Development of rapid atmospheric source term estimation system for AP1000 nuclear power plant

Progress in Nuclear Energy 81 (2015) 264e275

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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene

Development of rapid atmospheric source term estimation system for AP1000 nuclear power plant Yunfei Zhao, Liguo Zhang*, Jiejuan Tong Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084, PR China

a r t i c l e i n f o

a b s t r a c t

Article history: Received 22 December 2014 Received in revised form 2 February 2015 Accepted 10 February 2015 Available online 6 March 2015

The development of rapid atmospheric source term estimation system (RASTES) for AP1000 nuclear power plant (NPP) is described. The system is designed to give rapid estimation of the amount of fission products released to the atmosphere, and to provide necessary input for fission products atmospheric dispersion simulation and public dose calculation, during a nuclear accident. The estimation method in RASTES mainly refers to NUREG-1228. However, some adjustments are made based on an investigation of studies in nuclear emergency response and design features of AP1000 NPP during development of RASTES. These adjustments enable the system to be more flexible and realistic in atmospheric source term estimation, and more applicable to AP1000 NPP. These adjustments include modifications and validations of aerosol removal factors in reactor containment in both natural process and spray process, as well as addition of aerosol removal factors in passive containment cooling process. Besides, three reactor core damage assessment methods are integrated into atmospheric source term estimation. In addition, containment radiation levels for specified fuel damage which are used in core damage assessment are recalculated. The calculation function of RASTES is confirmed through comparison with RASCAL. Atmospheric source terms in different fission product removal processes are also calculated and compared. The result shows that passive containment cooling is able to further reduce aerosol release by about 50% compared with natural process. Development of RASTES will enhance the capability of nuclear emergency response of AP1000 NPP. © 2015 Elsevier Ltd. All rights reserved.

Keywords: Nuclear emergency response Atmospheric source term estimation Aerosol removal factor Reactor core damage assessment Containment radiation level AP1000 nuclear power plant

1. Introduction Since Three Mile Island accident, Chernobyl nuclear accident and especially Fukushima accident, more emphasis has been put on the capability of emergency response after a nuclear accident occurs. Estimation of the amount of fission products released to the atmosphere, which is usually termed atmospheric source term, constitutes an important part of nuclear emergency response. The estimation result provides necessary input for atmospheric dispersion simulation of fission products and public dose calculation. Based on the dose calculated, corresponding emergency response actions are recommended, such as whether people should evacuate or shelter in place. As a representative of generation-III advanced reactors, AP1000 nuclear power plant (NPP) has drawn

* Corresponding author. Tel.: þ86 10 62792767. E-mail addresses: [email protected] tsinghua.edu.cn (L. Zhang). http://dx.doi.org/10.1016/j.pnucene.2015.02.008 0149-1970/© 2015 Elsevier Ltd. All rights reserved.

(Y.

Zhao),

lgzhang@

numerous attention because of its passive safety feature, and two reactors are under construction currently in China. To enhance the capability of nuclear emergency response of AP1000 NPP, an emergency assisting system needs to be developed. The system is expected to consist of rapid atmospheric source term estimation, fission products atmospheric dispersion simulation and public dose calculation. In this paper, development of the rapid atmospheric source term estimation system (RASTES) for AP1000 NPP, which constitutes one part of the integrated system, is described. In fact, numerous activities had been carried out in the field of atmospheric source term for pressurized water reactor (PWR) and boiling water reactor (BWR) early in the development history of nuclear energy. In 1962, Atomic Energy Commission of U.S. issued Technical Information Document (DiNunno et al., 1962), Calculation of Distance Factors for Power and Test Reactors, which was mainly for evaluation of site suitability. In 1975, Nuclear Regulatory Commission of U.S. (U.S. NRC) issued Reactor Safety Study (Lewis et al., 1975), which applied probabilistic risk assessment method to examine the safety level of reactors. Although the primary purposes

Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275

of the two reports differed from atmospheric source term estimation in emergency response, they still provided valuable reference data about fission products release in nuclear reactors. In 1988, Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents (MeKenna and Giltter, 1988) was released by U.S. NRC. The report had a detailed description of a rapid and simple method to estimate atmospheric source term. Based on studies before the publication, it recommended some useful parameters, such as aerosol fission product removal factors in reactor containment, which would be used during the estimation process. Not only the method to estimate atmospheric source term, but also a body of the parameters recommended in the report had a wide application in subsequent nuclear emergency guiding reports and nuclear power plant safety studies. These included RTM-96: Response Technical Manual which was released by U.S. NRC (McKenna et al., 1996), Generic Assessment Procedures for Determining Protective Actions during a Reactor Accident which was released by IAEA (IAEA-TECDOC-955, 1997), and Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants (NUREG1150) which was finished by U.S. NRC (Ross et al., 1990). Furthermore, Radiological Assessment System for Consequence Analysis (RASCAL) was developed over 25 years ago and recommended by U.S. NRC (Ramsdell et al., 2012). In the system, the module of atmospheric source term mainly referred to the method and parameters recommended in NUREG-1228, with tiny modifications. During the process of atmospheric source term estimation, one important step is to determine status of reactor core in order to calculate release of fission products from reactor pressure vessel (RPV). In 1999, Westinghouse Owners Group published Westinghouse Owners Group Core Damage Assessment Guidance (Lutz, 1999), in which core damage assessment was thoroughly depicted. The report was regarded as a revision of Westinghouse Owners Group Post-Accident Core Damage Assessment Methodology (Westinghouse Electric Company LLC, 1984), which was published earlier in 1984. RTM-96 and IAEA-TECDOC-955 also proposed some core damage assessment methods. Based on these methods, three major reactor core damage assessment methods can be summarized: based on reactor core exit temperature, based on containment radiation level, and based on time duration that reactor core is uncovered. Although the framework of nuclear emergency response has been established in guiding reports mentioned above and RASCAL has been developed, they cannot be directly applied in the development of RASTES because of the design changes of AP1000 NPP. For example, in AP1000 NPP, reactor containment is cooled by passive containment cooling system (PCS) after an accident, which is different from large dry containment of traditional PWRs. This change will affect fission products removal in containment and ultimately affect atmospheric source term. Another change of AP1000 NPP is that the spray system in containment is non-safetyrelated and can last for less than three hours, which also has an effect on fission products removal in containment. These changes should be considered in the development of RASTES. Besides, since publication of NUREG-1228 and development of RASCAL, further studies regarding fission products removal in reactor containment have been conducted and these later understandings should be reflected in RASTES. In RASCAL, the module of reactor core damage assessment is not available, and this input mainly relies on expert judgments. If automatic assessment of reactor core is integrated into RASTES, the system will be more convenient and flexible in real application. These considerations form the basis and highlights of the paper. In addition to the work on atmospheric source term estimation, some progresses have also been achieved in recent years in the field of fission product atmospheric dispersion and public dose calculation (Bocquet, 2012; Cheng et al., 2008; Han

