This chapter provides supporting information for the present committee’s findings and recommendations in Chapter 4 (Table 4.1) related to understanding and mitigating loss-of-coolant events in spent fuel pools:
FINDING 4.5: Technical analyses undertaken by the U.S. Nuclear Regulatory Commission and Sandia National Laboratories after 2004 confirm that reconfiguring spent fuel in pools can be an effective strategy for reducing the likelihood of fuel damage and zirconium cladding fires following loss-of-pool-coolant events. However, reconfiguring spent fuel in pools does not eliminate the risks of zirconium cladding fires, particularly during certain periods following reactor shutdowns or for certain types of pool drainage conditions. These technical studies also illustrate the importance of maintaining water coolant levels in spent fuel pools so that fuel assemblies do not become uncovered.
FINDING 4.6: Additional analyses and physical experiments carried out by the U.S. Nuclear Regulatory Commission and Sandia National Laboratories since NRC (2006) was completed have substantially improved the state of knowledge of boiling water reactor (BWR) and pressurized water reactor (PWR) spent fuel behavior following partial or complete loss of pool water. These studies and experiments have addressed the following important issues:
- Fuel damage state and timing as a function of fuel age and pool water loss.
- Propagation of zirconium cladding fires to other assemblies in the pool.
- Potential mitigation strategies (dispersion of hot fuel assemblies in the pool, water sprays, water replacement) for delaying or preventing fuel damage following pool water loss.
These experiments have resulted in significant validation of the Methods for Estimation of Leakages and Consequences of Releases (MELCOR) code that is used to model coolant loss in spent fuel pools. However, the code is unable to adequately model flows when stratification occurs and plumes form in the pool and/or above-pool environment. Moreover, key portions of code lack validation, and there has been no end-to-end validation of the code for modeling coolant loss in spent fuel pools.
RECOMMENDATION 4.6 The U.S. Nuclear Regulatory Commission should (1) sponsor an end-to-end validation of the MELCOR code for use in modeling coolant loss in spent fuel pools, and (2) validate key submodels in the code with particular attention paid to
- Modeling the thermal and chemical behavior of spent fuel assemblies in partially drained pools.
- Modeling the thermal and chemical response of spent fuel assemblies to the application of water sprays.
- Modeling and validating for stratified flows in fully and partially drained pools.
FINDING 4.8: The U.S. Nuclear Regulatory Commission and the U.S. nuclear industry have made good progress in implementing actions to address Recommendation 3E-2 in NRC (2006). The U.S. Nuclear Regulatory Commission has directed plant licensees to
- Reconfigure their spent fuel in pools to achieve at least a 1 × 4 dispersion of high- and low-decay-heat assemblies, unless such configuration can be shown to be inapplicable or unachievable. This configuration must be achieved following each fuel offload from the reactor not later than 60 days after reactor shutdown.
- Develop guidance and implement strategies to maintain and restore spent fuel pool cooling following explosions and fires. To address this requirement, the U.S. nuclear industry has developed and adopted guidance and strategies for spent fuel pool water makeup and water sprays.
However, additional work is needed to more fully implement Recommendation 3E-2 in NRC (2006).
RECOMMENDATION 4.8: The U.S. Nuclear Regulatory Commission should take the following actions to more fully implement Recommendation 3E-2 in NRC (2006):
- Reexamine the need for the 60-day limit for fuel dispersion and reduce the allowable time if feasible.
- Reexamine and, if needed, redesign the water makeup and spray systems and strategies to ensure that they can be implemented when physical access to pools is hindered or the site becomes inaccessible.
This chapter is organized as follows:
- Section 6.1 provides supporting information for Findings 4.5 and 4.6 and Recommendation 4.6 on understanding the response of spent fuel pools to loss-of-coolant events.
- Section 6.2 provides supporting information for the present committee’s Finding 4.8 and Recommendation 4.8 on mitigating loss-of-coolant events in spent fuel pools.
The U.S. Nuclear Regulatory Commission (USNRC) and its contractor Sandia National Laboratories (Sandia) have expended considerable effort to understand the response of spent fuel pools to loss-of-coolant events since NRC (2006) was published:
- Physical experiments have been conducted on full-sized BWR and PWR assemblies to study fuel cladding ballooning and rapid zirconium oxidation (Lindgren and Durbin, 2007; NEA, 2015).
- Various loss-of-coolant scenarios have been examined for BWR and PWR assemblies in pools using an improved version of the MELCOR code (see Sidebar 6.1) that addressed some of the deficiencies identified in NRC (2006). Many of the publications from these studies contain security-related information and have not been released to the public. These publications are described in Appendix 6A of this chapter.
