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Study on mitigation of in-vessel release of fission products in severe accidents of PWR
Nuclear Engineering and Design 240 (2010) 3888–3897
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Study on mitigation of in-vessel release of fission products in severe accidents of PWR G.F. Huang ∗ , L.L. Tong, J.X. Li, X.W. Cao School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200240, China
a r t i c l e
i n f o
Article history: Received 9 April 2010 Received in revised form 10 August 2010 Accepted 20 August 2010
a b s t r a c t During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, including in-vessel and ex-vessel release. Mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. Mitigation countermeasures to in-vessel release are studied for Chinese 600 MW pressurized water reactor (PWR), including feed-andbleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling. SBO, LOFW, SBLOCA and LBLOCA are selected as typical severe accident sequences. Based on the evaluation of in-vessel release with different startup time of countermeasure, and the coupling relationship between thermohydraulics and in-vessel release of fission products, some results are achieved. Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products, and earlier startup time of countermeasure is more feasible. Feed-and-bleed in secondary circuit is also an effective countermeasure to mitigate in-vessel release for most severe accident sequences that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Ex-vessel cooling has no mitigation effect on in-vessel release owing to inevitable core melt and relocation. © 2010 Elsevier B.V. All rights reserved.
1. Introduction During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, which includes in-vessel and ex-vessel release (Denning et al., 1986). In-vessel release is occurred inside reactor pressure vessel (RPV) during degradation of core, ex-vessel release is occurred outside RPV after RPV is failed. For in-vessel release, released fission products transport into containment through primary system (Wright et al., 1994). If containment has great leakage, much fission products will release into environment, which is contributed to severe accident source term (Lee et al., 2008). This may induce health risk of public. So, mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. If integrity of RPV is maintained and in-vessel release of fission products is mitigated, radiological consequence will be alleviated and public health will be protected in the early phase of severe accidents. Strategies that aim at preventing or mitigating core damage all focus on reducing the fuel temperature. To the extent that these strategies do cool the fuel or slow the temperature rise. They may
∗ Corresponding author. Tel.: +86 21 34205495; fax: +86 21 34205495. E-mail addresses: [email protected] (G.F. Huang), [email protected] (X.W. Cao). 0029-5493/$ – see front matter © 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2010.08.010
also mitigate in-vessel release of fission products that is very dependent on fuel temperature (NEA, 2000). For instance, addition of coolant has a profound effect on mitigation of in-vessel release of fission products even if additive coolant is insufficient to cool the fuel. The additive coolant may limit the amounts of radioactive material released from the fuel that passes through the reactor coolant system (RCS) and reaches the containment. Mitigation of in-vessel release of fission products during severe accidents can be achieved by implementing some severe accident countermeasures. Many in-pile and out-of-pile experiments are conducted to better understand fission product release behavior in severe accidents. In-pile experiments include: the Phebus fission product and severe fuel damage (SFD) tests (Dubourg et al., 2005), the power burst facility (PBF) SFD tests (Martinson et al., 1986), the source term experiments project (STEP) (Baker et al., 1988), etc. Out-of-pile experiments include: HEVA program (Leveque et al., 1994), VERCOR program (Andre et al., 1996), verification experiments of radionuclides gas/aerosol release (VEGA) program (Kudo et al., 2001), etc. Most of them are concentrated on fission product release behavior in the accident environment, and fission product release models are developed based on these experiments. However, the research to effect of reflood/quenching on fission products release is relatively less, especially for the actual severe accident management countermeasures. Thus, the study of effect of severe accident management countermeasures on fission products release is needed for actual nuclear power plant.
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Fig. 2. RCS model. Fig. 1. Sketch map of nuclear island and safety system of 600 MW PWR.
This paper presents analyses of in-vessel release of fission products in severe accidents for Chinese 600 MW pressurized water reactor (PWR). Mitigation effect of feed-and-bleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling on in-vessel release of fission products are investigated. 2. Study methodology 2.1. Plant model The reference plant is a two loop of Chinese 600 MW PWR with U-tube steam generators (SG). The Sketch map of nuclear island and safety system of 600 MW PWR is shown in Fig. 1. The key operating parameters of 600 MW PWR are illustrated in Table 1. The integrated safety analysis code is used to simulate PWR system response to severe accidents. RCS is modeled with a reactor vessel, active core region and two reactor coolant loops—the broken and unbroken loops, as illustrated in Fig. 2. Either loop models a single hot leg, SG, intermediate leg, reactor coolant pump, and cold leg. Additionally, the pressurizer is located on the broken loop hot leg. There are 14 nodes in primary circuit simulation, pressurizer is simulated with independent module. As shown in Fig. 3, the core region is modeled with seven radial fuel channels, plus one bypass region, thirteen axial active fuel nodes and two nonfuel nodes below the active fuel’s bottom and one non-fuel node above the active fuel’s top. Secondary circuit is modeled with main feedwater (MFW) system, auxiliary feedwater (AFW) system, main steam system and safety valves of SG. The safety injection system contains high-pressure (HPI), low-pressure injection (LPI) and two accumulators. The depressurization system includes three lines of SEBIM safety valves, relief pipe, quench tank and containment volume. Table 1 Key parameters of 600 MW PWR. Parameters Thermal power (MW) RCS average temperature (K) Pressurizer pressure (MPa) RCS mass flux (kg s−1 ) SG pressure (MPa) Mass of UO2 in core region (kg) Mass of clad in core region (kg)
Value 1930 583.15 15.5 4997.67 6.86 63,133 15,810
Fig. 3. Model of core region.
2.2. Fission products model The fission products are divided into 12 groups, as shown in Table 2, including noble gas in group 1; volatile fission products in groups 2, 3, 6, and 11; and nonvolatile fission products in other groups. The fission products in each group have common physical and chemical characteristics. Mass of the initial fission products are calculated with the hypothesis that the fuel elements are of three Table 2 Fission products groups and initial mass. Group
Fission products
1 2 3 4 5 6 7 8 9 10 11 12
Xe, Kr (noble gas) CsI, RbI TeO2 SrO MoO2 CsOH, RbOH BaO La2 O3 , Pr2 O3 , Nd2 O3 , Y2 O3 , Sm2 O3 CeO2 Sb Te2 UO2 , NpO2 , PuO2
Initial mass (kg) 239.2 20.2 0 50.2 183.7 145.9 66.6 355.0 153.4 1.4 19.9 63,643
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categories for 1, 2 and 3 years of use, respectively. The number of moles of CsI and RbI is equal to the number of moles of I, because all I is assumed to be in this form. For CsOH and RbOH, the number of moles available is calculated from the initial number of moles of Cs and Rb by subtracting the moles of I. There are similar physical and chemical characteristics for all volatile fission products groups. Therefore, CsI (group 2) is chosen to analyze in-vessel release of volatile fission products. Also, SrO (group 4) is chosen to analyze in-vessel release of nonvolatile fission products. The CORSOR-M model (Kuhlman et al., 1985) is used for invessel release of noble gas and volatile fission products. The CORSOR-O (Kuhlman et al., 1985) model is used for in-vessel release of nonvolatile fission products. Computation models are given as follows: fm = km e−q/RT fo = ko e
−q/RT
(1)
(fH2 fcoro1 + (1 − fH2 )fcoro1 )
(2)
where fm , fo is the release rate (1/min), km is the multiplier for the CORSOR-M fission product release rates, ko is the multiplier for the CORSOR-O fission product release rates, q is activation energy (kJ/mol), R is constant (0.01987), T is core node temperature (K), fH2 is molar fraction of H2 in gas flow by core node; fcoro1 is the CORSORO relative multiplier for the initial fuel (steam-rich) condition; fcoro2 is the CORSOR-O relative multiplier for the reduced fuel (H2 -rich) condition. 2.3. Accident sequences selection Because numerous severe accident sequences could induce in-vessel release, it is unnecessary to calculate all the postulated conditions, and typical accident sequences should be chosen. According to the sequence selection criteria in 10 CFR 50.54(f) of U.S. NRC (USNRC, 1988), which contribute both 1 × 10−6 /a or more to core damage and 5% or more to the total core damage frequency (CDF) are selected. Another consideration about sequences selection is the experience of NUREG-1150 (USNRC, 1990), which analyzes severe accident source term of five U.S. nuclear power plants. Finally, representative accident sequences are chosen, which are the large break loss-of-coolant (LBLOCA), the small break loss-of coolant accident (SBLOCA), the loss of feedwater accident (LOFW) and the station blackout accident (SBO). The assumptions of these sequences are failure of HPI, LPI, motor-driven AFW and turbine-driven AFW.
Fig. 4. Maximum temperature of core in cases without mitigation.