265

et al., 2012; Lauritzen et al., 2006; Nielsen et al., 1999; Sauer, 2007). As the paper focuses on atmospheric source term estimation, these progresses are not introduced at length in the paper. The paper is organized as follows. Section 2 has a brief description of atmospheric source term estimation method applied in RASTES. Section 3 introduces later research results regarding fission products removal in reactor containment, as well as the effect of AP1000 NPP passive containment cooling (PCC) process on fission products removal in containment. Corresponding adjustments in RASTES compared with NUREG-1228, RTM-96, IAEATECDOC-955 and RASCAL are presented. In Section 4, the calculation function of RASTES is confirmed through comparison with RASCAL for 13 calculation examples. In Section 5, atmospheric source terms in different fission products removal processes are compared and discussed. The paper is concluded in Section 6. 2. Atmospheric source term estimation method The atmospheric source term estimation method applied in RASTES mainly refers to NUREG-1228. In the method, nuclear power plant (NPP) is divided into several compartments, such as reactor pressure vessel (RPV), steam generator, reactor containment, auxiliary building and the atmosphere. By computing transfer of fission products between connected compartments and removal in each compartment, the amount of fission products that are released to the atmosphere is obtained. The method can be summarized into Eq. (1).

0 atmospheric source termi ¼ FPIi  CRFi  @

n Y

1 RDFði;jÞ A  EFi

j¼1

(1) where FPIi ¼ core or coolant inventory of radionuclide i CRFi ¼ ratio of the amount of radionuclide i released from core to the inventory in core RDF(i,j) ¼ ratio of the amount of radionuclide i available for release after reduction mechanisms to that available for release before reduction mechanisms EFi ¼ ratio of the amount of radionuclide i released to atmosphere to that available for release From Eq. (1), it is easy to see that the whole estimation process can be completed in 4 steps, as illustrated in Fig. 1. 2.1. Step 1: calculate fission products inventory in reactor core and primary coolant In chapter 11 of AP1000 Final Safety Evaluation Report (AP1000 FSER, 2004) Design Control Document (DCD) R17 Tier 2, realistic primary coolant activity concentrations of 61 radionuclides are provided, and based on primary coolant mass of 195044.719 kg, total amount of radionuclides in primary coolant are obtained. In Appendix 15A of AP1000 FSER DCD R17 Tier 2, the inventory of 64 radionuclides in reactor core immediately after reactor shutdown is calculated based on the assumption of three-region equilibrium cycle at end of life, after continuous operation at 2 percent above full core thermal power. Fission products inventory provided in AP1000 FSER are a little different from the nominal inventory proposed in RTM-96 and IAEA-TECDOC-955, but the difference is very small. Part of fission products inventory in AP1000 NPP reactor core and primary coolant are listed in Table 1 for illustration.

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Fig. 1. Flow chart of atmospheric source term estimation.

2.2. Step 2: calculate the amount of fission products released from pressure vessel In RASTES, the release process of fission products from pressure vessel is separated into a number of release steps, and time duration of one release step is set to be 15 min. Calculation in this step can be divided into 2 situations. In the first case, reactor core is still covered with water, which indicates that it has nearly no damage. As a result, the majority of fission products that have the potential to be released locate in primary coolant. Radionuclides in primary coolant are assumed homogenously distributed. In this case, the amount of fission products released in each release step can be computed by combining the mass of coolant released in the release step and fission products concentration in RPV. Dilution of fission products concentration in RPV by clean injection water is considered in the calculation. In the second case, reactor core is uncovered, and fuels are damaged. As a result, fission products stored in fuels are released from RPV. Calculation in this case can be further divided into 2 steps. First, fission products release in typical large break loss of coolant accident (LBLOCA) is introduced to represent the actual

fission products release process during an accident. This mainly refers to the analysis result in Accident Source Term for Light-Water Nuclear Power Plants which was published by U.S. NRC (Soffer et al., 1995). It generalized reactor core damage progression and radionuclide release fractions under different fuel damage phases for LBLOCA, based on results obtained in NUREG-1150 and some other MELCOR and STCP calculations. The analysis result in NUREG-1465 had an extensive application in source term estimation, such as RTM-96, IAEA-TECDOC-955, RASCAL and AP1000 FSER. Table 2 reproduces the result. Second, actual fuel damage during an accident is assessed. WCAP-14696, RTM-96 and IAEA-TECDOC-955 recommended several reactor core assessment methods respectively. These methods can be summarized into three major classes, which are based on reactor core exit thermocouple readings, based on containment radiation level and based on time duration that reactor core is uncovered.