The committee finds (Finding 4.6; see Table 4.1 in Chapter 4) that these analyses and physical experiments have substantially improved the state of knowledge of BWR and PWR spent fuel behavior following partial or complete loss of pool water. These analyses and experiments addressed three of the questions raised in Recommendation 3E-1 (NRC, 2006; see Table 4.1 in Chapter 4):
- Is it feasible to reconfigure the spent fuel within pools to prevent zirconium cladding fires1 given the actual characteristics (i.e., heat
1 As noted in Chapter 4, the term zirconium cladding fire is used to describe the self-sustaining oxidation of zirconium fuel cladding. This oxidation results in a temperature runaway that can generate enough heat to melt the fuel pellets.
- In the event of a localized zirconium cladding fire, will such rearrangement prevent its spread to the rest of the pool?
- How much spray cooling is needed to prevent zirconium cladding fires and prevent propagation of such fires?
generation) of spent fuel assemblies in the pool, even if the fuel were damaged in an attack?
The committee finds (Finding 4.5; see Table 4.1 in Chapter 4) that technical analyses undertaken by the U.S. Nuclear Regulatory Commission and Sandia National Laboratories confirm that reconfiguring spent fuel in pools
can be an effective strategy for reducing the likelihood of fuel damage and zirconium cladding fires following loss-of-pool-coolant events. Indeed, the analyses indicate that fuel configuration in a pool can affect its coolability2 in air following a loss-of-pool-coolant event. There are certain pool water states and time periods when the fuel is not coolable in air and is therefore subject to damage (e.g., fuel rod ballooning) and rapid oxidation (i.e., zirconium cladding fires), regardless of its configuration in the pool.
The following sections provide brief summaries of this USNRC and Sandia work. Some details of the analyses, for example, the particular conditions under which fuel damage and zirconium cladding fires can occur, as well as the timing of such occurrences, are not provided in this report because they are security sensitive.
6.1.1 Physical Experiments
In the late 1990s, USNRC staff evaluated spent fuel pool accident risks at decommissioned nuclear plants in the United States. The results of this evaluation are documented in Collins and Hubbard (2001). Some assumptions in the accident progression were known to be conservative, especially fuel damage estimates. The USNRC subsequently initiated efforts to evaluate severe accident progression in spent fuel pools at operating plants using best-estimate computer codes. These code calculations identified various modeling and phenomenological uncertainties.
Following the release of the NRC (2004) report, the USNRC initiated an experimental program at Sandia to investigate thermal-hydraulic phenomena associated with complete loss-of-coolant accidents in spent fuel pools in light-water reactors. The ultimate objective of this program was to simulate accident conditions of interest for the spent fuel pool in a full-scale prototypic fashion (electrically heated, prototypic assemblies in a prototypic spent fuel pool rack). A major impetus for this work was to allow code validation (primarily MELCOR) and reduce modeling uncertainties. The results of this work are documented by Lindgren and Durbin (2007).
As shown in Table 6.1, Sandia used a phased approach with three basic types of experiments to complete this program. As a proof of concept, two heater-design tests were first performed to determine the suitability of the electrically heated, zirconium-clad spent fuel rod simulators. Three separate-effects tests were then conducted to study specific phenomena
2 The committee uses the term coolability to indicate whether air cooling of stored fuel is sufficient to prevent self-sustaining (runaway) oxidation of its zirconium cladding. Runaway oxidation may occur in the range 900°C-1200°C depending on thermal-hydraulic conditions. Fuel that is coolable in air will not reach these temperatures. Section 6.1.3 in this chapter describes the reactions that occur if fuel reaches this temperature.
TABLE 6.1 Description of Tests in Lindgren and Durbin (2007)
|Test Description||Purpose||Assembly||Rod Material|
|Heater Design||Test electrical heater performance, preliminary data on zirconium fire; conducted at normal and reduced oxygen concentrations||12-rod bundle||Zircaloy|
Hydraulics: Determine viscous and form loss coefficients for laminar volumetric flow rates
Thermal hydraulics: Determine input conditions for partial-length experiments
Thermal radiation: Determine radiation coupling in a 1 × 4 arrangement
Prototypic: Partial length
Axial ignition: Determine temperature profiles, induced flow, axial O2 profile, nature of rapid zirconium oxidation
Determine radial fire propagation in a 1 × 4 arrangement
Prototypic: Single full-length assembly
such as heat transfer (e.g., thermal radiative coupling) and fluid flow (e.g., induced natural convective flow). These tests were nondestructive and involved some nonprototypic materials (e.g., stainless steel and Incoloy). Finally, two prototypic assemblies were heated to ignition in a series of integral-effects tests.
The heater-design tests were conducted with a 12 × 12 rod bundle configuration with zirconium cladding. Sandia researchers demonstrated that ignition was possible when the test design minimized heat loss and maximized gas preheating and bundle power.
The separate-effects tests utilized a single full-length or partial-length “highly prototypic” BWR 9 × 9 fuel rod assembly. Sandia researchers measured the thermal-hydraulic response and determined appropriate loss coefficients as a function of bundle mass flow under adiabatic conditions.