RCS pressure rise up rapidly, and the power-operated relief valves (PORV) at the top of pressurizer begins to open and close automatically for a long time. Large amount of coolant loses through the PORV, and the water level in reactor pressurizer vessel (RPV) begins to drop. Then the core is uncovered, and maximum temperature of core is risen quickly (Fig. 4). When core temperature is more than 1000 K, gap release is occurred (Fig. 5). About 5% of noble gas releases from gap. When core temperature is more than 2499 K, core begins to melt. Then in-vessel release of fission products increases quickly (Figs. 5–7). Core relocation is occurred at 15,716 s and RPV fails at 17,369 s. After all of the corium drop into cavity, in-vessel release is ceased. In the end, in-vessel release fraction of noble gas is 1.0, in-vessel release fraction of CsI and SrO are 0.993 and 1.02E−2, respectively. For the case of LOFW, reactor cannot scram immediately. Safety systems including HPI, LPI and AFW are all unavailable. Since reactor scrams later than SBO, then heat generated in the core is more than SBO. As a result, core melt progression is more quickly than SBO, RPV fails at 11,704 s. In the end, in-vessel release fraction of noble gas is 1.0, in-vessel release fraction of CsI and SrO are 0.999 and 9.39E−3, respectively. For the case of SBLOCA, break with equivalent diameter of 25 mm in cold leg is assumed. Safety systems including HPI, LPI and AFW are all unavailable. Coolant ejects into containment through
3. In-vessel release of fission products without mitigation Table 3 indicates key events for different accidents without mitigation (reference cases). For the case of SBO, reactor scrams after SBO is initiated. Safety systems including the HPI, LPI and AFW are all unavailable. The decay heat generated in the core transfers to SG. Due to loss of AFW, the secondary sides of SG become empty quickly. After loss of secondary heat sink, the core temperature and
Table 3 Key events for different accidents without mitigation. SBO Reactor scrams (s) PORVs setpoint pressure (s) Rupture disk fails (s) Core uncovers (s) Core begins to melt (s) Core relocation (s) RPV fails (s)
0 6215 6864 7875 12,412 15,716 17,369
LOFW 38 1771 1796 3452 6448 10,502 11,704
SBLOCA
LBLOCA
278 4592 4594 5044 9603 12,267 14,737
0.2 – – 7.2 1040 3195 7850
Fig. 5. In-vessel release fraction of noble gas in cases without mitigation.
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vessel release fraction of noble gas is 1.0, in-vessel release fraction of CsI and SrO are 0.996 and 1.59E−3, respectively. For the case of LBLOCA, cold leg is assumed to break at both ends, HPI, LPI and AFW are also assumed to be unavailable. Different with other cases, RCS pressure and water level in RPV drop rapidly in LBLOCA due to loss more coolant. Core starts to melt at 1040 s and relocates at 3195 s. At last, RPV fails at 7850 s. In the end, In-vessel release fraction of noble gas is 1.0, in-vessel release fraction of CsI and SrO are 0.997 and 1.90E−2, respectively. 4. In-vessel release of fission products with mitigation 4.1. Feed-and-bleed in primary circuit
Fig. 6. In-vessel release fraction of CsI in cases without mitigation.
Fig. 7. In-vessel release fraction of SrO in cases without mitigation.
break, which makes RCS pressure drop, so reactor scrams at 278 s due to low RCS pressure. As a result of unavailability of AFW, the secondary sides of SG become empty quickly. Then core temperature and RCS pressure rise up rapidly. Rupture disk fails at 4594 s and core uncovers at 5044 s. At last, RPV fails at 14,737 s. Finally, in-
When assumed condition is achieved, the ability of HPI to RCS is recovered. For SBO, LOFW and SBLOCA accidents, the pressure in RCS keeps high. So, PORV is assumed to be intentional opened, which ensure feed-and-bleed in primary circuit successful. The bleed procedure is dependent on PORV and/or break of broken loop. Three kinds of different time for availability of feed-and-bleed in primary circuit are chosen. The assumed conditions of recovering ability of HPI to RCS are that core exit temperature exceeds 923.15 K, core exit temperature exceeds 1800 K and water level in RPV is under 0.5 m. As a result of feed-and-bleed, the water level in RPV is recovered rapidly, and the maximum core temperature descends gradually. Then the progression of core melt is ceased. Tables 4–6 show key events and in-vessel release fraction of fission products while release is stopped. Comparing with reference case (without mitigation), case of core exit temperature exceeds 923.15 K has the earliest injection time. Thus, water level in RPV is recovered quickly. Very small clad oxidation fraction is produced. After core melt progression is ceased and in-vessel release of fission products is stopped, final in-vessel release fraction of fission products is reduced greatly. For the case of core exit temperature exceeds 1800 K, injection time is later than case of core exit temperature exceeds 923.15 K, so clad oxidation fraction is increased. Especially, SBLOCA and LBLOCA have bigger clad oxidation fraction than reference cases due to more coolants interact with clad, but in-vessel release of fission products is still mitigated. For the case of water level in RPV is under 0.5 m, which has the latest injection time, so clad oxidation fraction is bigger than reference case. SBLOCA and LBLOCA have bigger in-vessel release fraction of SrO than reference case owing to SrO is easy to release after core relocation, but the majority release in accidents is noble gas and volatile fission products, so in-vessel release of fission products is still mitigated.
Table 4 Key events and in-vessel release fraction with availability of primary feed-and-bleed when core exit temperature exceeds 923.15 K.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO
SBO
LOFW
SBLOCA
LBLOCA
9511 10,061 2.38E−3 3.92E−2 2.58E−8 1.03E−9
4744 5330 4.91E−3 5.00E−2 3.36E−7 9.47E−9
6375 6849 1.85E−3 3.92E−2 1.43E−8 6.00E−10
700 1217 1.08E−2 5.01E−2 3.98E−5 5.78E−7
Table 5 Key events and in-vessel release fraction with availability of primary feed-and-bleed when core exit temperature exceeds 1800 K.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO
SBO
LOFW
SBLOCA
LBLOCA
12,466 13,165 0.27 0.57 0.54 1.91E−3
6729 7365 0.35 0.65 0.63 2.10E−3
9583 10,214 0.60 0.48 0.44 1.13E−3
1368 1983 0.36 0.74 0.71 1.54E−2
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Table 6 Key events and in-vessel release fraction with availability of primary feed-and-bleed when water level in RPV is under 0.5 m.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO
SBO
LOFW
SBLOCA
LBLOCA
16,631 17,539 0.459 0.99 0.968 7.04E−3
10,774 11,577 0.456 1.0 0.967 7.89E−3
13,071 13,793 0.599 1.0 0.993 9.39E−3
4735 5438 0.304 1.0 0.995 8.54E−2
In order to evaluate the relationship between in-vessel release fraction and injection time, more start time of primary feed-andbleed are chosen for SBO. Start time of primary feed-and-bleed are about from 9000 to 17,000 s, which can be divided into three phases. The first phase indicates that start time of feed-and-bleed is before core melt, the second phase is during core melt and core relocation, the third phase is after core relocation. Fig. 8 shows clad oxidation fraction and in-vessel release fraction of fission products with different start time of primary feed-and-bleed in case of SBO. The final clad oxidation fraction without feed-and-bleed is 0.436. If implementing feed-and-bleed before core melt, as a result of quick cooling of reactor, final clad oxidation fraction is less than 0.436. If implementing feed-and-bleed after core melt, clad oxidation fraction is more than 0.436 due to more coolant interact with clad. In-vessel release fraction of noble gas and CsI increases with start time of primary feed-and-bleed, the release trend of noble gas and CsI is similar except for gap release of noble gas. The final release fraction of noble gas and CsI in reference case are 1.0 and 0.993, respectively. Quick release of noble gas and CsI is near the time of core melt, which is about from 11,500 to 13,000 s. With the influence of feed-and-bleed, core melt progression is ceased. Invessel release of noble gas and CsI is stopped, final in-vessel release fraction is less than reference case. After quick release, in-vessel release trend is slow, which increases gradually. The final release fraction of SrO without mitigation is 1.02E−2. Before core relocation, in-vessel release fraction of SrO increases with start time of feed-and-bleed. But after core relocation, there is a fall for invessel release fraction. The majority release in accident is noble gas and volatile fission products, so in-vessel release fraction of fission products increases with start time of feed-and-bleed. The relationship between in-vessel release fraction and injection time in cases of LOFW, SBLOCA and LBLOCA are similar with SBO, as shown in Fig. 9. In order to evaluate the coupling relationship between thermohydraulics and in-vessel release of fission products, case of SBO
that activates primary feed-and-bleed when core exit temperature exceeds 1800 K is analyzed in detail. SBO is assumed to happen at 0 s, reactor scrams at once. At the beginning, safety systems including HPI, LPI and AFW system are all unavailable. The decay heat generated in the core transfers to the SG. Due to loss of AFW, the secondary sides of SG become empty quickly. Subsequently, the core temperature and RCS pressure begin to rise up (Fig. 10), while the water level in RPV drops quickly. When core exit temperature exceeds 1800 K at 12,205 s, the power supply is recovered and PORV is intentional opened. HPI is activated at 12,466 s (Fig. 11), but the core has begun to melt at 12,441 s. Water level in RPV is recovered quickly. Additionally, as decreasing pressure of RCS, accumulator is
Fig. 8. Clad oxidation fraction and in-vessel release fraction with different start time of primary feed-and-bleed in case of SBO.