Table 2 PWR accident progression and fission products release in LBLOCA.a Nuclide group

Table 1 Part of fission products inventory in AP1000 NPP reactor core and primary coolant.a Radionuclide

Reactor core inventory (Bq)

Primary coolant inventory (Bq)

I-131 I-133 Cs-137 Sr-90 Kr-85 Kr-88 Xe-133 Xe-135

3.561018 7.361018 4.181017 6.731018 3.921016 2.641018 7.031018 1.791018

7.94109 1.301010 1.081010 7.22107 1.80108 9.38107 7.94109 2.24108

a

Data comes from AP1000 FSER.

Noble gases (Kr, Xe) Halogens (I, Br) Alkali metals (Cs, Rb) Tellurium group (Te, Sb, Se) Barium, strontium (Ba, Sr) Noble metals (Ru, Rh, Pd, Mo, Tc, Co) Cerium group (Ce, Pu, Np) Lanthanides (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) a

Data comes from NUREG-1465.

PWR release fraction Cladding failure (cladding gap release) (0.5 h)

Core melt (in-vessel release) (1.3 hours)

0.05 0.05 0.05 0 0 0

0.95 0.35 0.25 0.05 0.02 0.0025

0 0

0.0005 0.0002

Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275

In AP1000 NPP, there are 42 thermocouples located at reactor core exit, and 4 containment radiation detectors installed on the inner surface of the containment wall. Before the assessment, coolant temperatures at the core exit that indicate occurrence of cladding failure or core melt are determined. As well, containment radiation levels for specified degree of core damage, such as 100% cladding failure and 100% core melt, are also determined. When an accident occurs, by comparing actual core exit thermocouple readings and containment radiation level with the specified set points, the phase of reactor core damage (cladding failure or core melt), and the degree of core damage can be obtained. The assessment process is demonstrated in Fig. 2(a). In the figure, CRM and CET represent containment radiation monitor and core exit thermocouple readings respectively. During development of the system, the set points of core exit temperature refer to the ones proposed in WCAP-14696. The set points of containment radiation level for typical PWR and BWR containments are proposed in IAEA-TECDOC-955, RTM-96 and RASCAL. However, the containment of AP1000 NPP is passively cooled, which is different in aerosol removal from traditional large dry containment of PWRs. Besides, the location of containment radiation monitors in AP1000 NPP is different from traditional containment. As a result, the set points for AP1000 NPP are recalculated using point-kernel integration method, and this will be discussed in more detail in section 3.

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When applying the method based on time duration that core is uncovered, set points that are able to indicate uncover and recovery of reactor core are determined. These set points include pressure vessel water level, hot leg coolant temperature, core exit coolant temperature, and minimum core injection flows needed to remove decay heat. When an accident occurs, by comparing the actual indications obtained from the plant with these set points, time duration that reactor core is uncovered can be determined. Comparing the time duration with the typical LBLOCA progression concluded in NUREG-1465, reactor core damage status and the degree of damage can be obtained. The integral assessment process is demonstrated in Fig. 2(b). In the figure, RVL3 and the terms with similar form represent set points of the indicators. Each of the three methods is able to give an assessment result about reactor core damage. This strategy of providing three alternative choices on the one hand enables emergency responders to choose the most reasonable assessment result based on accident condition and their judgments. On the other hand, it provides a redundancy for fuel damage assessment. The fraction of fission products that are released from RPV in each release step can be calculated through the two steps. Combining the calculation result with fission products inventory in reactor core, the amount of fission products that are released from RPV is obtained.

Fig. 2. Flow chart of reactor core damage assessment.

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2.3. Step 3: calculate fission products removal on release paths Fission products that escape from pressure vessel may be released to the atmosphere through three major paths, which are reactor containment, steam generator, and auxiliary building. If the release is through reactor containment, all of the fission products except for noble gases will undergo removal by natural process and spray process. In addition, one of the prominent features of AP1000 NPP is its passively cooled containment. This feature not only helps keep the integration of reactor containment, but also enhances aerosol removal in containment. Detailed analysis of aerosol removal in containment will be discussed in section 3. If the release is through steam generator, all of the fission products except for noble gases will be partitioned into two phases. They are steam phase, which is available for release to the atmosphere, and water phase, which is retained in steam generator. Partition factor depends on whether the steam generator tube break above or below water level. If the release is through auxiliary building, only filters in the building are effective in removing aerosol fission products. 2.4. Step 4: calculate fission products release to the atmosphere Release paths of fission products in this step are consistent with the ones described in step 3. If the release is through reactor containment, the amount of fission products that are released to the atmosphere can be obtained by combining the amount of fission products retained in containment atmosphere after removal and escape fraction from containment. In this case, escape fraction is expressed by dividing the volume released from containment at unit time by the total containment volume, such as 0.1%/day, which usually represents containment leakage under design pressure. If the release is through steam generator, the amount of fission products that are released to the atmosphere can be obtained by combining the amount of fission products retained in steam phase after partitioning and escape fraction from steam generator. In this case, escape fraction can be calculated by dividing the mass of steam released from steam generator by the total mass in steam generator. Auxiliary building is assumed to have no retention on fission products release. For illustration, the source term calculation algorithm for LOCA in which only PCS is operated is listed as follows. i. Determine release fraction of radionuclide j from reactor vessel in release step i Rij M represents total number of radionuclides, N represents total release steps. ii. Calculate initial release of radionuclide j from reactor vessel in release step i (without decay calculation) IRij ¼ Ij $Rij Ij represents inventory of radionuclide j in reactor core or primary coolant. iii. Calculate radionuclide release in release step i after decay calculation DRi ¼ DecayðIRi Þ Decay() is the decay calculator. iv. Calculate inventory of radionuclide j in reactor containment that is available for release to the atmosphere in release step i Calculate the first step if radionuclide j is aerosol SI1j ¼ DR1j $expðPCC1 $StepDuraÞ else

