the building, and the stair room were modeled using the HS (Heat Structure) package for an accurate simulation of the condensation and evaporation of the SFP coolant on the surface, the deposition and resuspension of radionuclides on the surface as well as the heat transfer through the surface itself. To simulate leakages of steam and of gases produced during an accident, both of which can carry radionuclides, a flow path to the environment was modeled along a manway door (2.32 m2) between the upper volume of the SFP building and the stair room. The temperature differences between the channels and inside of a channel, induced by the depletion of SFP water and the uncovery of SF channels (racks), may result in a large amount of driving force for thermal radiation among different channels and between spent fuel channels and the pool wall. To predict thermal behavior in the pool, fuel rods, and structures of the SFP, the present analysis considers (a) the natural circulation heat transfer in the pool and (b) the radiation heat transfer between the fuel channels, racks, and SFP wall (with an axial and radial view factor of 0.1 as a reference case). Here, it should be noted that while an increase in the radial view factor increases the corresponding heat loss to neighboring cells, there is no formal guidance to specify the proper values for a plant-level SFP. The prediction differences during recent benchmark studies in which integral code simulations were conducted (Durbin and Lindgren, 2012; Ahn, 2013; Coindreau, 2016) varied (especially for the progression of fuel damage) when different modeling options were employed. According to a recommendation in NUREG/CR-6119 (Gauntt et al., 2005), the CAV (Cavity) package and a hypothetical reactorvessel lower head (LH) model were used to simulate relocated spent fuel assemblies, damaged and/or melted fuel debris falling to the bottom of the SFP, and molten core (melted spent fuel assemblies in this case) concrete interaction (MCCI) phenomena, all of which could occur after the collapse of a spent fuel assembly onto the concrete floor of a SFP. When melted fuel debris is contacted with the SFP structure, the SFP steel liner is assumed to be failed at its melting temperature (set to be 1273.15 K in the present analysis). As shown in Fig. 5, the COR package was hardwired into the CVH and CAV packages to include the CVH volume below the lower head, representing the cavity volume below the reactor core vessel. 2.3.3. RN package Finally, the RN (RadioNuclide) package was used to evaluate the amount of radionuclides released into the environment. The initial inventory data as calculated using the isotope generation and depletion code developed by ORNL (Oak Ridge National Laboratories), ORIGEN2 (Croff, 1980), which takes into account both the storage periods of the spent fuel assemblies and their burnup rates was used instead of the direct use of the default data. The initial

Fig. 5. Modeling with hypothetical LH and Cavity.

masses of two typical radionuclides, i.e., Cesium and Iodine, were defined as 3358.5 kg and 379.1 kg in the normal operation mode and 3609.1 kg and 404.8 kg in the refueling mode, respectively.

3. Analysis results 3.1. Base case analysis Fig. 6 and Table 4 summarize the major MELCOR analysis results for the four base accident cases, which were simulated for up to 168 hours (7 days) after the accident. For the LOCA scenario in the normal operation mode (LOCA-N00), the level of the SFP coolant continued to fall but did so slower than that in any of the other scenarios, resulting in the full uncovery of the fuel rods in 162.4 hours. However, the SFP coolant did not completely dry out by the end of calculation. The SFP coolant level reached nearly 0.12 m (just above the bottom of the SFP) at the end of the calculation. During the calculation, the maximum fuel cladding temperature rose to 2365 K, and 1148.2 kg of hydrogen was generated by a cladding oxidation under steam. In this case, the fuel assemblies did not collapse by the end of the calculation. However, Fig. 6(c) also shows that if the analysis would be extended the spent fuel would be collapsed very soon. Actually, when the analysis was extended up to 200 hours the fuel assemblies in Ring-1 and Ring-2 collapsed in 170 hours and 180 hours, respectively. On the other hand, 628.1 kg of radioactive Cesium and 71.0 kg of Iodine (18.7% of their initial inventory levels) were released into the environment by the end of the calculation. Here, the released Iodine was calculated from its compound (CsI), while Cs was calculated from its compound (CsI and CsOH) as well as the Cs element itself. The foregoing result is understandable when considering that the reference SFP building is not constructed as a perfectly leak-tight structure; therefore, there always exist a certain rate of coolant leakage from the building, even in LOCA scenarios with no breaks. This case witnessed cladding oxidation beyond the capability of the SFP to retain the resulting radionuclides by deposition, consequently leading to a release of radionuclides into the environment. For the LOCA scenario in the refueling mode (LOCA-R-00), the level of the SFP coolant continued to drop much more rapidly than that in the LOCA-N-00 case, resulting in the full uncovery of the fuel rods in 73.0 hours and complete dryout of the SFP coolant in 79.4 hours. During the calculation, the maximum fuel cladding temperature rose to 2388 K and 1440.2 kg of hydrogen was generated by a cladding oxidation under steam. In this case, the integrity of the fuel assemblies in both the Ring 1 and 2 regions was lost, and the assemblies finally collapsed in 78.3 hours (Ring 1) and 96.1 hours (Ring 2). The base support assemblies in both regions also collapsed soon after. As a result, the melted fuel assemblies weighing 726,709 kg fell onto the bottom of the SFP, 4138.0 kg of hydrogen was generated from MCCI, and the concrete ablation depth reached 1.41 m by the end of the calculation. Eventually, 1421.4 kg of radioactive Cesium and 157.7 kg of Iodine (39.4% and 39.0% of the respective initial inventory levels) were released into the environment by the end of the calculation. For the LOPI scenario in the normal operation mode (LOPI-N00), the level of the SFP coolant continued to fall significantly more rapidly than that of either LOCA-N-00 or LOCA-R-00, resulting in the full uncovery of the fuel rods in 10.1 hours and complete dryout of the SFP coolant in 10.5 hours. During the calculation, the maximum fuel cladding temperature rose to 2066 K and 298.9 kg of hydrogen was generated by a cladding oxidation under steam. In this case, the integrity of the fuel assemblies of both Ring 1 and 2 regions was lost, and the assemblies finally collapsed in 11.5 hours (Ring 1) and 60.8 hours (Ring 2). The base support