Fig. 10. Maximum temperature of core and clad oxidation fraction in case of SBO with primary feed-and-bleed.
Fig. 9. In-vessel release fraction with different start time of primary feed-and-bleed in cases of LOFW, SBLOCA and LBLOCA.
G.F. Huang et al. / Nuclear Engineering and Design 240 (2010) 3888–3897
Fig. 11. Injection flux and water level in RPV in case of SBO with primary feed-andbleed.
also activated and maximum temperature of core decreases gradually. In the end, core melt progression is ceased. As illustrated in Fig. 12. For noble gas, when core temperature exceeds 1000 K, gap release is occurred at 9495 s. About 5% of noble gas releases from gap. Before reflood, maximum temperature of core increases rapidly, in-vessel release fraction of noble gas is 0.337. Subsequently, since clad and coolant interact during reflood phase. Clad oxidation fraction is risen quickly (Fig. 10), which induces that in-vessel release fraction of noble gas increases rapidly. After reflood phase stops, maximum temperature of core decreases to about 2200 K, the release of noble gas is very little. After 13,811 s, there is almost none noble gas release from core. In the end, water level in RPV is maintained and core melt progression is ceased. In-vessel release fraction of noble gas is 0.567. For CsI, only a little amount exists in gap, so the fraction of gap release is less than noble gas. Since reflood water lead to clad oxidation severe, there is also a quick release of CsI during reflood phase. Subsequently, the trend of in-vessel release is similar to noble gas. After core cooling is recovered, in-vessel release fraction of CsI is 0.543. For SrO, there is almost none gap release. In-vessel release increases with core temperature. When entry reflood phase, in-
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Fig. 12. In-vessel release fraction of fission products in case of SBO with primary feed-and-bleed.
vessel release rises quickly, but comparing with noble gas and CsI, the magnitude of release fraction is very small, which is decided by nonvolatile characteristic. In the end, in-vessel release fraction of SrO is only 1.91E−3. According to above analysis, it can be concluded that feed-andbleed in primary circuit has mitigation effect on in-vessel release, and earlier start time of feed-and-bleed has better mitigation effect. 4.2. Feed-and-bleed in secondary circuit When assumed condition is achieved, AFW is recovered and SG safety valves are intentional opened, which makes feed-and-bleed in secondary circuit successful. The bleed procedure is dependent on SG safety valves. The same to feed-and-bleed in primary circuit, three kinds of different time for availability of feed-and-bleed in secondary circuit are chosen. The assumed condition of recovering AFW are that core exit temperature exceeds 923.15 K, core exit temperature exceeds 1800 K and water level in RPV is under 0.5 m. Tables 7–9 show key events and final in-vessel release fraction of fission products. For SBO, LOFW and SBLOCA, as a result of feed-and-bleed in secondary circuit, the phenomenon of SG reflux-condensation
Table 7 Key events and in-vessel release fraction with availability of secondary feed-and-bleed when core exit temperature exceeds 923.15 K.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO a
SBO
LOFW
SBLOCA
LBLOCAa
9184 9292 4.94E−4 0 0 0
4378 4487 4.35E−4 0 0 0
6150 6328 4.50E−4 0 0 0
700 – 0.328 1.0 0.998 6.68E−2
RPV fails at 10,514 s.
Table 8 Key events and in-vessel release fraction with availability of secondary feed-and-bleed when core exit temperature exceeds 1800 K.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO a
RPV fails at 8635 s.
SBO
LOFW
SBLOCA
LBLOCAa
12,205 13,625 0.245 0.498 0.471 3.73E−3
6429 8325 0.342 0.563 0.539 5.22E−3
9583 11,149 0.640 0.958 0.942 1.35E−3
1368 – 0.296 1.0 0.997 1.94E−2
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Table 9 Key events and in-vessel release fraction with availability of secondary feed-and-bleed when water level in RPV is under 0.5 m.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO a
SBO
LOFW
SBLOCA
LBLOCAa
16,474 17,479 0.445 0.992 0.969 8.14E−3
10,611 12,716 0.456 0.999 0.972 7.25E−2
13,071 13,583 0.600 1.0 0.993 8.87E−2
4735 – 0.302 1.0 0.998 6.47E−2
RPV fails at 7799 s.
appears, water level in RPV is recovered and maximum temperature of core descends gradually. Subsequently, RCS pressure decreases gradually. Accumulator is activated, so water level in RPV is recovered more quickly. Then accident progression is ceased and fission product release is stopped. In case of core exit temperature exceeds 923.15 K, injection time of secondary is the earliest, and fission product release is not occurred. In cases of core exit temperature exceeds 1800 K and water level in RPV is under 0.5 m have later injection time, clad oxidation have occurred. With the influence of feed-and-bleed in secondary circuit, in-vessel release of fission products is mitigated. For LBLOCA, owing to quick depressurization of RCS, accumulator is activated and depleted before startup of secondary feed-and-bleed. This leads to that large amount of coolant releases into containment and core uncovers. Water level in RPV is not recovered by the effect of SG reflux-condensation. Core melt and relocation are occurred. Finally, RPV fails due to creep rupture. Invessel release of fission products is not mitigated. Similar to feed-and-bleed in primary circuit, more start time of feed-and-bleed in secondary circuit are chosen for SBO, which is used to evaluate the relationship between in-vessel release fraction and injection time. The evaluated result is shown in Fig. 13. Quick release of fission products is near the time of core melt, which is about from 11,500 to 13,000 s. After activation of feed-and-bleed, water level in RPV is recovered by SG reflux-condensation and accumulator injection. Maximum temperature of core decreases, then accident progression is ceased and in-vessel release of fission products is stopped. The in-vessel release fraction is less than reference case. Final in-vessel release fraction of fission products increases with start time of secondary feed-and-bleed. The relationship between in-vessel release fraction and injection time in cases of LOFW and SBLOCA are similar with SBO, as shown in Fig. 14.
In order to evaluate the coupling relationship between thermohydraulics and in-vessel release of fission products, case of SBO that activates secondary feed-and-bleed when core exit temperature exceeds 1800 K is analyzed in detail. When core exit temperature exceeds 1800 K at 12,205 s, the power supply is recovered and secondary injection is activated, which makes successful of SG reflux-condensation (Fig. 15). Subsequently, water level in RPV increases and RCS pressure decreases. The accumulator is activated while RCS pressure is less than initial accumulator pressure. Water level in RPV is recovered at 13,625 s and maximum temperature of
Fig. 13. In-vessel release fraction of fission products with different start time of secondary feed-bleed.
Fig. 15. Condensate flow rate on SGs and flux of accumulator injection in case of SBO with secondary feed-and-bleed.
Fig. 14. In-vessel release fraction of fission products with different start time of secondary feed-and-bleed in cases of LOFW and SBLOCA.