SI1j ¼ DR1j Calculate subsequent release steps SIi ¼ DecayðSIi1 Þ$ð1  EFÞ þ DRi if radionuclide j is aerosol SIij ¼ DRij $expðPCCi $StepDuraÞ EF is the escape fraction from containment to the atmosphere in each release step PCCi is aerosol removal coefficient in release step i achieved by PCS StepDura is the time duration of one release step v. Calculate radionuclide release to the atmosphere in release step i SRi ¼ SIi $EF vi. Calculate total amount of radionuclide released to the atmosphere P TRj ¼ N i¼1 SRij 3. Adjustments in RASTES As introduced in section 1, further studies in the field of aerosol removal in reactor containment have been conducted. Besides, there are some design changes of AP1000 NPP containment, such as passively cooled containment and non-safety-related spray system. As a result, some adjustments are made during development of RASTES, compared with RASCAL and current nuclear emergency guiding reports, such as NUREG-1228, RTM-96 and IAEA-TECDOC955. The major adjustments fall into two aspects. First, aerosol removal factors in containment are modified, based on investigation of later studies in this field and consideration of the design changes of AP1000 NPP containment. Second, containment radiation levels for specified fuel damage are recalculated, taking into account the location change of radiation monitors in AP1000 NPP containment and modifications of aerosol removal factors in containment. 3.1. Aerosol removal in containment When fission products are released to the atmosphere through reactor containment, a number of mechanisms will be effective to remove aerosols from containment atmosphere. Aerosol removal process in AP1000 NPP containment can be divided into three categories, which are natural process, passive containment cooling (PCC) process and spray process, according to the operating status of corresponding systems. Three parameters can be used to represent the magnitude of aerosol removal in containment. Usually, aerosol removal process can be expressed in Equation (2):

ZT mðTÞ ¼ m0 expð 

lðtÞdtÞ

(2)

0

where m0 ¼ mass of aerosols before removal at time 0; m(T) ¼ mass of aerosols after removal at time T; l(t) ¼ aerosol removal coefficient, a function of time duration of removal. As a result, aerosol removal coefficient is able to reflect the magnitude of aerosol removal. The second parameter can be expressed by the ratio of m(T) divided by m0, which is usually called aerosol reduction factor (RDF). The third parameter is the inverse of RDF, which is termed aerosol decontamination factor (DF). Because different parameters are used in literature, all of the three parameters will appear in following sections of the paper.

Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275 Table 3 Summary of aerosol reduction factors in natural process. NUREG-1228

RTM-96, IAEA-TECDOC-955

Holdup time t (h)

RDF

Holdup time t (h)

RDF

Holdup time t (h)

0.5 2e12 24

0.4 0.04 0.01

<1 1e12 >12

0.75 0.36 0.03

<1.75 1.75e2.25 >2.25

a

RASCAL

exp(1.2t)a exp(0.64t)a exp(0.15t)a

This is the reduction factor multiplier. There is a lower limit of 0.001 on RDF.

3.1.1. Natural process and PCC process When spray system and passive containment cooling system (PCS) in AP1000 NPP containment are not operated, several removal mechanisms, such as gravitational settling, thermophoresis and diffusiophoresis, are still effective in removing aerosols. NUREG-1228 recommended aerosol reduction factors based on studies before its publication (Hilliard and Postma, 1981). RTM-96 and IAEA-TECDOC-955 adjusted the reduction factors recommended in NUREG-1228 to be representative of the analysis result in NUREG-1150. In RASCAL, aerosol DFs also refer to the analysis result in NUREG-1150. All of the reduction factors or calculation formulae are summarized in Table 3. In 1995, U.S. NRC released A Simplified Model of Aerosol Removal by Natural Process in Reactor Containments (Powers et al., 1995). The report took advantage of the state-of-the-art knowledge about aerosol removal by natural process at that time to have a detailed analysis of aerosol behavior in containment, such as aerosol coagulation and deposition. Based on the analysis, the mechanistic model to calculate aerosol removal by natural process was constructed. The model was verified to be effective in predicting aerosol removal in natural process by comparing the predicted results with the ones predicted with CONTAIN code, as shown in Fig. 3. Uncertainty distribution of aerosol removal coefficients was analyzed by Monte Carlo method. Based on the analysis, median, 90th and 10th percentile values of the uncertainty distribution of aerosol removal coefficients were correlated with time duration of removal and reactor thermal power, as in Table 4. Four fission products release phases, which

269

were gap release, in-vessel release, ex-vessel release and late release, were analyzed in the report. Each release phase had independent aerosol removal coefficients during its own release process, while the removal coefficients will become the same with earlier release phases after termination of its own release process. In this paper, aerosol removal coefficients in in-vessel release phase are chosen to represent aerosol removal in natural process because of the large and early release of fission products in this phase. Full thermal power of AP1000 NPP is 3400 MW, so aerosol removal coefficients for AP1000 NPP is easy to obtain by the correlations in Table 4. Passive containment cooling is one of the prominent features of AP1000 NPP. In essence, aerosols are subject to the same removal mechanisms, such as gravitational settling, thermophoresis and diffusiophoresis, in PCC process as in natural process. However, cooling of the steal containment increases the temperature gradient in areas near the inner surface of containment wall and enhances steam condensation on the inner surface of containment wall, and therefore enhances thermophoresis and diffusiophoresis of aerosols. Besides, cooling of containment enhances air and steam convection in containment atmosphere, and therefore enhances aerosol deposition by gravitational settling. As a result, PCC process is expected to give rise to higher aerosol removal coefficients than natural process. In AP1000 NPP FSER DCD R17 Tier 2 Appendix 15B, aerosol removal coefficients following a postulated loss-of-coolantaccident (LOCA) are calculated as follows. First, the aerosol removal model is well established and the removal processes in the model are confirmed in many separate effect experiments. Then, the environment that would be expected to exist as a result of LOCA is analyzed (such as containment thermal-hydraulic data calculated by MAAP). Last, aerosol removal coefficients are obtained by combining the removal model and containment environment. In Table 5, part of the calculation result of aerosol removal coefficients in AP1000 NPP FSER are selected for illustration. To have a visual comparison of these aerosol removal parameters, all of the parameters are transformed into the form of aerosol decontamination factors (DFs) and are plotted in Fig. 4 as a function of time duration of removal.