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359

Fig. 6. MELCOR analysis results for the base accident cases.

assemblies in both regions also collapsed soon after. As a result, the melted fuel assemblies weighing 659,103 kg dropped onto the bottom of the SFP, 3940.9 kg of hydrogen was generated from MCCI,

and the concrete ablation depth reached 1.37 m by the end of the calculation. Eventually, 1533.1 kg of radioactive Cesium and 172.6 kg of Iodine (45.6% and 45.5% of the respective initial

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Fig. 6 (continued)

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K.-I. Ahn et al. / Annals of Nuclear Energy 113 (2018) 353–379 Table 4 Major MELCOR analysis results of the base accident cases. Base Cases

Time to a Change in the SFP Coolant Level (Inventory) [hrs] Start of SF Racks Uncovery (4.718 m)

Start of SF Rods Uncovery (4.166 m)

SF Rods Half Uncovery (2.261 m)

SF Rods Full Uncovery (0.356 m)

SFP Complete Dryout (<0.01 m)

Cladding Oxidation

MCCI (Concrete Ablation [m]) (168 hrs)

LOCA-N00 LOCA-R00 LOPI-N00 LOPI-R00

116.4

122.8

137.6

162.4

1148.2

–

52.8

55.9

63.2

73.0

N/A* (0.115 m) 79.4

1440.2

4138.0 (1.41)

6.1

6.4

8.2

10.1

10.5

298.9

3940.9 (1.37)

6.4

7.3

9.1

10.8

11.4

538.9

5271.7 (1.90)

Hydrogen Generation [kg]

Note: (*) The level of SFP coolant approached the corresponding level.

Fig. 7. MELCOR analysis results for different injection rates (LOCA-N, MATRIX-A).

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Fig. 7 (continued)

inventories) were released into the environment by the end of the calculation. For the LOPI scenario in the refueling mode (LOPI-R-00), the level of the SFP coolant continued to drop somewhat slower than in the LOPI-N-00 case, resulting in the full uncovery of the fuel rods in 10.8 hours and complete dryout of the SFP coolant in 11.4 hours. During the calculation, the maximum fuel cladding temperature rose to 2168 K and 538.8 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies in both the Ring 1 and 2 regions was lost and the assemblies finally collapsed in 13.2 hours (Ring 1) and 45.3 hours (Ring 2), respectively. The base support assemblies in both regions also collapsed soon after. Hence, melted fuel assemblies weighing 719,743 kg settled onto the bottom of the SFP, 5271.7 kg of hydrogen was generated from MCCI, and the concrete ablation depth reached 1.90 m by the end of the calculation. Eventually, 2136.8

kg of radioactive Cesium and 240.5 kg of Iodine (59.2% and 59.4% of the respective initial inventory levels) were released into the environment by the end of the calculation. In summary, the level of the SFP coolant continuously decreased, and the fuel rods were fully uncovered by the end of the calculation in all four cases. In all cases except for LOCA-N00, the SFP coolant completely dried out, leading to a collapse of the fuel assemblies, MCCI phenomena, and a release of radionuclides into the environment by the end of the calculation. As mentioned before, however, it is expected that LOCA-N-00 would give a similar trend with the other three cases if the analysis would be extended to a longer time than 168 hours, which can cause a challenge for the SFP. The foregoing results for the LOCA scenarios also show that the refueling mode (LOCA-R-00), due to its higher heat load, resulted in more rapid accident progression as well as more severe accident

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363

Fig. 8. MELCOR analysis results for different injection rates (LOCA-N, MATRIX-B).

consequences than in the corresponding normal operation mode (LOCA-N-00). Meanwhile, there was no significant difference in the time to SFP exposure and complete dryout between two LOPI scenarios (LOPI-N-00 and LOPI-R-00) compared to the corresponding LOCA scenarios, most likely due to the recirculating upward flows caused by natural convection in the lower gap region and the driving force of which is provided through heat transfer from the spent fuel rods. The higher decay heat load in the refueling mode caused a larger amount of coolant to recirculate through the spent fuel storage racks, reducing the mass of coolant leaking out through the break point at the bottom center position of the SFP. The differences in hydrogen generation caused by cladding oxidation between LOCA and LOPI are mainly due to the relatively rapid water drainage which allows the high temperature to be reached in the presence of steam in the LOPI case.