The integral-effects tests used five one-third-length zirconium fuel assemblies arranged in a 1 × 4 configuration (i.e., a center assembly and four face assemblies; see Figure 7.2 in Chapter 7) in a 3 × 3 pool rack. The tests were designed to simulate the middle to upper portions of an array of full-length assemblies.
An important aspect of this project was the deliberate close coupling of experiments with numerical analysis. The project utilized the severe accident code MELCOR as (1) a tool for the experimental design, (2) for pretest results prediction, and (3) for post-test analysis of the calculated
and measured responses. The post-test MELCOR analysis helped refine some model parameters, which led to improvements in the fidelity of model predictions.
6.1.2 MELCOR Analyses
The MELCOR code (Sidebar 6.1) was originally developed for analysis of severe accidents in BWR and PWR reactors. The code is based on control volumes connected by flow paths that are treated as “pipes” with specified frictional losses. Flow areas are either given or are determined as degraded material relocates. Generally, one-dimensional (1D) flows are considered. However, the code can be exercised in a pseudo-2D manner to calculate radial flow rates.
MELCOR has been updated for use in investigating conditions in spent fuel pools under various loss-of-coolant conditions. The updated version of MELCOR includes models for
- Fuel degradation;
- Radiative, convective, and conductive heat transfer;
- Air and steam oxidation;
- Hydrogen production and combustion;
- Boiling and two-phase thermal hydraulics; and
- Fission-product release and transport.
Recent enhancements to the code include a new air-oxidation kinetics model and a new spent fuel assembly flow-resistance model based on the experiments described in the previous section of this chapter. Nevertheless, the MELCOR model still has several limitations:
- It cannot model stratified flow or buoyancy-driven flow (i.e., formation of plumes and circulatory flow patterns in large spent fuel pools or above-pool environments) or open-rack configurations in which fuel assemblies are not contained in solid-wall boxes as is current practice in dense-packed pools.
- It cannot model two-phase flow and boiling heat transfer when structural debris falls into the fuel assemblies in the pool.
- It cannot model nitriding reactions with zirconium.
- For spray-cooling scenarios (i.e., when water is sprayed onto the tops of the fuel assemblies in the pool to cool them), it cannot model entrainment and deentrainment of water droplets.3 Droplet
3 Wallis’ correlation is used for the flooding limit, but spray penetration depth calculations are suspect.
behavior during travel from the spray nozzle to the fuel assemblies is not modeled, and modeling of heat transfer from the cladding to the impacting droplets is lacking.
- It cannot model the simultaneous flow of air, steam, or water droplets through fuel assemblies during spray cooling of a fully drained pool.
- Fuel cladding degradation models are empirical and are based on user-specified criteria.
MELCOR has been used by Sandia to analyze complete and partial loss-of-coolant scenarios at the fuel assembly and whole-pool level. These analyses show that several factors affect spent fuel coolability, including
- Fuel aging time,
- Fuel configuration in the pool,
- Size and location of coolant leaks in the pool,
- Ventilation above the pool,
- Radial thermal coupling, and
- Deformation of the fuel rod bundle geometry.
These factors are described below.
220.127.116.11 Fuel Aging Time
Fuel age refers to the elapsed time since the fuel was in an operating reactor. Once the reactor is shut down, heat production from radioactive decay in the fuel decreases rapidly. In general, the older the fuel, the less heat it generates. If the fuel is offloaded from the reactor to the pool, there is a certain period of time after offload during which the fuel is not coolable in air, even if the pool is completely drained (see, for example, USNRC , which is discussed in Chapter 7).
18.104.22.168 Fuel Configuration in the Pool
Sandia analysts used MELCOR to examine five configurations of fuel storage in spent fuel pools:
- A uniform configuration in which all of the fuel assemblies from a reactor offload are grouped together in the pool;
- A 1 × 4 configuration in which a hotter fuel assembly is surrounded on four sides with colder assemblies;
- A 1 × 4 configuration in which some of the surrounding cold assemblies are missing (i.e., the spent fuel rack is empty in some locations);
- A checkerboard configuration of hot and cold assemblies; and
- A checkerboard configuration where some of the surrounding cold assemblies are missing.
The 1 × 4 and checkerboard configurations are illustrated in Figure 7.2 in Chapter 7. When higher-decay-heat (i.e., hotter) assemblies are surrounded by lower-decay-heat (i.e., colder) assemblies, the temperature rise of the hotter assemblies is slowed, mostly because of heat loss by radiation to the colder assemblies and their thermal inertia. In other words, the thermal capacities of the colder assemblies play an important role in regulating temperature rise of the hotter assemblies.
The Sandia analyses showed that, at higher temperatures, the heat-up rate of 1 × 4 configurations is slower compared to the other configurations studied. They also showed that zirconium cladding fires, once initiated, can spread to other assemblies in the pool. At high temperatures, fuel rod integrity is lost and fuel material is rearranged into debris piles or molten pools. Radioactive materials are released from the fuel as noble gases and aerosols (see Sidebar 2.2 in Chapter 2).