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Fig. 16. Maximum temperature of core and water level in RPV in case of SBO with secondary feed-and-bleed. Fig. 18. Scheme of cavity structure.
core decreases gradually (Fig. 16). Finally, core melt progression is ceased and fission products release is stopped. As shown in Fig. 17, for noble gas, gap release happens while core temperature exceeds 1000 K, about 5% of noble gas releases. During reflood phase, coolant which is from reflux-condensation and accumulator is interacted with clad, great clad oxidation fraction is produced. At the same time, in-vessel release fraction of noble gas increases rapidly. After water level in RPV is recovered, increase of in-vessel release fraction is very small. After 15,075 s, there is almost none noble gas release. At last, in-vessel release fraction of noble gas is 0.498. For CsI, in-vessel release fraction of CsI is very small during gap release. There is also a quick release during reflood phase owing to interaction between clad and coolant. Final in-vessel release fraction of CsI is 0.471. For SrO, there is almost none gap release. The release trend is similar to CsI, but in-vessel release fraction is less than CsI owing to nonvolatile characteristic. Final in-vessel release fraction of SrO is only 3.73E−3. Based on above evaluation, it can be concluded that feed-andbleed in secondary circuit has mitigation effect on in-vessel release
Fig. 17. In-vessel release fraction of fission products in case of SBO with secondary feed-and-bleed.
of fission products for most accidents that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Earlier start time of secondary feed-and-bleed has better mitigation effect. But due to quick loss of coolant for LBLOCA, water level in RPV is not recovered and accident progression is not ceased, in-vessel release is not mitigated. 4.3. Ex-vessel cooling Ex-vessel cooling is a useful countermeasure to make successful of in-vessel retention (NEA, 1998). According to previous study on Chinese 600 MW PWR, cavity flooding is a feasible countermeasure to keep RPV lower head integrated. In this paper, the mitigation effect of ex-vessel cooling on in-vessel release of fission products is analyzed. As shown in Fig. 18, RPV is located in the cavity. The height of cavity is 14.8 m. There is a cavity vent at the height of 11.4 m. The distance between bottom of RPV and cavity floor is 4.48 m. When cavity flooding is implemented, water level in cavity increases gradually and reaches bottom of RPV. The water goes into insulation and absorbs heat from RPV wall, then some water transforms into steam and leaves insulation through outlet. After water level in cavity reaches 11.4 m, water will overflow through cavity vent. As a result, core residual heat transfers to cavity water via RPV wall, which keeps integrity of RPV. In this paper, the entry condition of cavity flooding is core exit temperature exceeds 923.15 K. In SBO, LOFW and SBLOCA, one PORV is intentional opened for the sake of avoiding creep rupture of RPV. Table 10 shows key events and final in-vessel release fraction of fission products. With the influence of cavity flooding, water level in cavity of each case reaches 11.4 m, but water level in RPV is not recovered. Core temperature increases, then core melt and relocation are occurred. Clad oxidation fraction is very big. In-vessel release fraction of noble gas and CsI are basically equal to reference case, and in-vessel release fraction of SrO is bigger than reference case. Table 11 shows comparison results of fraction of release into containment between ex-vessel cooling cases and reference cases. For SrO, fraction of release into containment is largely less than reference case. This is due to ex-vessel release fraction of nonvolatile fission products is more than noble gas and volatile fission products (Huang et al., 2009). In ex-vessel cooling cases of SBO, SBLOCA and LOFW, RCS pressure is low during core melts, so less volatile fission products deposit in RCS comparing with reference cases. This leads to that more volatile fission products release into
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Table 10 Key events and in-vessel release fraction with availability of cavity flooding when core exit temperature exceeds 923.15 K.
Start time of cavity flooding (s) Water level in cavity reaches 11.4 m (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO
SBO
LOFW
SBLOCA
LBLOCA
9184 11,111 – 0.529 1.0 0.994 1.67E−2
4378 6519 – 0.618 1.0 0.995 1.88E−2
6150 8187 – 0.590 1.0 0.984 5.23E−2
700 2555 – 0.301 1.0 0.997 1.70E−2
Table 11 Comparison of fraction of release into containment between ex-vessel cooling cases and reference casesa .
Noble gas (ex-vessel cooling) Noble gas (reference cases) CsI (ex-vessel cooling) CsI (reference cases) SrO (ex-vessel cooling) SrO (reference cases) a
SBO
LOFW
SBLOCA
LBLOCA
0.996 0.995 0.425 0.198 9.72E−3 8.98E−2
0.990 0.996 0.347 0.102 6.86E−3 6.23E−2
0.997 0.997 0.397 0.183 2.07E−2 8.30E−2
0.997 0.996 0.635 0.642 9.98E−3 3.10E−2
Ex-vessel release is computed in reference cases, duration of ex-vessel release is 10 h.
containment. Nevertheless, success of IVR avoids direct containment heating (DCH) in cases of core melts at high pressure and largely reduces probability of containment failure. Finally, fraction of release into environment can be largely reduced. In order to evaluate the coupling relationship between thermohydraulics and in-vessel release of fission products, SBO is chosen to be analyzed in detail. Before core exit temperature exceeds 923.15 K, accident progression is the same to reference case. When core exit temperature exceeds 923.15 K, cavity flooding system is activated and one PORV is intentional opened at 9197 s. Water level in cavity reaches 11.4 m at 11,111 s. Subsequently, flux of cavity flooding is kept and excessive water is drained through cavity vent (Fig. 19). During this phase, water level in RPV drops rapidly and core uncovers. Accumulator is activated at 10,502 s, so water level in RPV increases a little. However, due to continuous evaporation of coolant and accumulator is depleted, water level in RPV still drops and core temperature increases. Finally, core relocates at 29,856 s (Fig. 20). Core residual heat transfers to cavity water via RPV wall, which keeps RPV lower head integrated. As shown in Fig. 21, in-vessel release fraction of fission products is very large. For noble gas and CsI, release is mostly occurred dur-
Fig. 19. Water level in cavity and heat transfer of RPV to cavity water in case of SBO with ex-vessel cooling.
Fig. 20. Water level in RPV and maximum temperature of core in case of SBO with ex-vessel cooling.
Fig. 21. In-vessel release fraction of fission products in case of SBO with ex-vessel cooling.
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ing core melt. After core relocates, the increased in-vessel release fraction is about 0.05. In the end, in-vessel release fraction of noble gas and CsI are 1.0 and 0.997, respectively. For SrO, the release trend is flatter than noble gas and CsI. In-vessel release fraction of SrO is only 0.0167 owing to nonvolatility. According to above analysis results of ex-vessel cooling, core melt and relocation are not ceased. In-vessel release fraction of noble gas and CsI are almost equal to reference case, and in-vessel release fraction of SrO is bigger than reference case. So, ex-vessel cooling has no mitigation effect on in-vessel release. Nevertheless, success of IVR can largely reduce fraction of release into environment.
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(3) Ex-vessel cooling is used to prevent RPV failure, it is not an effective countermeasure to control in-vessel release of fission products owing to inevitable core melt and relocation. In-vessel release fraction in ex-vessel cooling case is almost equal to reference case. Nevertheless, success of IVR avoids DCH in cases of core melts at high pressure and largely reduces probability of containment failure, which can largely reduce fraction of release into environment. Acknowledgement The authors acknowledge National Basic Research Program of China (no. 2009CB724301) support.
5. Conclusions References The mitigation effect of different severe accident management countermeasures on in-vessel release of fission products is studied for Chinese 600 MW PWR. According to the analytical results, it can be concluded as follows: (1) Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products. Implementation of feed-and-bleed in primary circuit when core exit temperature exceeds 923.15 K, in-vessel release fraction of noble gas is 1E−2 orders of magnitude, in-vessel release of other fission products can be ignored. With later time to implement the countermeasure, accident progression is ceased later, so final in-vessel release fraction of fission products is bigger. (2) Feed-and-bleed in secondary circuit has mitigation effect on in-vessel release of fission products for most accidents that can cease core melt progression. For SBO, LOFW and SBLOCA, water level in RPV is recovered owing to SG reflux-condensation and accumulator injection as a result of feed-and-bleed in secondary circuit. Accident progression is ceased and in-vessel release is stopped. Implementation of feed-and-bleed in secondary circuit when core exit temperature exceeds 923.15 K, there are not fission products release. Bigger in-vessel release fraction of fission products is produced with later time of countermeasure implementation. However, in case of LBLOCA, water level in RPV is not recovered due to loss more coolant. Core melt and relocation are occurred. In-vessel release of fission products is not mitigated.
Andre, B., et al., 1996. Fission product releases at severe light water reactor accident conditions: ORNL/CEA measurements versus calculations. Nuclear Technol. 114, 23–49. Baker Jr., L., et al., 1988. Source Term Experiments Project (STEP): A Summary, NP5753M, Electric Power Research Institute. Denning, R.S., et al., 1986. Radionuclide Release Calculations for Selected Severe Accident Scenarios, NUREG/CR-4624. Dubourg, R., et al., 2005. Fission product release in the first two PHEBUS tests FPT0 and FPT1. Nuclear Eng. Des. 235, 2183–2208. Huang, G.F., et al., 2009. Source term in severe accident induced by large break loss of coolant accident for nuclear power plant. Atomic Energy Sci. Technol. 43 (7), 609–615 (in Chinese). Kudo, T., et al., 2001. Influence of pressure on cesium release from irradiated fuel at temperatures up to 2773 K. Nuclear Sci. Technol. 38 (10), 910–911. Kuhlman, et al., 1985. CORSOR User’s Manual. Battelle Columbus Laboratories, NUREG/CR-4173. Lee, M., et al., 2008. Quantification of severe accident source terms of a Westinghouse 3-loop plant. Nuclear Eng. Des. 238, 1080–1092. Leveque, J.P., et al., 1994. The fission product aerosols in the programmes heva and vercors. J. Aerosol Sci. 25, 87–88. Martinson, Z.R., et al., 1986. PBF Severe Fuel Damage Test 1-1 Test Results Report, NUREG/CR-4684. NEA, 1998. Workshop on In-Vessel Core Debris Retention and Coolability, NEA/CSNI/R(98)21. NEA, 2000. Insights into the control of the release of iodine, cesium, strontium and other fission product in the containment by severe accident management, NEA/CSNI/R(2000)9. USNRC, 1988. Individual plant examination for severe accident vulnerabilities 10 CFR 50.54 (f). Generic Letter No. 88-20. USNRC, 1990. Reactor Risk Reference Document, NUREG-1150. Wright, A.L., et al., 1994. Primary System Fission Product Release and Transport, NUREG/CR-6193.
Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Study on mitigation of in-vessel release of fission products in severe accidents of PWR G.F. Huang ∗ , L.L. Tong, J.X. Li, X.W. Cao School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200240, China
a r t i c l e
i n f o
Article history: Received 9 April 2010 Received in revised form 10 August 2010 Accepted 20 August 2010
a b s t r a c t During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, including in-vessel and ex-vessel release. Mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. Mitigation countermeasures to in-vessel release are studied for Chinese 600 MW pressurized water reactor (PWR), including feed-andbleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling. SBO, LOFW, SBLOCA and LBLOCA are selected as typical severe accident sequences. Based on the evaluation of in-vessel release with different startup time of countermeasure, and the coupling relationship between thermohydraulics and in-vessel release of fission products, some results are achieved. Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products, and earlier startup time of countermeasure is more feasible. Feed-and-bleed in secondary circuit is also an effective countermeasure to mitigate in-vessel release for most severe accident sequences that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Ex-vessel cooling has no mitigation effect on in-vessel release owing to inevitable core melt and relocation. © 2010 Elsevier B.V. All rights reserved.
1. Introduction During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, which includes in-vessel and ex-vessel release (Denning et al., 1986). In-vessel release is occurred inside reactor pressure vessel (RPV) during degradation of core, ex-vessel release is occurred outside RPV after RPV is failed. For in-vessel release, released fission products transport into containment through primary system (Wright et al., 1994). If containment has great leakage, much fission products will release into environment, which is contributed to severe accident source term (Lee et al., 2008). This may induce health risk of public. So, mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. If integrity of RPV is maintained and in-vessel release of fission products is mitigated, radiological consequence will be alleviated and public health will be protected in the early phase of severe accidents. Strategies that aim at preventing or mitigating core damage all focus on reducing the fuel temperature. To the extent that these strategies do cool the fuel or slow the temperature rise. They may
∗ Corresponding author. Tel.: +86 21 34205495; fax: +86 21 34205495. E-mail addresses: [email protected] (G.F. Huang), [email protected] (X.W. Cao). 0029-5493/$ – see front matter © 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2010.08.010
also mitigate in-vessel release of fission products that is very dependent on fuel temperature (NEA, 2000). For instance, addition of coolant has a profound effect on mitigation of in-vessel release of fission products even if additive coolant is insufficient to cool the fuel. The additive coolant may limit the amounts of radioactive material released from the fuel that passes through the reactor coolant system (RCS) and reaches the containment. Mitigation of in-vessel release of fission products during severe accidents can be achieved by implementing some severe accident countermeasures. Many in-pile and out-of-pile experiments are conducted to better understand fission product release behavior in severe accidents. In-pile experiments include: the Phebus fission product and severe fuel damage (SFD) tests (Dubourg et al., 2005), the power burst facility (PBF) SFD tests (Martinson et al., 1986), the source term experiments project (STEP) (Baker et al., 1988), etc. Out-of-pile experiments include: HEVA program (Leveque et al., 1994), VERCOR program (Andre et al., 1996), verification experiments of radionuclides gas/aerosol release (VEGA) program (Kudo et al., 2001), etc. Most of them are concentrated on fission product release behavior in the accident environment, and fission product release models are developed based on these experiments. However, the research to effect of reflood/quenching on fission products release is relatively less, especially for the actual severe accident management countermeasures. Thus, the study of effect of severe accident management countermeasures on fission products release is needed for actual nuclear power plant.
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Fig. 2. RCS model. Fig. 1. Sketch map of nuclear island and safety system of 600 MW PWR.
This paper presents analyses of in-vessel release of fission products in severe accidents for Chinese 600 MW pressurized water reactor (PWR). Mitigation effect of feed-and-bleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling on in-vessel release of fission products are investigated. 2. Study methodology 2.1. Plant model The reference plant is a two loop of Chinese 600 MW PWR with U-tube steam generators (SG). The Sketch map of nuclear island and safety system of 600 MW PWR is shown in Fig. 1. The key operating parameters of 600 MW PWR are illustrated in Table 1. The integrated safety analysis code is used to simulate PWR system response to severe accidents. RCS is modeled with a reactor vessel, active core region and two reactor coolant loops—the broken and unbroken loops, as illustrated in Fig. 2. Either loop models a single hot leg, SG, intermediate leg, reactor coolant pump, and cold leg. Additionally, the pressurizer is located on the broken loop hot leg. There are 14 nodes in primary circuit simulation, pressurizer is simulated with independent module. As shown in Fig. 3, the core region is modeled with seven radial fuel channels, plus one bypass region, thirteen axial active fuel nodes and two nonfuel nodes below the active fuel’s bottom and one non-fuel node above the active fuel’s top. Secondary circuit is modeled with main feedwater (MFW) system, auxiliary feedwater (AFW) system, main steam system and safety valves of SG. The safety injection system contains high-pressure (HPI), low-pressure injection (LPI) and two accumulators. The depressurization system includes three lines of SEBIM safety valves, relief pipe, quench tank and containment volume. Table 1 Key parameters of 600 MW PWR. Parameters Thermal power (MW) RCS average temperature (K) Pressurizer pressure (MPa) RCS mass flux (kg s−1 ) SG pressure (MPa) Mass of UO2 in core region (kg) Mass of clad in core region (kg)
Value 1930 583.15 15.5 4997.67 6.86 63,133 15,810
Fig. 3. Model of core region.
2.2. Fission products model The fission products are divided into 12 groups, as shown in Table 2, including noble gas in group 1; volatile fission products in groups 2, 3, 6, and 11; and nonvolatile fission products in other groups. The fission products in each group have common physical and chemical characteristics. Mass of the initial fission products are calculated with the hypothesis that the fuel elements are of three Table 2 Fission products groups and initial mass. Group
Fission products
1 2 3 4 5 6 7 8 9 10 11 12
Xe, Kr (noble gas) CsI, RbI TeO2 SrO MoO2 CsOH, RbOH BaO La2 O3 , Pr2 O3 , Nd2 O3 , Y2 O3 , Sm2 O3 CeO2 Sb Te2 UO2 , NpO2 , PuO2
Initial mass (kg) 239.2 20.2 0 50.2 183.7 145.9 66.6 355.0 153.4 1.4 19.9 63,643
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categories for 1, 2 and 3 years of use, respectively. The number of moles of CsI and RbI is equal to the number of moles of I, because all I is assumed to be in this form. For CsOH and RbOH, the number of moles available is calculated from the initial number of moles of Cs and Rb by subtracting the moles of I. There are similar physical and chemical characteristics for all volatile fission products groups. Therefore, CsI (group 2) is chosen to analyze in-vessel release of volatile fission products. Also, SrO (group 4) is chosen to analyze in-vessel release of nonvolatile fission products. The CORSOR-M model (Kuhlman et al., 1985) is used for invessel release of noble gas and volatile fission products. The CORSOR-O (Kuhlman et al., 1985) model is used for in-vessel release of nonvolatile fission products. Computation models are given as follows: fm = km e−q/RT fo = ko e
−q/RT
(1)
(fH2 fcoro1 + (1 − fH2 )fcoro1 )
(2)
where fm , fo is the release rate (1/min), km is the multiplier for the CORSOR-M fission product release rates, ko is the multiplier for the CORSOR-O fission product release rates, q is activation energy (kJ/mol), R is constant (0.01987), T is core node temperature (K), fH2 is molar fraction of H2 in gas flow by core node; fcoro1 is the CORSORO relative multiplier for the initial fuel (steam-rich) condition; fcoro2 is the CORSOR-O relative multiplier for the reduced fuel (H2 -rich) condition. 2.3. Accident sequences selection Because numerous severe accident sequences could induce in-vessel release, it is unnecessary to calculate all the postulated conditions, and typical accident sequences should be chosen. According to the sequence selection criteria in 10 CFR 50.54(f) of U.S. NRC (USNRC, 1988), which contribute both 1 × 10−6 /a or more to core damage and 5% or more to the total core damage frequency (CDF) are selected. Another consideration about sequences selection is the experience of NUREG-1150 (USNRC, 1990), which analyzes severe accident source term of five U.S. nuclear power plants. Finally, representative accident sequences are chosen, which are the large break loss-of-coolant (LBLOCA), the small break loss-of coolant accident (SBLOCA), the loss of feedwater accident (LOFW) and the station blackout accident (SBO). The assumptions of these sequences are failure of HPI, LPI, motor-driven AFW and turbine-driven AFW.