For all of base cases, a large amount of burnable gases (such as hydrogen and carbon monoxide) were generated by a cladding oxidation and MCCI, but a concentration of oxygen reached 3–4% at the most, but a steam (H2O) concentration exceeded at least 60%. This is mainly due to non leak-tight feature of the present SFP building, consequently making most of air (oxygen and nitrogen) inside the building be released to the environment through the stair wary. Under the foregoing situation, there may be a small local burn of hydrogen, but a hydrogen deflagration, which can endanger the SFP building, is not expected to take place. Regarding an additional impact of MCCI, there was one case (LOPI-R-00), in which the SFP wall is failed by the radial erosion at the end of calculation. Moreover, Fig. 6(i) shows that if the present analysis is extended longer the SFP wall would be failed even for both LOCA-R-00 and LOPI-N-00 cases.

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Fig. 8 (continued)

3.2. Mitigation case analysis For a comparison with the base cases, corresponding sensitivity analyses were conducted while taking into account different mitigation strategies with different injection times and emergency makeup water rates. 3.2.1. LOCA Scenarios in the Normal Operation Mode (LOCA-N) Fig. 7 shows the results of the LOCA-N scenario analysis of the three different injection rates at 116.4 hours (when the fuel storage racks started to become uncovered; see MATRIX-A in Table 3). When the emergency makeup water injection rate was 0.75 kg/ s (30% of the maximum coolant evaporation rate, LOCA-N-03-A),

the level of SFP coolant continued to decrease somewhat slower than that in the LOCA-N-00 case. However, by the end of the calculation, the fuel rods were not fully uncovered and the SFP coolant did not completely dry out. The SFP coolant level was nearly 0.93 m (only about 0.58 m above the bottom of the fuel rods) at the end of the calculation. During the calculation, the maximum fuel cladding temperature rose to 2214 K and 60.4 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies was maintained and they did not collapse by the end of the calculation. Nonetheless, due to the occurrence of a coolant leak and given that the extent of cladding oxidation extended beyond the radionuclide retainability of the SFP, 27.8 kg of radioactive Cesium and 3.1 kg of Iodine (0.827% of

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Fig. 9. MELCOR analysis results for different injection rates (LOCA-R, MATRIX-A).

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Fig. 9 (continued)

their initial inventory levels) were released into the environment by the end of the calculation. When emergency makeup water was injected at a rate of 1.5 kg/s (60% of the maximum coolant evaporation rate, LOCA-N-06-A), the level of the SFP coolant continued to decline much more slowly than that of the LOCA-N-00 case. However, by the end of the calculation, the fuel rods were not fully uncovered and the SFP coolant did not completely dry out. The SFP coolant level reached approximately 3.35 m (more than 1.08 m above half the height of the fuel rods) at the end of the calculation. Only when emergency makeup water was injected at a rate of 2.5 kg/s (100% of the maximum coolant evaporation rate, LOCA-N-10-A) the level of the SFP coolant start to recover immediately after the injection, and this continued until it reached nearly 6.06 m (approximately 1.34 m above the top of the fuel rods) at the end of the calculation. In the two cases, the maximum fuel cladding temperatures rose to only 506 K (LOCA-N06-A) and 383 K (LOCA-N-10-A), respectively, leading to neither cladding oxidation nor a collapse of the fuel assembly and therefore no release of radionuclides into the environment by the end of the calculation. On the other hand, Fig. 8 shows the results of the LOCA-N scenario analysis for three different injection rates at a time of 137.6 hours (when half of the height of the fuel rods starts to become uncovered; see MATRIX-B in Table 3). When the emergency makeup water was injected at a rate of 0.75 kg/s (30% of the maximum coolant evaporation rate, LOCAN-03-B), the level of the SFP coolant continued to decrease at a rate slightly slower than that of LOCA-N-00 but slightly more rapidly than in the LOCA-N-03-A case. However, by the end of the calculation, the fuel rods were not fully uncovered and the SFP coolant was

not completely dried out. The SFP coolant level reached about 0.45 m (just above the bottom of the fuel rods) at the end of the calculation. During the calculation, the maximum cladding temperature rose to 2362 K and 397.8 kg of hydrogen was generated by the oxidation of the cladding under steam. In this case, the integrity of the fuel assemblies was maintained and they did not collapse by the end of the calculation. Nonetheless, due to the existence of a coolant leak and the extent of the cladding oxidation, which exceeded the radionuclide retainability of the SFP, 724.8 kg of radioactive Cesium and 81.9 kg of Iodine (21.6% of their initial inventory levels) were released into the environment by the end of the calculation. When the emergency makeup water was injected at a rate of 1.5 kg/s (60% of the maximum coolant evaporation rate, LOCA-N06-B), the level of the SFP coolant continued to decline at a rate slightly slower than that of LOCA-N-00 but significantly more rapidly than that of LOCA-N-06-A. However, by the end of the calculation, the fuel rods were not fully uncovered and the SFP coolant did not completely dry out. The SFP coolant level was close to 1.27 m (about 0.92 m above the bottom of the fuel rods) at the end of the calculation. Only when the emergency makeup water was injected at a rate of 2.5 kg/s (100% of the maximum coolant evaporation rate, LOCA-N-10-B) the level of the SFP coolant starts to recover immediately after the injection, and this continued until it reached approximately 3.27 m (about 1.01 m above half of the height of the fuel rods) at the end of calculation. In both cases, the maximum fuel cladding temperatures rose to only 605 K (LOCA-N-06-B) and 573 K (LOCA-N-10-B), respectively, leading to neither cladding oxidation nor a collapse of the fuel assemblies and therefore no release of radionuclides into the environment by the end of the calculation.