22.214.171.124 Size and Location of Coolant Leaks in the Pool
Development of an uncontrolled leak on the pool boundary can lead to the loss of pool coolant. The location of the leak determines whether the pool will drain partially or completely. The size and location of the leak will determine how long it will take to drain the pool.
Sandia analysts used MELCOR to investigate the effect of leak size and location on air coolability limits of BWR and PWR assemblies stored in spent fuel pools. Two scenarios were considered:
- Complete loss of coolant, where the pool was assumed to drain below the base plate of the spent fuel racks, allowing for natural convection of air through the assemblies, and
- Partial loss of coolant, where the pool was not drained completely, so water covered the lower portions of the fuel assemblies. This blocked airflow through the assemblies until water in the lower portion of the pool boiled off.
For both scenarios, it was assumed that spent fuel cooling and building ventilation systems were not functioning and that no water was being added to the pool.
126.96.36.199 Ventilation Above the Pool
The Sandia analyses showed that ventilation of the enclosed space above a spent fuel pool affects the coolability of stored fuel in air. Under poor ventilation conditions (e.g., when the building ventilation system is not functioning) in a fully drained pool, stored fuel of a certain age4 can undergo self-sustaining oxidation (zirconium cladding fire) that can spread to other assemblies in the pool and release fission products to the environment. However, fission-product aerosols are mostly retained in the pool enclosure as long as it is not damaged. If the enclosure is damaged, for example by a hydrogen explosion, then fewer aerosols will be retained. Under good ventilation conditions (e.g., the blowout panels above the spent fuel pool are open), the fuel may be coolable in air under complete loss-of-coolant conditions, again depending on its age, but the fuel may not be coolable in air under partial-loss-of-coolant conditions.
188.8.131.52 Radial Thermal Coupling
Heat transfer from hot fuel assemblies to adjacent cold assemblies in a spent fuel pool plays an important role in the propagation of zirconium fires and the consequent release of radioactive material from the fuel. Radiation is the dominant mode of heat transfer at high temperatures, so accuracy in the calculation of radiative heat transfer is important for estimating assembly temperatures. The limiting condition occurs when a hot assembly is totally isolated from surrounding colder assemblies. In this case, the heat-up rate of the hot assembly will be much higher, leading to early fuel damage and radioactive material releases. In the absence of heat gain from the hot assemblies, the heat-up rate of the colder fuel assemblies will be lower, limiting the spread of a zirconium cladding fire from hotter to colder assemblies. MELCOR sensitivity analyses of fuel heat-up have been carried out for three different radial thermal coupling configurations (i.e., no coupling, normal coupling, and reduced coupling).
184.108.40.206 Deformation of the Fuel Rod Bundle Geometry
Internal pressures in the fuel rods can cause localized ballooning as mechanical properties of the cladding degrade with increases in temperature. The occurrence of ballooning can increase flow resistance within the fuel assembly and impair cooling, possibly leading to further cladding degradation. Clad ballooning cannot be modeled mechanistically using
4 The exact ages of the fuel for this condition is security-related information and therefore not disclosed in this report.
MELCOR but is instead modeled parametrically by reducing the flow area. A co-planar blockage is created at the exit of the rod bundle when the peak temperature exceeds an established criterion for the onset of ballooning. A more than minor reduction in flow area,5 which occurs when neighboring fuel rods in an assembly are in physical contact, will produce a small increase in the calculated maximum cladding temperature but will have little effect on the initiation of cladding oxidation.
Crushing of fuel rods due to mechanical loading can also reduce flow areas and impose additional flow resistance. MELCOR was used to study the effects of crushing on the coolability of spent fuel in air for a completely drained pool. Crushing was assumed to occur along the entire length of the fuel rod, and flow area was reduced parametrically. The reduced flow area allows temperatures in the cladding to increase sufficiently to cause self-sustaining oxidation, even in older spent fuel. However, when flow-area reductions are high,6 the reduced rate of air flow through the assemblies limits the extent of oxidation and thus slows the rise of cladding temperatures, delaying the onset of self-sustaining oxidation conditions.
A reduction in flow area and accompanied increase in the flow resistance can also occur if structural debris falls onto the spent fuel (such as occurred at the Fukushima Daiichi plant; see Chapter 2). The flow resistance and thermal insulation created by this debris can reduce the coolability of the stored fuel and accelerate fuel rod heat-up. To the committee’s knowledge, this scenario has not been evaluated by the USNRC or Sandia.
Sandia has carried out hand calculations to assess the coolability of debris beds formed from the relocation of degraded fuel rods, cladding, and structural material to the bottom of a spent fuel pool. Both liquid-saturated debris beds with an overlying layer of liquid and dry debris beds have been considered. The coolability limit strongly depends on effective particle diameter, porosity, and depth of the debris bed. Calculations have also been carried out for a dried-out debris bed that is cooled by the flow of air. The coolability of the fuel in the debris bed depends on the bed’s effective particle diameter and porosity.