Table 4 Correlation of aerosol removal coefficients with reactor thermal power.

Fig. 3. Comparison of aerosol removal predicted with CONTAIN code and with the mechanistic model in NUREG/CR-6189 (Fig. 20 in NUREG/CR-6189).

Time interval (s)

Correlationsa (h1)

0e4680

lð90Þ ¼ 0:0505 þ 0:94  106 P b lð50Þ ¼ 0:0257 þ 3:87  106 P lð10Þ ¼ 0:0166 þ 3:49  106 P

4680e11,880

lð90Þ ¼ 0:1146 þ 371:9=P lð50Þ ¼ 0:0948 þ 141:2=P lð10Þ ¼ 0:0472 þ 62:0=P

11,880e40,680

lð90Þ ¼ 0:378 þ 161:6=P lð50Þ ¼ 0:269 þ 141:2=P lð10Þ ¼ 0:068 þ 81:8=P

40,680e78,200

lð90Þ ¼ 0:210 þ 50:6=P lð50Þ ¼ 0:144 lð10Þ ¼ 0:0915½1  expð2:216P=1000Þ

78,200e98,200

lð90Þ ¼ 0:0933 þ 12:0=P lð50Þ ¼ 0:0838 lð10Þ ¼ 0:0377

98,200e118,200

lð90Þ ¼ 0:0717 þ 10:8=P lð50Þ ¼ 0:0669 lð10Þ ¼ 0:0277

a lð90Þ, lð50Þ and lð10Þ are 90th median and 10th percentile values of the uncertainty distribution respectively. b P is reactor thermal power (unit is MW).

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Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275 Table 5 Aerosol removal coefficients as a result of passive containment cooling. Time interval (h)

Removal coefficient (h1)

0.433e0.631 0.893e1.033 1.991e 2.067 4.928e5.362 9.553e11.189 14.937e17.610

0.901 0.699 0.429 0.435 0.530 0.506

From the figure, it is easy to see that aerosol DFs in natural process applied in current nuclear emergency guiding reports, such as NUREG-1228 and RTM-96, are very rough, which are merely segmented functions of aerosol removal time. Comparison of aerosol DFs in natural process shows that at the beginning of aerosol removal process, all of the DFs are higher than the 10th, 50th and 90th percentile values calculated by the simplified model in NUREG/CR-6189. Particularly, aerosol DFs recommended by RASCAL and NUREG-1228 are even higher than the ones achieved by passive containment cooling system, which is unrealistic. As aerosols are released continuously from RPV at the beginning of the removal process and it needs time for aerosol coagulation to facilitate removal, aerosol DFs cannot be very high at this period and the values recommended in NUREG/CR-6189 are more reasonable. There are differences between aerosol DFs from various sources. However, these differences are not great. It should also be noticed that although both of RASCAL and RTM-96 refer to the analysis result in NUREG-1150, aerosol DFs from the two sources are different. In fact, five different uncertain distributions of aerosol DFs are given independently by five experts respectively in NUREG-1150, and these distributions are combined to give an aggregate distribution. As a result, the difference may be explained by adoption of different aerosol DF distributions in RASCAL and RTM-96. The difference also illustrates the uncertainty about aerosol removal in reactor containment from another side. Median aerosol DFs calculated by the simplified model in NUREG/CR-6189 have similar trend with the ones proposed in RTM-96, which represent one of the analysis results in NUREG-1150. Compared with natural process, PCC process is more effective in removing aerosols in containment atmosphere, especially in the late stage of the removal process, which can be higher by a factor of about 1000. In RASTES, median values of aerosol decontamination

factors in NUREG/CR-6189 are adopted to represent aerosol removal process when neither spray system nor PCS is operated. Aerosol decontamination factors in AP1000 PCC process, taken from AP1000 NPP FSER, are adopted to represent aerosol removal process when PCS is operated. 3.1.2. Spray process When spray system in containment is operated, spray water will wash out aerosols suspended in containment atmosphere. NUREG1228 proposed aerosol reduction factors, which were based on studies before its publication (Hilliard and Postma, 1981). RTM-96 and IAEA-TECDOC-955 adopted the factors proposed in NUREG1228, with tiny modifications taking account of the analysis result in NUREG-1150. RASCAL proposed simple aerosol reduction factor calculation formulae, mainly referring to MENU-TACT code which is also designed to estimate atmospheric source term during reactor accident emergency (Sjoreen et al., 1987). Aerosol reduction factors or calculation methods are summarized in Table 6. In 1993, U.S. NRC released A Simplified Model of Aerosol Removal by Containment Sprays (Powers and Burson, 1993). In the report, aerosol removal mechanisms by water droplets, such as impaction, interception and diffusion, were analyzed in detail, and based on the analysis, a mechanistic model was established. Important factors that would contribute to uncertainty of aerosol removal in spray process were also discussed. Uncertainty distribution of aerosol removal in spray process was developed with Monte Carlo method, and aerosol removal coefficients of three levels of uncertainty (10th, 50th and 90th percentile) were correlated as in Formulae 3 through 5. Parameters in the formulae for 10th, 50th and 90th percentile calculations differ from each other.