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Fig. 10. MELCOR analysis results for different injection rates (LOCA-R, MATRIX-B).

3.2.2. LOCA scenarios in the refueling operation mode (LOCA-R) LOCA scenarios in the refueling mode (LOCA-R) were analyzed for a direct comparison with the corresponding LOCA-N cases. Fig. 9 shows the results of the LOCA-R scenario analysis for three different injection rates at a time of 52.8 hours (when the fuel storage racks start to become uncovered; see MATRIX-A in Table 3). When the emergency makeup water was injected at a rate of 1.56 kg/s (30% of the maximum coolant evaporation rate, LOCAR-03-A), the level of the SFP coolant continued to decrease somewhat slower than that of LOCA-R-00, resulting in the full uncovery of the fuel rods in 87.0 hours and a complete dryout of the SFP coolant in 102.4 hours. During the calculation, the maximum fuel cladding temperature rose to 2410 K and 1133.1 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies in both the Ring 1 and 2 regions was lost and these regions finally collapsed in 86.7 hours (Ring 1) and 113.1 hours (Ring 2). The base support assemblies in the Ring 1

region collapsed in 102.5 hours, whereas those in the Ring 2 region did not collapse by the end of the calculation. As a result, 477,920 kg of melted fuel assemblies sank to the bottom of the SFP, 1240.5 kg of hydrogen was generated from MCCI, and the concrete ablation depth reached 0.86 m by the end of the calculation. Eventually, 1715.6 kg of radioactive Cesium and 196.4 kg of Iodine (47.5% and 48.5% of the respective initial inventory levels) were released into the environment by the end of the calculation. When emergency makeup water was injected at a rate of 3.12 kg/s (60% of the maximum coolant evaporation rate, LOCA-R-06-A), the level of the SFP coolant continued to decrease far more slowly than that of LOCA-R-00, resulting in the full uncovery of the fuel rods in 156.0 hours. However, the SFP coolant did not completely dry out by the end of calculation. The SFP coolant level reached nearly 0.18 m (just above the bottom of the SFP) by the end of the calculation. During the calculation, the maximum fuel cladding temperature rose to 2388 K and 548.5 kg of hydrogen was gener-

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Fig. 10 (continued)

ated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies in the Ring 1 region was lost, and this component finally collapsed in 155.8 hours. However, the fuel assemblies in the Ring 2 region did not collapse by the end of the calculation. As a result, the base support assemblies in both regions did not collapse by the end of the calculation. Despite the absence of MCCI and an ejection of melted fuel assemblies, due to the existence of a coolant leak and the extent of cladding oxidation, which exceeded the radionuclide retainability of the SFP, 1421.5 kg of radioactive Cesium and 159.5 kg of Iodine (39.4% of the corresponding initial inventory levels) were released into the environment by the end of the calculation. Only when emergency makeup water was injected at a rate of 5.2 kg/s (100% of the maximum coolant evaporation rate, LOCA-R-10-A), the level of the SFP coolant start to recover immediately after the injection, and this continued until the level reached approximately 9.38 m (about

4.66 m above the top of the fuel storage racks) by the end of the calculation. In this case, the maximum fuel cladding temperature remained below 400 K, leading to neither cladding oxidation nor a collapse of the fuel assemblies and therefore no release of radionuclides into the environment by the end of the calculation. On the other hand, Fig. 10 shows the results of the LOCA-R scenario analysis for the three different injection rates made at a time of 63.2 hours (when half of the height of the fuel rods starts to become uncovered; see MATRIX-B in Table 3). When the emergency makeup water was injected at 1.56 kg/s (30% of the maximum coolant evaporation rate, LOCA-R-03-B), the level of the SFP coolant continued to decrease significantly more slowly than that of LOCA-R-00, and even slightly more slowly than that of LOCA-R-03-A, resulting in the full uncovery of the fuel rods in 93.8 hours and a complete dryout of the SFP coolant in 152.1 hours. During the calculation, the maximum fuel cladding

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Fig. 11. MELCOR analysis results for different injection rates (LOPI-N, MATRIX-A).