6.1.3 Spent Fuel Pool Loss-of-Coolant Accidents (LOCAs)
In a complete-loss-of-pool-coolant scenario, most of the oxidation of zirconium cladding occurs in an air environment:
5 The specific percentage is security-related information.
6 The specific percentage is security-related information.
For a partial-loss-of-pool-coolant scenario (or slow drainage in a complete-loss-of-pool-coolant scenario), the initial oxidation of zirconium cladding will occur in a steam environment:
Both of these reactions are highly exothermic. The zirconium-steam reaction leads to the formation of hydrogen, which can undergo rapid deflagration in the pool enclosure, resulting in overpressures and structural damage. This damage can provide a pathway for air ingress to the pool, which can promote further zirconium oxidation and allow radioactive materials to be released into the environment. Debris from the damaged enclosure can fall into the pool and block coolant passages.
The MELCOR analyses show that fuel in a completely drained pool can be more easily cooled by air than in a partially drained pool. Coolability in a completely drained pool is promoted by the establishment of natural air circulation through the fuel assemblies.
In a partially drained pool, cooling of the uncovered portions of fuel assemblies occurs mainly by steam that is generated as the pool water boils off. The steam production rate depends on the decay-heat generation rate in the fuel and the portion of the fuel rods that are covered with the two-phase mixture of steam and water. The rate affects the temperature rise in the dry regions of the rods and the rate of hydrogen production as a result of the zirconium-steam reaction. The exothermic zirconium-steam reaction also adds heat to the fuel rod.
After the water level drops below the rack base plate, convective air flow is established. If the steam is exhausted, then the zirconium-steam reaction is replaced by the zirconium-oxygen reaction. However, prior to the onset of convective air flow, fuel cladding temperatures can exceed the threshold for oxidation, and fuel damage and radioactive material release can occur. The time to damage and release depends on pool water depth relative to the stored fuel assemblies.
Significant validation of the MELCOR code and improvement of thermal, chemical, and mechanical models has been carried out with experiments on electrically heated rods to examine the behavior of spent fuel assembly heat-up, oxidation, and loss of coolable geometry in air environments corresponding to a completely drained pool. However, no experimental validation of these phenomena has been performed for a partially drained pool. The present committee’s Recommendation 4.6 (see Table 4.1 in Chapter 4) calls for validation of the thermal and chemical behavior of spent fuel assemblies in partially drained pools. This validation is important because, as noted previously, there is a higher hazard for zirconium cladding fires in partially drained pools.
6.1.4 Mitigation Strategies for Spent Fuel Pool LOCA
At least two mitigation strategies are available to mitigate a loss-of-coolant event in a spent fuel pool: Repair the leak that is causing water to be lost and/or add makeup water. In the absence of leak repair, the location, size of the leak, magnitude of decay heat in the pool, and rate and timing of makeup water addition determines the effectiveness of the mitigation strategy. Sandia used MELCOR to determine the desired flow rate of makeup water under different accident scenarios. In the case where a spent fuel pool drains completely, adding makeup water will cover the lower portions of the fuel assemblies and block air convection. This could lead to heat-up of the fuel and production of hydrogen as a result of zirconium-steam reaction, loss of coolable geometry in the assembly, and eventually self-sustaining oxidation of zirconium. The flow of makeup water must be high enough to cover a certain portion7 of the active fuel height before these conditions occur.
Spraying water on top of the fuel assemblies may also be an effective strategy to provide additional cooling if makeup water capabilities are inadequate to maintain pool water levels above the tops of the fuel racks. Sandia used MELCOR to study the effectiveness of spraying with a certain flow rate and delay time after a loss-of-pool-coolant event. Analysts found that, for certain fuel configurations,8 spraying the fuel can be an effective strategy for maintaining coolability of the fuel. To the committee’s knowledge, no experimental verification of MELCOR calculations with respect to droplet size, effect of counterflowing steam and/or gases on droplet carryover, and wetting of surface by droplets and associated heat transfer has been provided.
The present committee’s Recommendation 4.6 (see Table 4.1 in Chapter 4) calls for the validation of the thermal and chemical response of spent fuel assemblies to the application of water sprays. Analysis should be carried to determine the envelope of water flow-rate conditions, either as water makeup or spray, for fuel assembly coolability. The presence of debris in the pool, which can block water sprays, should be considered in the assessment.
Validating the responses of stored fuel to water sprays is essential for confirming the effectiveness of existing mitigation capabilities for loss-of-coolant events in spent fuel pools. These capabilities are discussed in the next section.