 h  i l ¼ l mf ¼ 0:9 l=l mf ¼ 0:9

(3)

  ln l mf ¼ 0:9 ¼ A þ B ln Q þ CH þ DQ 2 H þ EQH2 þ FQ

(4)

  c   c   mf mf þ l=l mf ¼ 0:9 ¼ ½a þ b log10 Q  1  0:9 0:9

(5)

where Q ¼ spray water flux into reactor containment, cm3/cm2 s H ¼ falling distance of water droplets, cm mf ¼ aerosol reduction factor A, B, C, D, E, F ¼ parameters to calculate lðmf ¼ 0:9Þ a, b, c ¼ parameters to calculate l=lðmf ¼ 0:9Þ AP1000 NPP includes a non-safety-related spray system in containment which is used for fire protection, rather than a safetyrelated spray system. The spray system can last for less than three hours as a result of loss of spray water supply. In AP1000 FSER DCD R17 Tier2 chapter 6, the calculation formula for aerosol removal

Table 6 Summary of aerosol reduction factors in spray process.

Fig. 4. Comparison of aerosol removal during natural process.

NUREG-1228

RTM-96, IAEA-TECDOC-955

Holdup time t (h)

RDF

Holdup time t (h)

RDF

Holdup time t (h)

0.5 2e12 24

0.03 0.02 0.002

<1 1e12 >12

0.03 0.02 0.01

<0.25 >0.25

a

RASCAL

exp(12t)a exp(6t)a

This is the reduction factor multiplier. There is a lower limit of 0.03 on RDF.

Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275

coefficients during spray process in Standard Review Plan (SRP) is adopted as Eq. (6):

l ¼ 3hfE=2Vd

(6)

where h ¼ falling distance of water droplets, m f ¼ spray water flow rate, m3/h E ¼ collection efficiency V ¼ volume of containment exposed to spray, m3 d ¼ average of spray droplet diameter, m In AP1000 FSER and SRP, E/d is identified to be 10 m1 until aerosol concentration is reduced by a factor of 50. After a DF of 50 is reached, E/d is reduced by a factor of 10. This calculation formula can be validated by the experiment made in Hanford Engineering Development Laboratory (Postma and Coleman, 1970). h, f and V in the experiment are 10.3 m, 11.13m3/h and 596 m3 respectively. Putting these parameter values into formula 5, aerosol removal coefficient before and after DF reaches 50 are calculated to be 2.88 h1 and 0.288 h1 respectively. The experiment result shows that in the early stage of spray process, aerosol removal coefficient is 2.60 h1, and in the late stage of spray process, aerosol removal coefficient is 0.308 h1. The predicted result agrees very well with the experiment result. h, f and V in AP1000 NPP are 38.1 m, 227.12 m3/h and 48138.64 m3 respectively, so aerosol removal coefficients are calculated to be 2.7 h1 and 0.27 h1 before and after DF of 50 is reached respectively. Parameters needed in validation and calculation of aerosol DFs in spray process for AP1000 NPP are summarized in Table 7. The recommended aerosol decontamination factors in spray process are plotted in Fig. 5 for comparison. From Fig. 5, three points are easy found. First, uncertainty of DFs in spray process is much greater than in natural process, which can be shown by the great differences between 10th, 50th and 90th percentile values calculated by the simplified model in NUREG/CR-5966. This also indicates that aerosol removal in spray process is more complicated than in natural process. Second, the differences of DFs from various sources are very small in the early stage of the process, in which spray may be most effective in removing aerosols. Third, there are upper limits on DFs in RTM-96, NUREG-1228 and RASCAL. This is reasonable because in emergency response, a relatively conservative estimation of atmospheric source term is needed, considering the great uncertainty in predicting aerosol removal. DFs in AP1000 FSER are also subject to an upper limit, but this results from loss of spray water supply.

In RASTES, DFs in AP1000 FSER are chosen to describe aerosol removal in spray process, as a result of the good agreement with experimental results and the non-safety-related design.

3.2. Recalculation of containment radiation levels for specified fuel damage There are 4 radiation monitors installed on the wall within AP1000 NPP containment. As described in section 2.2, containment radiation monitors provide an alternative approach to assess reactor core damage. In the approach, containment radiation levels for specified reactor core damage need to be pre-determined for comparison with actual monitor readings in a real accident. In RTM-96, IAEA-TECDOC-955 and RASCAL, pre-determined containment radiation levels for 100% cladding failure and 100% core melt are provided for traditional large dry containment. However, these results are based on the fact that the radiation monitor is located in the dome of the containment, which is different from AP1000 NPP. Furthermore, aerosol decontamination factors have been adjusted for natural process and spray process, and aerosol decontamination factors for PCC process are added in RASTES. The difference of radiation monitor location, and adjustment and addition of aerosol decontamination factors will affect radiation response in containment. As a result, it is necessary to recalculate these predetermined containment radiation levels for AP1000 NPP. In this paper, point-kernel integration method is introduced for the calculation as follows (Li et al., 2004). In the calculation, different conditions in which whether containment spray system or PCS is operated are considered. For the simplest situation, there is a point source of radiation at point T, the strength of which is S0 and isotropic. The radiation detector is arranged at point P, and its distance to point T is a. The projected area of the radiation detector normal to the direction to T is dA. For the condition of without any medium in the circumstance, the photon quantity the detector is able to receive equals:

S0

 dA 4p 2 a

Hanford experiment

AP1000 NPP

Spray water falling distance h (m) Spray water flow rate f (m3/h) Containment volume (m3) Aerosol removal coefficient in early stage in Hanford experiment (h1) Aerosol removal coefficient in early stage calculated through equation 6 (h1) Aerosol removal coefficient in late stage in Hanford experiment (h1) Aerosol removal coefficient in late stage calculated through equation 6 (h1)

10.3 11.13 596 2.60

38.1 227.12 48138.64 e

2.88

2.7

0.308

e

0.288

0.27

(7)

For the condition of atmosphere, the radiation strength is attenuated exponentially. Taking account of the accumulation of scattering radiation, the photon quantity that the detector is able to receive becomes

Table 7 Parameters needed in validation and calculation of aerosol DFs in spray process for AP1000 NPP. Parameters

271

Fig. 5. Comparison of aerosol decontamination factors in spray process.