temperature rose to 2564 K and 3509.9 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies in the Ring 1 region was lost and they finally collapsed in 120.5 hours. However, in contrast to LOCA-R-03-A, the fuel assemblies in the Ring 2 region did not collapse by the end of the calculation. As a result, the base support assemblies in both regions did not collapse by the end of the calculation. Although there were no MCCI effects and no ejection of melted fuel assemblies, due to the existence of a coolant leak and the extent of cladding oxidation beyond the radionuclide retainability of the SFP, 1644.0 kg of radioactive Cesium and 185.9 kg of Iodine (45.6% and 45.9% of the respective initial inventory levels) were released to the environment by the end of the calculation. When emergency makeup water was injected at a rate of 3.12 kg/s (60% of the maximum coolant evaporation rate, LOCA-R-06-B),

the level of the SFP coolant was initially decreased and the fuel rods became fully uncovered for some time, reaching, at a minimum, 0.26 m (just below the bottom of the fuel rods) in 134.7 hours. Subsequently, however, the level of the SFP coolant started to recover, and this continued until it reached approximately 1.49 m (only about 0.77 m below half of the height of the fuel rods) at the end of the calculation, unlike LOCA-R-06-A, which showed no recovery of the SFP coolant. During the calculation, the maximum fuel cladding temperature rose to 2561 K and 3745.2 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies in the Ring 1 region was lost and they eventually collapsed in 155.0 hours. However, the fuel assemblies in the Ring 2 region did not collapse by the end of the calculation. As a result, the base support assemblies in both regions did not collapse by the end of the calculation. Although

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Fig. 11 (continued)

the effects of MCCI and ejection of melted fuel assemblies were not noted, due to the existence of a coolant leak and the extent of cladding oxidation, which exceeded the radionuclide retainability of the SFP, 1826.3 kg of radioactive Cesium and 205.0 kg of Iodine (50.6% of their initial inventory levels) were released into the environment by the end of the calculation. Only when emergency makeup water was injected at 5.2 kg/s (100% of the maximum coolant evaporation rate, LOCA-R-10-B), the level of the SFP coolant start to recover immediately after the injection, and this continued until it reached nearly 7.14 m (about 2.43 m above the top of the fuel storage racks) at the end of the calculation. In this case, the integrity of the fuel assemblies was maintained and they did not collapse by the end of the calculation. Nevertheless, during the calculation, the maximum fuel cladding temperature rose to 788 K and 1.58 kg of hydrogen was generated by cladding oxidation under steam in contrast to the MATRIX-A

case (i.e., LOCA-R-10-A). This occurred due to the later injection of water compared to that in the MATRIX-A case, which allowed the upper parts of the fuel rods to become uncovered for a while. In this case, because the integrity of the fuel assemblies was maintained and cladding oxidation was limited to a negligible amount such that the SFP could retain all of the resulting radionuclides by deposition, there were no radionuclides released into the environment by the end of the calculation. 3.2.3. LOPI scenarios in the normal operation mode (LOPI-N) Fig. 11 shows the results of the LOPI-N scenario analyses of the three different injection rates made at a time of 6.1 hours (when the fuel storage racks start to become uncovered; see MATRIX-A in Table 3). When emergency makeup water was injected at a rate of 11.4 kg/s (30% of the maximum coolant leakage rate, LOPI-N-03-A), the

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Fig. 12. MELCOR analysis results for different injection rates (LOPI-N, MATRIX-B).

level of the SFP coolant continued to decrease somewhat more slowly than that of LOPI-N-00, resulting in the full uncovery of the fuel rods in 17.5 hours. However, the SFP coolant did not completely dry out by the end of the calculation. The SFP coolant level reached nearly 0.02 m (just above the bottom of the SFP) by the end of the calculation. During the calculation, the maximum fuel cladding temperature rose to 2368 K and 156.5 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies was maintained and they did not collapse by the end of the calculation. Nonetheless, due to the existence of a coolant leak and the extent of cladding oxidation, which exceeded the radionuclide retainability of the SFP, 2602.4 kg of radioactive Cesium and 293.7 kg of Iodine (77.5% of their initial inventory levels) were released into the environment by the end of the calculation. When emergency makeup water was injected at a rate of 22.8 kg/s (60% of the maximum coolant leakage rate, LOPI-N-06-A), con-

trary to the LOPI-N-03-A case, the level of the SFP coolant started to recover immediately after the injection. After a partial recovery to approximately 5.26 m (only about 0.54 m above the top of the fuel storage racks), it remained nearly constant until the end of the calculation. When emergency makeup water was injected at a rate of 38 kg/s (100% of the maximum coolant leakage rate, LOPI-N-10-A), the level of the SFP coolant was fully recovered to approximately 12 m, which was mostly maintained until the end of the calculation. In both cases, the maximum fuel cladding temperature remained below 400 K, leading to neither cladding oxidation nor a collapse of the fuel assemblies and therefore no release of radionuclides into the environment by the end of the calculation. On the other hand, Fig. 12 shows the results of the LOPI-N scenario analyses for the three different injection rates made at a time of 8.2 hours (when half of the height of the fuel rods starts to be uncovered; see MATRIX-B in Table 3).