7 The exact height is security-related information and is not disclosed in this report.
8 The exact configuration is security-related information and is not disclosed in this report.
NRC (2006) recommended that the USNRC ensure that nuclear plant operators take prompt and effective measures to reduce the consequences of loss-of-pool-coolant events that could result in propagating zirconium cladding fires. Two specific measures were recommended for prompt implementation (Recommendation 3E-2 [NRC, 2006]; see Table 4.1 in Chapter 4):
- Reconfigure the fuel in the pools so that high-decay-heat fuel assemblies are surrounded by low-decay-heat assemblies, and
- Make provision for water-spray systems that would be able to cool the fuel even if the pool or overlying building were severely damaged.
The present committee finds (Finding 4.8; see Table 4.1 in Chapter 4) that the USNRC and the U.S. nuclear industry have made good progress in implementing this NRC recommendation. The USNRC has directed plant licensees to
- Reconfigure their spent fuel in pools to achieve at least a 1 × 4 dispersion of high- and low-decay-heat assemblies (see Figure 7.2 in Chapter 7), unless such configuration can be shown to be inapplicable or unachievable. This configuration must be achieved following each fuel offload from the reactor not later than 60 days after reactor shutdown.
- Develop guidance and implement strategies to maintain and restore spent fuel pool cooling following explosions and fires.
The USNRC informed the committee that most U.S. nuclear plants have implemented the first recommended action. Additionally, the U.S. nuclear industry has developed and adopted guidance and strategies to implement the second recommended action. Further discussion is provided below.
Following the September 11, 2001, terrorist attacks on the United States, the USNRC issued Order EA-02-026 (Order for Interim Safeguards and Security Compensatory Measures).9 Section B.5.b of the order directed nuclear plant licensees to develop and implement strategies to maintain or restore core, containment, and spent fuel pool cooling capabilities following large explosions or fires that damaged large areas of the plant. The order
9 The order is designated as Safeguards Information and has not been released to the public, but its requirements have been codified in 10 CFR 50.54(hh)(2).
directed licensees to identify mitigation measures that could be implemented with resources already existing or readily available at the plant, including strategies for fire-fighting, operations to minimize and mitigate fuel damage, and actions to minimize radiological releases.
Parallel but separate plant-specific studies were carried out by the USNRC and the nuclear industry to identify readily available resources to mitigate damage to spent fuel pools and nearby areas from large explosions and fires. The plant conditions evaluated in these site-specific assessments were beyond design basis. The assessments utilized a threat-independent methodology to identify potential plant-specific strategies for preventing or mitigating damage to reactors and spent fuel pools. As described in the industry-sponsored assessment report NEI 06-1210 (NEI, 2009), the overall strategy involves a diverse capability within plants to provide at least 500 gallons per minute (gpm) of makeup water to the plant’s spent fuel pools for 12 hours.11 The balance of the strategy involves the use of a portable spent fuel pool makeup capability as well as a 200 gpm spray capability from that same water source to enhance the robustness and flexibility of site responses.
The Fukushima Daiichi accident renewed and heightened interest in the potential vulnerability of spent fuel pools to extreme natural events (see Chapter 2 of this report). The USNRC issued Order EA-12-049 (USNRC, 2012a), which directed nuclear plant licensees to develop, implement, and maintain guidance and strategies to maintain or restore core cooling, containment, and spent fuel pool cooling capabilities following a beyond-design-basis event. The industry responded with the Diverse and Flexible Coping Strategies (FLEX) initiative (NEI, 2012; see also NRC, 2014, Appendix F). The USNRC subsequently endorsed the industry’s FLEX initiative for meeting the Order (see USNRC  for the latest guidance). The FLEX initiative is designed to increase defense-in-depth for beyond-design-basis accident scenarios, including the extended loss of AC power and ultimate heat sink at multiple units on a site.
The objectives of FLEX are to establish resources and associated procedures for an indefinite coping capability to prevent damage to fuel in the
10 NEI 06-12 is USNRC-endorsed guidance for meeting the requirements of 10 CFR 50.54(hh)(2).
11 This makeup capability can be used to replace water lost from the pool due to breaches in the pool boundary or from evaporation. (A 500-gpm flow rate approximately corresponds to the drainage rate resulting from an 8-cm-diameter hole in the pool wall 5 m below the pool’s free surface.) The water makeup requirements for evaporative losses are well below 500 gpm. For example, only about 40 gpm of water would have been required to remove decay heat in the Unit 4 pool at the Fukushima Daiichi plant 10 days after the entire core was offloaded to the pool.
reactor and spent fuel pool and maintain containment functions.12 The operator would first use installed equipment, if available, to meet these goals. If such equipment is not available, then operators would provide makeup water (e.g., from the condensate storage tank) with a portable injection source (pump, flexible hoses to standard connections, and associated diesel engine-generator) that can provide at least 500 gpm of spent fuel pool makeup. The portable equipment would be staged on site and could also be brought in from regional staging facilities.