Fig. 4. Maximum temperature of core in cases without mitigation.
RCS pressure rise up rapidly, and the power-operated relief valves (PORV) at the top of pressurizer begins to open and close automatically for a long time. Large amount of coolant loses through the PORV, and the water level in reactor pressurizer vessel (RPV) begins to drop. Then the core is uncovered, and maximum temperature of core is risen quickly (Fig. 4). When core temperature is more than 1000 K, gap release is occurred (Fig. 5). About 5% of noble gas releases from gap. When core temperature is more than 2499 K, core begins to melt. Then in-vessel release of fission products increases quickly (Figs. 5–7). Core relocation is occurred at 15,716 s and RPV fails at 17,369 s. After all of the corium drop into cavity, in-vessel release is ceased. In the end, in-vessel release fraction of noble gas is 1.0, in-vessel release fraction of CsI and SrO are 0.993 and 1.02E−2, respectively. For the case of LOFW, reactor cannot scram immediately. Safety systems including HPI, LPI and AFW are all unavailable. Since reactor scrams later than SBO, then heat generated in the core is more than SBO. As a result, core melt progression is more quickly than SBO, RPV fails at 11,704 s. In the end, in-vessel release fraction of noble gas is 1.0, in-vessel release fraction of CsI and SrO are 0.999 and 9.39E−3, respectively. For the case of SBLOCA, break with equivalent diameter of 25 mm in cold leg is assumed. Safety systems including HPI, LPI and AFW are all unavailable. Coolant ejects into containment through
3. In-vessel release of fission products without mitigation Table 3 indicates key events for different accidents without mitigation (reference cases). For the case of SBO, reactor scrams after SBO is initiated. Safety systems including the HPI, LPI and AFW are all unavailable. The decay heat generated in the core transfers to SG. Due to loss of AFW, the secondary sides of SG become empty quickly. After loss of secondary heat sink, the core temperature and
Table 3 Key events for different accidents without mitigation. SBO Reactor scrams (s) PORVs setpoint pressure (s) Rupture disk fails (s) Core uncovers (s) Core begins to melt (s) Core relocation (s) RPV fails (s)
0 6215 6864 7875 12,412 15,716 17,369
LOFW 38 1771 1796 3452 6448 10,502 11,704
SBLOCA
LBLOCA
278 4592 4594 5044 9603 12,267 14,737
0.2 – – 7.2 1040 3195 7850
Fig. 5. In-vessel release fraction of noble gas in cases without mitigation.
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vessel release fraction of noble gas is 1.0, in-vessel release fraction of CsI and SrO are 0.996 and 1.59E−3, respectively. For the case of LBLOCA, cold leg is assumed to break at both ends, HPI, LPI and AFW are also assumed to be unavailable. Different with other cases, RCS pressure and water level in RPV drop rapidly in LBLOCA due to loss more coolant. Core starts to melt at 1040 s and relocates at 3195 s. At last, RPV fails at 7850 s. In the end, In-vessel release fraction of noble gas is 1.0, in-vessel release fraction of CsI and SrO are 0.997 and 1.90E−2, respectively. 4. In-vessel release of fission products with mitigation 4.1. Feed-and-bleed in primary circuit
Fig. 6. In-vessel release fraction of CsI in cases without mitigation.
Fig. 7. In-vessel release fraction of SrO in cases without mitigation.
break, which makes RCS pressure drop, so reactor scrams at 278 s due to low RCS pressure. As a result of unavailability of AFW, the secondary sides of SG become empty quickly. Then core temperature and RCS pressure rise up rapidly. Rupture disk fails at 4594 s and core uncovers at 5044 s. At last, RPV fails at 14,737 s. Finally, in-
When assumed condition is achieved, the ability of HPI to RCS is recovered. For SBO, LOFW and SBLOCA accidents, the pressure in RCS keeps high. So, PORV is assumed to be intentional opened, which ensure feed-and-bleed in primary circuit successful. The bleed procedure is dependent on PORV and/or break of broken loop. Three kinds of different time for availability of feed-and-bleed in primary circuit are chosen. The assumed conditions of recovering ability of HPI to RCS are that core exit temperature exceeds 923.15 K, core exit temperature exceeds 1800 K and water level in RPV is under 0.5 m. As a result of feed-and-bleed, the water level in RPV is recovered rapidly, and the maximum core temperature descends gradually. Then the progression of core melt is ceased. Tables 4–6 show key events and in-vessel release fraction of fission products while release is stopped. Comparing with reference case (without mitigation), case of core exit temperature exceeds 923.15 K has the earliest injection time. Thus, water level in RPV is recovered quickly. Very small clad oxidation fraction is produced. After core melt progression is ceased and in-vessel release of fission products is stopped, final in-vessel release fraction of fission products is reduced greatly. For the case of core exit temperature exceeds 1800 K, injection time is later than case of core exit temperature exceeds 923.15 K, so clad oxidation fraction is increased. Especially, SBLOCA and LBLOCA have bigger clad oxidation fraction than reference cases due to more coolants interact with clad, but in-vessel release of fission products is still mitigated. For the case of water level in RPV is under 0.5 m, which has the latest injection time, so clad oxidation fraction is bigger than reference case. SBLOCA and LBLOCA have bigger in-vessel release fraction of SrO than reference case owing to SrO is easy to release after core relocation, but the majority release in accidents is noble gas and volatile fission products, so in-vessel release of fission products is still mitigated.
Table 4 Key events and in-vessel release fraction with availability of primary feed-and-bleed when core exit temperature exceeds 923.15 K.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO
SBO
LOFW
SBLOCA
LBLOCA
9511 10,061 2.38E−3 3.92E−2 2.58E−8 1.03E−9
4744 5330 4.91E−3 5.00E−2 3.36E−7 9.47E−9
6375 6849 1.85E−3 3.92E−2 1.43E−8 6.00E−10
700 1217 1.08E−2 5.01E−2 3.98E−5 5.78E−7
Table 5 Key events and in-vessel release fraction with availability of primary feed-and-bleed when core exit temperature exceeds 1800 K.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO
SBO
LOFW
SBLOCA
LBLOCA
12,466 13,165 0.27 0.57 0.54 1.91E−3
6729 7365 0.35 0.65 0.63 2.10E−3
9583 10,214 0.60 0.48 0.44 1.13E−3
1368 1983 0.36 0.74 0.71 1.54E−2
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Table 6 Key events and in-vessel release fraction with availability of primary feed-and-bleed when water level in RPV is under 0.5 m.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO
SBO
LOFW
SBLOCA
LBLOCA
16,631 17,539 0.459 0.99 0.968 7.04E−3
10,774 11,577 0.456 1.0 0.967 7.89E−3
13,071 13,793 0.599 1.0 0.993 9.39E−3
4735 5438 0.304 1.0 0.995 8.54E−2
In order to evaluate the relationship between in-vessel release fraction and injection time, more start time of primary feed-andbleed are chosen for SBO. Start time of primary feed-and-bleed are about from 9000 to 17,000 s, which can be divided into three phases. The first phase indicates that start time of feed-and-bleed is before core melt, the second phase is during core melt and core relocation, the third phase is after core relocation. Fig. 8 shows clad oxidation fraction and in-vessel release fraction of fission products with different start time of primary feed-and-bleed in case of SBO. The final clad oxidation fraction without feed-and-bleed is 0.436. If implementing feed-and-bleed before core melt, as a result of quick cooling of reactor, final clad oxidation fraction is less than 0.436. If implementing feed-and-bleed after core melt, clad oxidation fraction is more than 0.436 due to more coolant interact with clad. In-vessel release fraction of noble gas and CsI increases with start time of primary feed-and-bleed, the release trend of noble gas and CsI is similar except for gap release of noble gas. The final release fraction of noble gas and CsI in reference case are 1.0 and 0.993, respectively. Quick release of noble gas and CsI is near the time of core melt, which is about from 11,500 to 13,000 s. With the influence of feed-and-bleed, core melt progression is ceased. Invessel release of noble gas and CsI is stopped, final in-vessel release fraction is less than reference case. After quick release, in-vessel release trend is slow, which increases gradually. The final release fraction of SrO without mitigation is 1.02E−2. Before core relocation, in-vessel release fraction of SrO increases with start time of feed-and-bleed. But after core relocation, there is a fall for invessel release fraction. The majority release in accident is noble gas and volatile fission products, so in-vessel release fraction of fission products increases with start time of feed-and-bleed. The relationship between in-vessel release fraction and injection time in cases of LOFW, SBLOCA and LBLOCA are similar with SBO, as shown in Fig. 9. In order to evaluate the coupling relationship between thermohydraulics and in-vessel release of fission products, case of SBO
that activates primary feed-and-bleed when core exit temperature exceeds 1800 K is analyzed in detail. SBO is assumed to happen at 0 s, reactor scrams at once. At the beginning, safety systems including HPI, LPI and AFW system are all unavailable. The decay heat generated in the core transfers to the SG. Due to loss of AFW, the secondary sides of SG become empty quickly. Subsequently, the core temperature and RCS pressure begin to rise up (Fig. 10), while the water level in RPV drops quickly. When core exit temperature exceeds 1800 K at 12,205 s, the power supply is recovered and PORV is intentional opened. HPI is activated at 12,466 s (Fig. 11), but the core has begun to melt at 12,441 s. Water level in RPV is recovered quickly. Additionally, as decreasing pressure of RCS, accumulator is
Fig. 8. Clad oxidation fraction and in-vessel release fraction with different start time of primary feed-and-bleed in case of SBO.