272

Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275

N ¼ S0

dA Bema 4pa2

(8)

where B ¼ accumulation factor

m ¼ the linear attenuation coefficient for the photon According to the definition, the photon fluence rate for point P becomes

N¼

S0 Bema 4pa2

(9)

This is the equation used for calculation of radiation level for a point source. For a power plant containment, to get the radiation level for a specific radiation detector, we can firstly discretize the containment atmosphere and the containment inner-wall into lots of tiny parts, and then calculate the radiation contribution from every tiny part respectively, and at last add these contributions together. As long as the discretized part is small enough, the difference from actual radiation level can be controlled to an acceptable level. The following equations calculate the radiation level contributed by radionuclide n, for the part of containment atmosphere and the part of inner-wall respectively.

CDðtÞ ¼

XX KE

f  E0  FEðn; KEÞ  m0a  expð  mRÞ

S

 BðE0 ; mRÞ  Cðn; tÞ  CDAðtÞ ¼

XX KE

rdr dc dz 4pR2

(10)

f  E0  FEðn; KEÞ  m0a  expð  mRÞ

S

 BðE0 ; mRÞ  CAðn; tÞ 

dS 4pR2

(11)

where f ¼ energy unit transfer coefficient, 1.6  1013 J/MeV KE ¼ the number of the photon created by the decay of radionuclide n E0 ¼ the energy of the KE numbered photon created by radionuclide n, (MeV) FE(n, KE) ¼ the photon yield of the KE numbered photon created by radionuclide n m0a ¼ the mass energy absorption coefficient for the KE numbered photon created by radionuclide n, (m2/kg) m ¼ the linear attenuation coefficient for the KE numbered photon created by radionuclide n, (m1) R ¼ the distance between the detector and the calculated location, (m) B(E0, mR) ¼ the accumulation factor, and for photons with different energy, the calculation equations are different

BðE0 ; mRÞ ¼ 1 þ mR þ

ðmRÞ2 7E02:4

ðE0  0:5 MevÞ

BðE0 ; mRÞ ¼ 1 þ 1:1mR þ ðmRÞ2

ðE0 < 0:5 MevÞ

Values of KE, E0, and FE (n, KE) refer to the radioactive nuclide database provided in the online cooperative research center of nuclear science and technology (photon coefficient, http://rsh.nst. pku.edu.cn/nuclide/, 2000). Values of m0a and m refer to U.S. National Institute of Standards and Technology (NIST) X-ray attenuation database (Hubble and Seltzer, http://physics.nist.gov/ PhysRefData/XrayMassCoef/tab4.html, 2014). In the calculation, several ideal assumptions are made, part of which are consistent with the ones in RTM-96 and IAEA-TECDOC955. First, radionuclides are released to the containment instantly after the accident. Second, radionuclides are uniformly mixed in containment atmosphere. Third, the radiation monitor is not shielded. Last, radionuclides scrubbed out in spray process cannot be detected by the monitor, while radionuclides removed to innersurface of containment wall can still be detected by the monitor. As the diameter and free volume of AP1000 NPP containment are 39.62 m and 58,333 m3 respectively, so the configuration of the containment is assumed to be with a cylinder which is 39.62 m in diameter and 34.11 m in height, and a hemi-sphere head which is 39.62 m in diameter. Radionuclide release fractions for 100% cladding failure and 100% core melt refer to the values concluded in NUREG-1465. The calculation results for 100% cladding failure when PCS is operated while spray system is not available, and for 100% core melt when neither PCS nor spray system is available, are demonstrated in Figs. 6 and 7 respectively. In the figures, the reference ones applied in RASCAL are also added to show their difference. In RASCAL, containment radiation levels 1 h and 24 h after reactor shutdown are specified. When the time after reactor shutdown is less than 1 h, then radiation level for 1 h is used. When the time after reactor shutdown is between 1 h and 24 h, then the levels for 1 h and 24 h are linearly interpolated to get a new radiation level for the actual time. The two figures show that at the beginning of the accident and a long time after the start of the accident (about 24 h), the difference between the radiation levels is not great. However, in the middle stage of accident progression, radiation levels proposed in RASCAL are larger than the ones recalculated by, as large as 100%. The linear interpolation strategy in RASCAL may not be appropriate, considering that both decay and removal of fission products follow exponential reduction. This conclusion is also supported by the example containment radiation levels shown in WCAP-14696, in which even logarithm of containment radiation level is a concave function of time after reactor shutdown.

(12) (13)

C(n, t) ¼ the concentration of nuclide n in the containment atmosphere CA(n, t) ¼ the concentration of nuclide n on the containment inner-wall

Fig. 6. Containment radiation level for 100% cladding failure (PCS on, spray system off).

Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275

273

Table 9 Comparison of results for example 2 between RASCAL and the integrated system. Typical radionuclide

Xe-133 Kr-85 I-131 I-133 Cs-137 Sr-90

Fig. 7. Containment radiation level for 100% core melt (PCS off, spray system off).