If pool water levels cannot be maintained above the tops of the fuel assemblies, then portable pumps and nozzles would be used to spray water on the uncovered fuel assemblies. FLEX requires a minimum of 200 gpm to be sprayed onto the tops of the fuel assemblies to cool them (NEI, 2012).13 These FLEX water flow requirements are consistent with the requirements in 10 CFR 50.54(hh)(2). The installed FLEX equipment has also been shown to have substantial margin (60 percent) above the required flow based on equipment testing.
Although good progress has been made by the USNRC and the U.S. nuclear industry in implementing Recommendation 3E-2 in NRC (2006), the committee recommends two additional actions (Recommendation 4.8; see Table 4.1 in Chapter 4) be taken:
- Reexamine the need for the 60-day limit for fuel dispersion after reactor shutdown and reduce the allowable time if feasible. It is not clear what the technical basis is for this 60-day time period. Not all spent fuel is air-coolable within 60 days of reactor shutdown. The USNRC’s Spent Fuel Study (described in Chapter 7 of this report), for example, shows that spent fuel cannot be air cooled for about the first 30 days after its offload from a reactor regardless of fuel configuration in the pool. A shorter-than-60-day time limit may provide an improved safety posture if it is operationally feasible and does not increase the risk of other types of spent fuel pool accidents. A risk-informed examination of this issue could have safety benefits.
- Reexamine and, if needed, redesign the water makeup and spray systems and strategies to ensure that they can be implemented when physical access to pools is hindered, for example, by structural damage and/or radiation levels or the site becomes inaccessible. The FLEX strategy for spent fuel pool cooling assumes that
12 Dry-cask storage facilities are outside the scope of FLEX.
13 USNRC (2016) notes that the FLEX portable spray capability (utilizing portable spray nozzles from the refueling floor with portable pumps) is not required when a pool is located below grade or when a seismic hazard analysis shows that the pool will maintain its integrity.
workers will have physical access to the pools to install hoses and spray nozzles if permanently installed equipment is damaged. However, physical access might not be possible if the building is damaged or the pool is drained (in the latter case, high radiation levels would likely limit physical access to the pool). The spent fuel pools in Units 1-4 of the Fukushima Daiichi plant were not accessible after the hydrogen explosions because of debris and high radiation levels.
This Appendix describes a series of USNRC-sponsored and Sandia-executed technical studies that were carried out to improve the understanding of loss-of-cooling accidents in spent fuel pools. These studies were initiated following completion of NRC (2004) and were published between 2006 and 2008. The publications have not been released to the public because they contain security-related information. However, the committee was provided copies of these publications by the USNRC and also received detailed briefings from the USNRC and Sandia on some of this work.
Summaries of these technical studies are provided below to demonstrate the breadth of information that committee considered outside of publicly available documents. These summaries omit technical details that are considered by the USNRC to be security sensitive.
Mitigation of Spent Fuel Pool Loss-of-Coolant Inventory Accidents and Extension of Reference Plant Analyses to Other Spent Fuel Pools
K.C. Wagner and R.O. Gauntt, SANDIA Letter Report, Revision 2, Completed November 2006; 121 pages.
Simulations were carried out to examine mitigation strategies for representative spent fuel pools in PWR and BWR reference plants. The following mitigation strategies were examined: makeup water, pool leak repair, fuel dispersion, emergency sprays, building ventilation, and pool configuration.
Analysis of BWR Spent Fuel Pool Flow Patterns Using Computational Fluid Dynamics: Supplemental Air Cases
K.C. Wagner and R.O. Gauntt, SANDIA Letter Report, Revision 3, Completed January 2008; 63 pages.
This study is a follow-up to earlier studies (e.g., NUREG-1726), which used computational fluid dynamics (FLUENT) to simulate air-flow patterns in drained spent fuel pools. This study used FLOW 3-D to model natural air convection in a fully drained BWR (Mark I) spent fuel pool. A porous media model for the fuel racks and fuel was used together with a computational fluid dynamics model of the pool and reactor building. Air-flow patterns within and outside the fuel were simulated. The effect of open areas in the pool (i.e., areas of the pool with no racking) on peak air temperatures was examined in a series of parametric computations. Fuel
oxidation or mechanical response was not modeled, so no conclusions were reached regarding fuel damage or zirconium fires.
Analysis of Emergency Spray Mitigation of Spent Fuel Pool Loss-of-Coolant Inventory Accidents
K.C. Wagner and R.O. Gauntt, SANDIA Letter Report, Revision 2, Completed January 2008; 115 pages.
MELCOR was used to simulate water sprays in mitigating loss-of-coolant accidents in a BWR (Mark I) spent fuel pool. Parametric studies were performed to examine the effects of pool leak location and size, spray flow rate, and fuel arrangement pattern (uniform, 1 × 4, and checkerboard; see Figure 7.2 in Chapter 7). Steam and air oxidation as well as hydrogen generation and combustion were modeled. The time to fuel ignition was determined as a function of fuel age (i.e., time since reactor shutdown). Scenarios with water levels above and below the baseplate of the fuel racks were considered. Experimental measurements of pressure drops in a prototypic BWR fuel assembly were used to provide realistic correlations for flow resistance within the fuel bundle.