Fig. 10. Maximum temperature of core and clad oxidation fraction in case of SBO with primary feed-and-bleed.
Fig. 9. In-vessel release fraction with different start time of primary feed-and-bleed in cases of LOFW, SBLOCA and LBLOCA.
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Fig. 11. Injection flux and water level in RPV in case of SBO with primary feed-andbleed.
also activated and maximum temperature of core decreases gradually. In the end, core melt progression is ceased. As illustrated in Fig. 12. For noble gas, when core temperature exceeds 1000 K, gap release is occurred at 9495 s. About 5% of noble gas releases from gap. Before reflood, maximum temperature of core increases rapidly, in-vessel release fraction of noble gas is 0.337. Subsequently, since clad and coolant interact during reflood phase. Clad oxidation fraction is risen quickly (Fig. 10), which induces that in-vessel release fraction of noble gas increases rapidly. After reflood phase stops, maximum temperature of core decreases to about 2200 K, the release of noble gas is very little. After 13,811 s, there is almost none noble gas release from core. In the end, water level in RPV is maintained and core melt progression is ceased. In-vessel release fraction of noble gas is 0.567. For CsI, only a little amount exists in gap, so the fraction of gap release is less than noble gas. Since reflood water lead to clad oxidation severe, there is also a quick release of CsI during reflood phase. Subsequently, the trend of in-vessel release is similar to noble gas. After core cooling is recovered, in-vessel release fraction of CsI is 0.543. For SrO, there is almost none gap release. In-vessel release increases with core temperature. When entry reflood phase, in-
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Fig. 12. In-vessel release fraction of fission products in case of SBO with primary feed-and-bleed.
vessel release rises quickly, but comparing with noble gas and CsI, the magnitude of release fraction is very small, which is decided by nonvolatile characteristic. In the end, in-vessel release fraction of SrO is only 1.91E−3. According to above analysis, it can be concluded that feed-andbleed in primary circuit has mitigation effect on in-vessel release, and earlier start time of feed-and-bleed has better mitigation effect. 4.2. Feed-and-bleed in secondary circuit When assumed condition is achieved, AFW is recovered and SG safety valves are intentional opened, which makes feed-and-bleed in secondary circuit successful. The bleed procedure is dependent on SG safety valves. The same to feed-and-bleed in primary circuit, three kinds of different time for availability of feed-and-bleed in secondary circuit are chosen. The assumed condition of recovering AFW are that core exit temperature exceeds 923.15 K, core exit temperature exceeds 1800 K and water level in RPV is under 0.5 m. Tables 7–9 show key events and final in-vessel release fraction of fission products. For SBO, LOFW and SBLOCA, as a result of feed-and-bleed in secondary circuit, the phenomenon of SG reflux-condensation
Table 7 Key events and in-vessel release fraction with availability of secondary feed-and-bleed when core exit temperature exceeds 923.15 K.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO a
SBO
LOFW
SBLOCA
LBLOCAa
9184 9292 4.94E−4 0 0 0
4378 4487 4.35E−4 0 0 0
6150 6328 4.50E−4 0 0 0
700 – 0.328 1.0 0.998 6.68E−2
RPV fails at 10,514 s.
Table 8 Key events and in-vessel release fraction with availability of secondary feed-and-bleed when core exit temperature exceeds 1800 K.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO a
RPV fails at 8635 s.
SBO
LOFW
SBLOCA
LBLOCAa
12,205 13,625 0.245 0.498 0.471 3.73E−3
6429 8325 0.342 0.563 0.539 5.22E−3
9583 11,149 0.640 0.958 0.942 1.35E−3
1368 – 0.296 1.0 0.997 1.94E−2
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Table 9 Key events and in-vessel release fraction with availability of secondary feed-and-bleed when water level in RPV is under 0.5 m.
Start time of feed-and-bleed (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO a
SBO
LOFW
SBLOCA
LBLOCAa
16,474 17,479 0.445 0.992 0.969 8.14E−3
10,611 12,716 0.456 0.999 0.972 7.25E−2
13,071 13,583 0.600 1.0 0.993 8.87E−2
4735 – 0.302 1.0 0.998 6.47E−2
RPV fails at 7799 s.
appears, water level in RPV is recovered and maximum temperature of core descends gradually. Subsequently, RCS pressure decreases gradually. Accumulator is activated, so water level in RPV is recovered more quickly. Then accident progression is ceased and fission product release is stopped. In case of core exit temperature exceeds 923.15 K, injection time of secondary is the earliest, and fission product release is not occurred. In cases of core exit temperature exceeds 1800 K and water level in RPV is under 0.5 m have later injection time, clad oxidation have occurred. With the influence of feed-and-bleed in secondary circuit, in-vessel release of fission products is mitigated. For LBLOCA, owing to quick depressurization of RCS, accumulator is activated and depleted before startup of secondary feed-and-bleed. This leads to that large amount of coolant releases into containment and core uncovers. Water level in RPV is not recovered by the effect of SG reflux-condensation. Core melt and relocation are occurred. Finally, RPV fails due to creep rupture. Invessel release of fission products is not mitigated. Similar to feed-and-bleed in primary circuit, more start time of feed-and-bleed in secondary circuit are chosen for SBO, which is used to evaluate the relationship between in-vessel release fraction and injection time. The evaluated result is shown in Fig. 13. Quick release of fission products is near the time of core melt, which is about from 11,500 to 13,000 s. After activation of feed-and-bleed, water level in RPV is recovered by SG reflux-condensation and accumulator injection. Maximum temperature of core decreases, then accident progression is ceased and in-vessel release of fission products is stopped. The in-vessel release fraction is less than reference case. Final in-vessel release fraction of fission products increases with start time of secondary feed-and-bleed. The relationship between in-vessel release fraction and injection time in cases of LOFW and SBLOCA are similar with SBO, as shown in Fig. 14.
In order to evaluate the coupling relationship between thermohydraulics and in-vessel release of fission products, case of SBO that activates secondary feed-and-bleed when core exit temperature exceeds 1800 K is analyzed in detail. When core exit temperature exceeds 1800 K at 12,205 s, the power supply is recovered and secondary injection is activated, which makes successful of SG reflux-condensation (Fig. 15). Subsequently, water level in RPV increases and RCS pressure decreases. The accumulator is activated while RCS pressure is less than initial accumulator pressure. Water level in RPV is recovered at 13,625 s and maximum temperature of
Fig. 13. In-vessel release fraction of fission products with different start time of secondary feed-bleed.
Fig. 15. Condensate flow rate on SGs and flux of accumulator injection in case of SBO with secondary feed-and-bleed.
Fig. 14. In-vessel release fraction of fission products with different start time of secondary feed-and-bleed in cases of LOFW and SBLOCA.