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Fig. 12 (continued)

When emergency makeup water was injected at a rate of 11.4 kg/s (30% of the maximum coolant leakage rate, LOPI-N-03-B), the level of the SFP coolant continued to decrease slightly more slowly than that of LOPI-N-00 but slightly more rapidly than that of LOPIN-03-A, resulting in the full uncovery of the fuel rods in 13.7 hours. However, the SFP coolant did not completely dry out by the end of the calculation. The SFP coolant level reached nearly 0.02 m (just above the bottom of the SFP) at the end of the calculation. During the calculation, the maximum fuel cladding temperature rose to 2350 K and 189.8 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies was maintained and they did not collapse by the end of the calculation. Nonetheless, due to the existence of a coolant leak and the extent of cladding oxidation, which exceeded the radionuclide retainability of the SFP, 2618.6 kg of radioactive Cesium and

295.5 kg Iodine (78.0% of their initial inventory levels) were released into the environment by the end of the calculation. When emergency makeup water was injected at a rate of 22.8 kg/s (60% of the maximum coolant leakage rate, LOPI-N-06-B), contrary to the LOPI-N-03-B case, the level of the SFP coolant started to recover immediately after the injection. After a partial recovery to approximately 3.65 m (only about 0.52 m below the top of the fuel rods), it was mostly retained until the end of the calculation. When emergency makeup water was injected at a rate of 38 kg/s (100% of the maximum coolant leakage rate, LOPI-N-10-B), even the level of the SFP coolant was fully recovered to nearly 12 m and later remained generally at this level until the end of the calculation. In both cases, the integrity of the fuel assemblies was maintained without a collapse by the end of the calculation. Nevertheless, in both cases, during the calculation, the maximum fuel

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Fig. 13. MELCOR analysis results for different injection rates (LOPI-R, MATRIX-A).

cladding temperature rose to 1922 K (LOPI-N-06-B) and 782 K (LOPI-N-10-B), respectively, and amounts of 105.0 kg (LOPI-N-06B) and 7.64 kg (LOPI-N-10-B) of hydrogen, respectively, were generated by cladding oxidation under steam, in contrast to the MATRIX-A cases (i.e., LOPI-N-06-A and LOPI-N-10-A). This occurred because a later injection of water than that in the MATRIX-A cases caused the upper parts of the fuel rods to become uncovered for a while. In the LOPI-N-06-B case, although the fuel assemblies maintained their integrity, 534.8 kg of radioactive Cesium and 60.4 kg of Iodine (15.9% of their initial inventory levels) were released into the environment by the end of the calculation. This is clear when considering the existence of a coolant leak and the fact that cladding oxidation exceeded the radionuclide retainability of the SFP. Meanwhile, no radionuclides were released into the environment by the end of the calculation in the LOPI-N10-B case, which had a negligible degree of cladding oxidation such that the SFP could retain all the resulting radionuclides by deposition.

3.2.4. LOPI scenarios in the refueling operation mode (LOPI-R) LOPI scenarios in the refueling mode (LOPI-R) were analyzed for a direct comparison with the corresponding LOPI-N cases. Fig. 13 shows the results of the LOPI-R scenario analysis for three different injection rates made at a time of 6.4 hours (when the fuel storage racks started to become uncovered; see MATRIX-A in Table 3).. When emergency makeup water was injected at a rate of 11.4 kg/s (30% of the maximum coolant leakage rate, LOPI-R-03-A), the level of the SFP coolant continued to decrease somewhat more slowly than that of LOPI-R-00, resulting in the full uncovery of the fuel rods in 20.2 hours and a complete dryout of the SFP coolant in 21.4 hours. During the calculation, the maximum fuel cladding temperature rose to 2375 K and 1877.7 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies in both the Ring 1 and 2 regions was lost and they finally collapsed in 40.8 hours (Ring 1) and 36.4 hours (Ring 2). While the base support assemblies in the both regions did not collapse during the end of the calculation,

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Fig. 13 (continued)

melted fuel assemblies weighing 698,543 kg were relocated to the bottom of the SFP, 2693.9 kg of hydrogen was generated from MCCI, and the concrete ablation depth reached 1.46 m by the end of the calculation. Eventually, 2140.0 kg of radioactive Cesium and 239.2 kg of Iodine (59.3% and 59.1% of the respective initial inventory levels) were released into the environment by the end of the calculation. When emergency makeup water was injected at a rate of 22.8 kg/s (60% of the maximum coolant evaporation rate, LOPI-R-06-A), contrary to the LOPI-R-03-A case, the level of the SFP coolant started to recover directly after the injection. After a partial recovery to approximately 7.43 m (about 2.71 m above the top of the fuel storage racks), the level was mostly retained until the end of the calculation.

When emergency makeup water was injected at a rate of 38 kg/ s (100% of the maximum coolant evaporation rate, LOPI-R-10-A), even the level of the SFP coolant was fully recovered at approximately 12 m, after which it held steady until the end of the calculation. In both cases, the maximum fuel cladding temperature remained below 400 K, leading to neither cladding oxidation nor a collapse of the fuel assemblies and therefore no release of radionuclides into the environment by the end of the calculation. On the other hand, Fig. 14 shows the results of the LOPI-R scenario analysis for three different injection rates made at a time of 9.1 hours (when half of the height of the fuel rods started to become uncovered; see MATRIX-B in Table 3). When emergency makeup water was injected at a rate of 11.4 kg/s (30% of the maximum coolant leakage rate, LOPI-R-03-B), the

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Fig. 14. MELCOR analysis results for different injection rates (LOPI-R, MATRIX-B).