Additional MELCOR Analyses of BWR Spent Fuel Pool Assembly Accident Response
K.C. Wagner and R.O. Gauntt, SANDIA Letter Report, Revision 3, Completed June 2008; 206 pages.
MELCOR was used to simulate full and partial loss-of-coolant accidents in a spent fuel pool in a reference BWR (Mark I) plant. Several improvements were made to MELCOR for these analyses, including increasing the fidelity of the rack model, oxidation kinetics, a hydraulic resistance model, and the physical modeling of the fuel assembly and racks. A parametric study of decay heat, “bypass” flow, oxidation-layer thickness, and fuel configuration, including checkerboard, 1 × 4, and uniform configuration cases (see Figure 7.2 in Chapter 7), was performed. Parametric computations were carried out to determine the effect of fuel age and distribution of fuel in the pool on coolability under complete- and partial-loss-of-coolant conditions. The time until fuel ignition (i.e., runaway zirconium oxidation) was quantified for individual assemblies, and source-term computations were carried out for the entire pool. The ability to cool a debris bed with water and air were investigated. The effect of removing the BWR fuel assembly channel on ignition time was also investigated. A simplified analysis of fuel ballooning was carried out.
Evaluation of a BWR Spent Fuel Pool Accident Response to Loss-of-Coolant Inventory Scenarios Using MELCOR 1.8.5
K.C. Wagner and R.O. Gauntt, SANDIA Letter Report, Revision 3, Completed June 2008; 172 pages (Part I), 175 pages (Part II).
MELCOR was used to examine full- and partial-loss-of-pool-coolant accidents in a BWR (Mark I) reference plant when there is no mitigation and building ventilation is lost. Parameters examined included decay heat, radial thermal coupling scheme, the leakage hole size (small, medium, and large), and an open (i.e., blowout panel removed) versus a closed reactor building. The effect of hydrogen combustion was investigated for the partial-loss-of-pool-coolant cases. Radioactive material releases from the stored fuel were estimated.
Evaluation of a PWR Spent Fuel Pool Accident Response to Loss-of-Coolant Inventory Scenarios Using MELCOR 1.8.5
K.C. Wagner and R.O. Gauntt, SANDIA Letter Report, Revision 4, Completed June 2008; 156 pages.
MELCOR was used to simulate complete- and partial-loss-of-coolant accidents in a PWR spent fuel pool. For complete loss of coolant, the peak cladding temperature history depends on radial thermal coupling scheme, fuel age, building ventilation rate, and flow resistance of the fuel assemblies. The effect of leak size, crushing of fuel, rod ballooning, and reduced radial thermal coupling were examined through sensitivity calculations. A few simulations were used to compute the magnitude of fission product releases outside the spent fuel pool building.
MELCOR 1.8.5 Separate Effects Analyses of PWR Spent Fuel Pool Assembly Accident Response
K.C. Wagner and R.O. Gauntt, SANDIA Letter Report, Revision 4, Completed June 2008; 100 pages, +100 pages for part 2.
MELCOR was used to examine the effects of varying the following parameters on peak cladding temperatures within fuel assemblies of a single PWR spent fuel pool: decay heat, gas speed and temperature, oxide-layer thickness, flow resistance, rod ballooning, oxidation kinetics, rack configuration, and water level. For multiple fuel assemblies, the effects of fuel dispersion, including 1 × 4 and checkerboard (see Figure 7.2 in Chapter 7), and the location of empty cells were examined. For full pools, the effects of drain-down time on peak cladding temperature history were also examined. The effect of each of these parameters on the peak cladding temperature
history was ranked in order of impact. The minimum fuel age required to prevent ignition was determined for each fuel configuration.
Investigations of Zirconium Fires during Spent Fuel Pool LOCAs: PWR Assemblies
Division of System Analysis, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission; June 24, 2013; presentation to committee; 23 pages.
This presentation covered the results of Sandia experiments on PWR bundles. Testing used prototypic assemblies to measure hydraulic resistance to air flow to calibrate the MELCOR model used to model air cooling of fully drained pools. Internal electrical heating was used to simulate decay heat with low power and tests were carried out to determine fuel bundle temperatures and air-flow rates as a function of time prior to ignition. Tests were carried out at higher power levels that resulted in ignition. Measurements were made of flow rate, peak cladding temperatures, oxygen concentration in the exit flow, and time to ignition. Tests were carried out to simulate both the 1 × 4 and uniform loading arrangements that are used for spent fuel in pools. MELCOR simulations had been completed for single-assembly tests and were in progress for testing simulating multiple assemblies. Portions of the Sandia work were carried out in the context of an Organisation for Economic Co-operation and Development/Nuclear Energy Agency (NEA) Sandia Fuel Project and some of the results are now available publicly in NEA (2015).