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Fig. 16. Maximum temperature of core and water level in RPV in case of SBO with secondary feed-and-bleed. Fig. 18. Scheme of cavity structure.
core decreases gradually (Fig. 16). Finally, core melt progression is ceased and fission products release is stopped. As shown in Fig. 17, for noble gas, gap release happens while core temperature exceeds 1000 K, about 5% of noble gas releases. During reflood phase, coolant which is from reflux-condensation and accumulator is interacted with clad, great clad oxidation fraction is produced. At the same time, in-vessel release fraction of noble gas increases rapidly. After water level in RPV is recovered, increase of in-vessel release fraction is very small. After 15,075 s, there is almost none noble gas release. At last, in-vessel release fraction of noble gas is 0.498. For CsI, in-vessel release fraction of CsI is very small during gap release. There is also a quick release during reflood phase owing to interaction between clad and coolant. Final in-vessel release fraction of CsI is 0.471. For SrO, there is almost none gap release. The release trend is similar to CsI, but in-vessel release fraction is less than CsI owing to nonvolatile characteristic. Final in-vessel release fraction of SrO is only 3.73E−3. Based on above evaluation, it can be concluded that feed-andbleed in secondary circuit has mitigation effect on in-vessel release
Fig. 17. In-vessel release fraction of fission products in case of SBO with secondary feed-and-bleed.
of fission products for most accidents that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Earlier start time of secondary feed-and-bleed has better mitigation effect. But due to quick loss of coolant for LBLOCA, water level in RPV is not recovered and accident progression is not ceased, in-vessel release is not mitigated. 4.3. Ex-vessel cooling Ex-vessel cooling is a useful countermeasure to make successful of in-vessel retention (NEA, 1998). According to previous study on Chinese 600 MW PWR, cavity flooding is a feasible countermeasure to keep RPV lower head integrated. In this paper, the mitigation effect of ex-vessel cooling on in-vessel release of fission products is analyzed. As shown in Fig. 18, RPV is located in the cavity. The height of cavity is 14.8 m. There is a cavity vent at the height of 11.4 m. The distance between bottom of RPV and cavity floor is 4.48 m. When cavity flooding is implemented, water level in cavity increases gradually and reaches bottom of RPV. The water goes into insulation and absorbs heat from RPV wall, then some water transforms into steam and leaves insulation through outlet. After water level in cavity reaches 11.4 m, water will overflow through cavity vent. As a result, core residual heat transfers to cavity water via RPV wall, which keeps integrity of RPV. In this paper, the entry condition of cavity flooding is core exit temperature exceeds 923.15 K. In SBO, LOFW and SBLOCA, one PORV is intentional opened for the sake of avoiding creep rupture of RPV. Table 10 shows key events and final in-vessel release fraction of fission products. With the influence of cavity flooding, water level in cavity of each case reaches 11.4 m, but water level in RPV is not recovered. Core temperature increases, then core melt and relocation are occurred. Clad oxidation fraction is very big. In-vessel release fraction of noble gas and CsI are basically equal to reference case, and in-vessel release fraction of SrO is bigger than reference case. Table 11 shows comparison results of fraction of release into containment between ex-vessel cooling cases and reference cases. For SrO, fraction of release into containment is largely less than reference case. This is due to ex-vessel release fraction of nonvolatile fission products is more than noble gas and volatile fission products (Huang et al., 2009). In ex-vessel cooling cases of SBO, SBLOCA and LOFW, RCS pressure is low during core melts, so less volatile fission products deposit in RCS comparing with reference cases. This leads to that more volatile fission products release into
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Table 10 Key events and in-vessel release fraction with availability of cavity flooding when core exit temperature exceeds 923.15 K.
Start time of cavity flooding (s) Water level in cavity reaches 11.4 m (s) Water level in RPV is recovered (s) Clad oxidation fraction In-vessel release fraction of noble gas In-vessel release fraction of CsI In-vessel release fraction of SrO
SBO
LOFW
SBLOCA
LBLOCA
9184 11,111 – 0.529 1.0 0.994 1.67E−2
4378 6519 – 0.618 1.0 0.995 1.88E−2
6150 8187 – 0.590 1.0 0.984 5.23E−2
700 2555 – 0.301 1.0 0.997 1.70E−2
Table 11 Comparison of fraction of release into containment between ex-vessel cooling cases and reference casesa .
Noble gas (ex-vessel cooling) Noble gas (reference cases) CsI (ex-vessel cooling) CsI (reference cases) SrO (ex-vessel cooling) SrO (reference cases) a
SBO
LOFW
SBLOCA
LBLOCA
0.996 0.995 0.425 0.198 9.72E−3 8.98E−2
0.990 0.996 0.347 0.102 6.86E−3 6.23E−2
0.997 0.997 0.397 0.183 2.07E−2 8.30E−2
0.997 0.996 0.635 0.642 9.98E−3 3.10E−2
Ex-vessel release is computed in reference cases, duration of ex-vessel release is 10 h.
containment. Nevertheless, success of IVR avoids direct containment heating (DCH) in cases of core melts at high pressure and largely reduces probability of containment failure. Finally, fraction of release into environment can be largely reduced. In order to evaluate the coupling relationship between thermohydraulics and in-vessel release of fission products, SBO is chosen to be analyzed in detail. Before core exit temperature exceeds 923.15 K, accident progression is the same to reference case. When core exit temperature exceeds 923.15 K, cavity flooding system is activated and one PORV is intentional opened at 9197 s. Water level in cavity reaches 11.4 m at 11,111 s. Subsequently, flux of cavity flooding is kept and excessive water is drained through cavity vent (Fig. 19). During this phase, water level in RPV drops rapidly and core uncovers. Accumulator is activated at 10,502 s, so water level in RPV increases a little. However, due to continuous evaporation of coolant and accumulator is depleted, water level in RPV still drops and core temperature increases. Finally, core relocates at 29,856 s (Fig. 20). Core residual heat transfers to cavity water via RPV wall, which keeps RPV lower head integrated. As shown in Fig. 21, in-vessel release fraction of fission products is very large. For noble gas and CsI, release is mostly occurred dur-
Fig. 19. Water level in cavity and heat transfer of RPV to cavity water in case of SBO with ex-vessel cooling.
Fig. 20. Water level in RPV and maximum temperature of core in case of SBO with ex-vessel cooling.
Fig. 21. In-vessel release fraction of fission products in case of SBO with ex-vessel cooling.
G.F. Huang et al. / Nuclear Engineering and Design 240 (2010) 3888–3897
ing core melt. After core relocates, the increased in-vessel release fraction is about 0.05. In the end, in-vessel release fraction of noble gas and CsI are 1.0 and 0.997, respectively. For SrO, the release trend is flatter than noble gas and CsI. In-vessel release fraction of SrO is only 0.0167 owing to nonvolatility. According to above analysis results of ex-vessel cooling, core melt and relocation are not ceased. In-vessel release fraction of noble gas and CsI are almost equal to reference case, and in-vessel release fraction of SrO is bigger than reference case. So, ex-vessel cooling has no mitigation effect on in-vessel release. Nevertheless, success of IVR can largely reduce fraction of release into environment.
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(3) Ex-vessel cooling is used to prevent RPV failure, it is not an effective countermeasure to control in-vessel release of fission products owing to inevitable core melt and relocation. In-vessel release fraction in ex-vessel cooling case is almost equal to reference case. Nevertheless, success of IVR avoids DCH in cases of core melts at high pressure and largely reduces probability of containment failure, which can largely reduce fraction of release into environment. Acknowledgement The authors acknowledge National Basic Research Program of China (no. 2009CB724301) support.
5. Conclusions References The mitigation effect of different severe accident management countermeasures on in-vessel release of fission products is studied for Chinese 600 MW PWR. According to the analytical results, it can be concluded as follows: (1) Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products. Implementation of feed-and-bleed in primary circuit when core exit temperature exceeds 923.15 K, in-vessel release fraction of noble gas is 1E−2 orders of magnitude, in-vessel release of other fission products can be ignored. With later time to implement the countermeasure, accident progression is ceased later, so final in-vessel release fraction of fission products is bigger. (2) Feed-and-bleed in secondary circuit has mitigation effect on in-vessel release of fission products for most accidents that can cease core melt progression. For SBO, LOFW and SBLOCA, water level in RPV is recovered owing to SG reflux-condensation and accumulator injection as a result of feed-and-bleed in secondary circuit. Accident progression is ceased and in-vessel release is stopped. Implementation of feed-and-bleed in secondary circuit when core exit temperature exceeds 923.15 K, there are not fission products release. Bigger in-vessel release fraction of fission products is produced with later time of countermeasure implementation. However, in case of LBLOCA, water level in RPV is not recovered due to loss more coolant. Core melt and relocation are occurred. In-vessel release of fission products is not mitigated.
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