level of the SFP coolant continually declined slightly more slowly than that of LOPI-R-00, but slightly more rapidly than that of LOPI-R-03-A, resulting in the full uncovery of the fuel rods in 16.1 hours and a complete dryout of the SFP coolant in 17.2 hours. During the calculation, the maximum fuel cladding temperature rose to 2388 K and 2727.2 kg of hydrogen was generated by cladding oxidation under steam. In this case, the integrity of the fuel assemblies in the Ring 1 region was lost and they finally collapsed in 18.8 hours. However, in contrast to the LOPI-R-03-A case, the fuel assemblies in the Ring 2 region did not collapse by the end of the calculation. As a result, the base support assemblies in both regions did not collapse by the end of the calculation. Although there were no MCCI effects and no ejection of melted fuel assemblies, due to the existence of a coolant leak and the extent of cladding oxidation, which exceeded the radionuclide retainability of the SFP, radioactive Cesium weighing 2733.6 kg and Iodine of an amount of 304.2 kg (75.7% and 75.1% of their initial inventory

levels, respectively) were released into the environment by the end of the calculation. When emergency makeup water was injected at a rate of 22.8 kg/s (60% of the maximum coolant leakage rate, LOPI-R-06-B), contrary to the LOPI-R-03-B case, the level of the SFP coolant started to recover right after the injection, After a partial recovery to approximately 7.41 m (about 2.69 m above the top of the fuel storage racks), the level was mostly retained until the end of the calculation. When emergency makeup water was injected at a rate of 38 kg/s (100% of the maximum coolant leakage rate, LOPI-R-10-B), even the level of the SFP coolant was fully recovered at approximately 12 m, after which was mostly maintained until the end of the calculation. In both cases, the integrity of the fuel assemblies was maintained without a collapse by the end of calculation. Nevertheless, in both cases, during the calculation, the maximum fuel cladding temperatures rose to 1047 K (LOPI-R-06-B) and 952 K

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Fig. 14 (continued)

(LOPI-R-10-B), respectively, and amounts of 0.80 kg (LOPI-R-06-B) and 2.50 kg (LOPI-R-10-B) of hydrogen, respectively, were generated by cladding oxidation under steam, in contrast to the MATRIX-A cases (i.e., LOPI-R-06-A and LOPI-R-10-A), as a later injection of water compared to that in the MATRIX-A cases caused the upper parts of the fuel rods to become uncovered for a while. In both cases, because the integrity of the fuel assemblies were retained and cladding oxidation was limited to a negligible extent such that the SFP could retain all the resulting radionuclides by deposition, there were no radionuclides released into the environment by the end of the calculation. 3.3. A summary of the cases analyzed Table 5 (Test MATRIX-A) summarizes the results of the MELCOR analysis (up to 7 days after accident) for (1) the two different plant

operating modes (normal and refueling), (2) the two different accident scenarios (LOCA and LOPI), and (3) different injection times and rates of emergency makeup water (corresponding to 0, 30, 60, 100% of the SFP coolant loss). According to Table 5, the fuel storage racks start to become uncovered after 116.4 hours (LOCA-N), 52.8 hours (LOCA-R), 6.1 hours (LOPI-N), and 6.4 hours (LOPI-R). Similar to earlier cases, Table 6 (Test MATRIX-B) also provides the corresponding MELCOR analysis results, form the point when half of the height of the spent fuel rods started to become uncovered (i.e., 137.6 hours for LOPI-N, 63.2 hours for LOCA-R, 8.2 hours for LOPI-N, and 9.1 hours for LOPI-R). According to Tables 5 and 6, the timings and rates for the emergency makeup water injection were found to be the key parameters for a successful mitigation. For the LOPI scenarios, emergency makeup water exceeding 60% of the maximum coolant leakage rate (i.e., 22.8 kg/s or more) should be injected to maintain the structural integrity of

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Table 5 Success criteria map for SFP SAM (up to 7 days after accident): MATRIX-A (an injection of emergency water at the time the SF racks start to become uncovered).

Note: (*) The level of the SFP coolant approached the corresponding level; (r) recovered above the corresponding level. Note: (**) Gray shade represents a successful SAM strategy for the corresponding parameters. Table 6 Success criteria map for SFP SAM (up to 7 days after accident): MATRIX-B (an injection of emergency water at the time half of the SF fuel rods start to become uncovered).

Note: (*) The level of the SFP coolant approached the corresponding level; (r) recovered above the corresponding level. Note: (**) Gray shade represents a successful SAM strategy for the corresponding parameters.

the spent fuel rods and prevent progression to a severe fuel damage accident regardless of the reactor operation modes in uses, whereas in the LOCA scenarios, successful mitigation could be achieved by injecting emergency makeup water exceeding 60% of the maximum coolant evaporation rate (i.e., 1.5 kg/s or more) in

the normal operation mode and exceeding 100% of the maximum coolant evaporation rate (i.e., 5.2 kg/s) in the refueling mode. With regard to the injection rate of the emergency makeup water, significant effects from the injection time of the emergency makeup water on accident mitigation were found when the

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delivery time was changed from the point at which the spent fuel storage racks started to become uncovered (Table 5) to when half the height of the spent fuel rods started to uncover (Table 6). For the latter injection scenarios in MATRIX-B, those injection rates cannot guarantee the successful mitigation from a conservative point of view. Despite a water injection at those rates, cladding oxidation occurred generating hydrogen and damaging the integrity of the spent fuel rods, although, in most cases, the damage was limited to a negligible extent and there was no significant amount of radionuclides released into the environment. For all test cases in the present analysis, the SFP was filled with steam, consequently leading to only a cladding oxidation under steam atmosphere rather than air atmosphere. This means that the high-exothermic Zirconium ignition (Durbin and Lindgren, 2012), which may be possible under a complete LOPI, did not take place in the present analysis. For most cases in the present analysis, a large amount of burnable gases (such as hydrogen and carbon monoxide) were generated by cladding oxidation or MCCI. Even for those cases. a deflagration, which could endanger the SFP building, is not expected to take place, due to a low concentration of oxygen (3  4% at the most) and a high steam concentration (at least 60%). The only exception is the LOPI-N-30 case, resulting in oxygen and hydrogen concentration of about 12% and 7% (MATRIX-A) and the corresponding concentration of about 12% and 10% (MATRIX-B). Under the foregoing condition, there may be a possibility of deflagration, which could endanger potentially the integrity of the SFP building. However, the present MELCOR analysis did not calculate such a deflagration of hydrogen since more sophisticated analysis is required for it. The present analysis has one case (LOPI-R-00), in which the SFP wall is failed by the radial erosion of MCCI at the end of calculation. However, it is expected that if the present analysis would extended longer the SFP wall would be failed even for both LOCA-R-00 and LOPI-N-00 cases.

4. Concluding remarks In the present study, a number of MELCOR simulations including reference and mitigation cases were carried out to assess the effectiveness and success conditions required for an emergency water injection as a typical SFP SAM strategy. The results of those simulations, based on the two test matrices shown in Table 3, MATRIX-A (an injection of emergency water at the time the fuel storage racks started to become uncovered) and MATRIX-B (an injection of emergency water when half of the height of the spent fuel rods started to become uncovered), are summarized in Tables 5 and 6, respectively. The following summarizes the relevance of the present MELCOR analysis results for the SFP SAM: (1) A SFP LOCA even in the refueling mode is a benign accident, and any small low-capacity pump (1.5–5.2 kg/s) is sufficient to mitigate a LOCA. There are large time margins (several days) and low-capacity pumps are likely sufficient to mitigate such an accident, even during the refueling phase. (2) A SFP LOPI is a much more serious accident with small time margins comparable to reactor accident time margins. For example, a seismically induced SBO + SFP LOPI would require operators to prioritize SBO and LOPI mitigation actions. Much larger equipment and water sources (possibly borated water) must be (>22.8 kg/s, preferably >38 kg/s) installed and operated during much shorter mission times. Promptly diagnosing a LOPI is critical (e.g. after a large seismic event, an operator or plant worker should visually

inspect the SFP and check the water level). Thus, it is necessary to allocate resources to studying these types of accidents and realistic initiating events leading to them. The degradation mechanisms of a monolithic SFP structure (concrete and steel liner) should be identified and inspections performed if necessary (e.g. boric acid corrosion of liner welds and water erosion of basement concrete over long periods). Table 6 shows there can be a release even if a 22.8 kg/s injection of water is achieved. (3) To prevent a failure of the SFP building by a detonation of burnable gases (such as hydrogen and carbon monoxide), which can take place in a situation like LOPI-N-30, hydrogen mitigation equipment must be installed inside the SFP building whenever necessary. When designing hydrogen mitigation systems, the system capacity, critical parameters such as a total amount of burnable gases with a time progression from cladding oxidation and MCCI, and their additional generation resulting from the SAM actions must be taken into account. (4) While an oxidation under air atmosphere did not happen in the present analysis, its possibility should be carefully investigated due to its high contribution to the SFP structure failure. (5) To prevent a failure of the SFP wall by a radial erosion of MCCI, which can take place in a few cases of the present study (i.e., LOPI-R-00, LOCA-R-00, and LOPI-N-00), much larger equipment and water sources must be provided and operated during much longer mission times than at least one week. However, more concrete conclusions regarding the efficiency of plant-level mitigation measures should be supported through further studies, which are devoted to the issues below. (1) The influence of different modeling options and user effects on simulation results to account for the prediction capability of the present MELCOR code should be considered. Typical example includes the following: (a) different thermal radiation heat transfers between hot and cold fuel assemblies in terms of view factors, (b) an additional flow path between the clad and the rack, (c) cladding oxidation under a mixture of steam, oxygen and nitrogen that drives the heating rate and onset of melting of the fuel assemblies and the amount of hydrogen production, and (d) more sophisticated numerical nodes (e.g., radial core rings employed in the MELCOR COR package). (2) Diverse simulations which take into account not only more detailed and practical times and rates of emergency makeup water injections but also the effects of other competitive mitigation strategies, such as spray systems and filtered venting systems, whenever necessary, must be considered.

Acknowledgements This work was supported by the National Research Foundation of Korea (NRF) grant funded by the Korean government (MSIP: Ministry of Science, ICT and Future Planning). References Adorni, Martina, Herranz, Luis E., Hollands, Thorsten, Ahn, Kwang-II, et al., 2016. OECD/NEA Sandia Fuel Project phase I: Benchmark of the Ignition Testing. Nucl. Eng. Des. 307, 418–430. Ahn, K.I., Shin, J.U., Kim, W.T., 2016. Severe accident analysis of plant-specific spent fuel pool to support a sfp risk and accident management. Ann. Nucl. Energy 89, 70–83.

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