4. Confirmation of calculation function of RASTES In this section, the correctness of calculation function of RASTES is confirmed through comparison with RASCAL. In the confirmation process, parameters embedded in RASTES are set to be the same as the ones in RASCAL. These parameters include such as radionuclide inventory in reactor core and primary coolant, containment radiation level for specified core damage, and aerosol fission product removal factors. Thirteen accident examples are calculated with both RASTES and RASCAL. Radioactive activity released in each release step and total activity released to the atmosphere and are compared between RASTES and RASCAL. All of the deviations of the results calculated with RASTES from the ones calculated with RASCAL are less than 1%. Therefore, the correctness of calculation function of RASTES is confirmed. Tables 8 and 9 demonstrate the results of two confirmation examples. In example 1, reactor is uncovered immediately after reactor shutdown in an LBLOCA, and the process lasts for 1 h until reactor core is recovered by injection water. Release of radionuclides to the atmosphere is through reactor containment, and the release is terminated 2 h after the start of release. In example 2, reactor is shut down immediately after a steam generator tube rupture (SGTR) accident. Reactor core is not damaged, but the release of normal coolant from RPV to the atmosphere through steam generator begins two hours and terminates four hours after reactor shutdown respectively. The leak rate from RPV to the steam generator is 0.03155 m3/s, and the steam leak rate from steam generator is 34 m3/h. The steam generator tube rupture is above water level, and steam is released to the atmosphere through safety release valve. Investigation of other AP1000 NPP related studies about atmospheric source term also indicates that RASTES is effective. Zhu et al.

Total activity released to the atmosphere (107 Bq) RASCAL

The assisting system

10,682 120,768 448 5362 9.24 2.09

10,678 120,768 447 5356 9.23 2.09

(2014) applied the same estimation procedure as in RASTES and NUREG-1465 source term to derive the atmospheric source term for a hypothetical station blackout accident. Unfortunately, as a result of lack of necessary information of the atmospheric source term result in the paper by Zhu, no further comparison can be implemented. The estimated source term by RASTES agrees very well with the study completed by Sholly et al. (2014) for the loss of coolant accident (for example, iodine group release fraction calculated as 0.0011% by RASTES versus 0.0012% by Sholly). 5. Comparison of atmospheric source terms in different fission product removal processes In this section, atmospheric source terms in three situations are calculated to compare the different effect of natural process, PCC process and spray process on fission products release. In the calculation, an LBLOCA is assumed to occur in AP1000 NPP. Water in RPV drains out very quickly, safety injection systems fail to inject water into RPV, and fuels in reactor core are assumed to be uncovered instantly after the accident. Uncover of fuels lasts for two hours before safety injection systems are repaired to inject water into RPV. During the process, all of the fuel claddings fail and fuels melt under high temperature. Fission product release rate from containment is assumed to be 5%/day, which is 50 times higher than design basis release rate, and the first ten hours' release (equivalent to forty release steps) to the atmosphere is calculated. In the first situation of the calculation, it is assumed that neither PCS nor spray system is operated, which means only natural process contributes to fission products removal. In the second situation, PCS is operated while spray system is not available. In the third situation, spray system is operated while PCS fails to start. Releases

Table 8 Comparison of results for example 1 between RASCAL and the integrated system. Typical radionuclide

Total activity released to the atmosphere (1010 Bq) RASCAL

The assisting system

Xe-133 Kr-85 I-131 I-133 Cs-137 Sr-90

16,866 67.54 1529 3018 115.67 4.31

16,865 67.56 1527 3018 115.63 4.32

Fig. 8. Noble gas release under different conditions.

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Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275

of noble gases, iodine and cesium in each release step are shown in Figs. 8 through 10, and total releases of the three categories of fission products in all of the release steps are summarized in Table 10. As noble gases are not affected by any removal process in containment, the difference of atmospheric source terms between the three situations is very small, as shown in Fig. 8. The minor difference can be explained by the fact that part of Xe-135 and Xe135m originate from the decay of I-135, which is subject to removal. Decrease of release with time in the late stage results from decrease of inventory in containment atmosphere as a result of release to atmosphere and decay of noble gases with time. Iodine and cesium have similar release process, as demonstrated in Figs. 9 and 10. During the period of reactor core uncover, fission products are released from RPV continuously to containment, as a result release of fission products in each release step increases with time. When reactor core is recovered with water, release of fission products from RPV terminates, and release in each release step decreases with time as a result of decontamination, decay and leak of fission products to the atmosphere. Both of natural process and PCC process are effective in reducing fission products release. Take cesium as an example, activity release from maximum release step (step 6) to the minimum release step (step 40) is reduced by a factor of about 100 for PCC process, and about 20 for natural process. It is evident that PCC process is more powerful in removing aerosols compared with natural process. For total release, PCS is able to reduce the release of both iodine and cesium by about 50% compared with natural process. The calculation result also shows that if spray system is available during release of fission products from RPV, it will be the most effective tool to reduce aerosol release to the atmosphere. 6. Conclusion In this paper, the development of RASTES is described. It can complete similar functions as the atmospheric source term estimation module in RASCAL. These include providing very rapid, although relatively rough estimation of the amount of fission products released to atmosphere, as well as providing necessary input for atmospheric dispersion simulation of fission products and people dose calculation. Based on an investigation of studies in nuclear emergency response and design features of AP1000 NPP, some adjustments are made during development of RASTES. These

Fig. 10. Cesium release under different conditions.

adjustments enable the system to be more flexible and realistic in estimating atmospheric source term, and more applicable to AP1000 NPP. These adjustments include: 1. In RASCAL, emergency responders need to input the assessment of fuel damage in RPV to complete source term estimation based on their own judgments. While in RASTES, three reactor core damage assessment methods are integrated into atmospheric source term estimation. This strategy provides more alternative choices for this input, and makes RASTES more flexible to be used in nuclear emergency response. 2. In both RASCAL and nuclear emergency guiding reports, such as RTM-96, aerosol removal factors in natural process and spray process are very rough. In RASTES, adjustments of these factors are made based on an investigation of aerosol removal factors in nuclear emergency response. These factors have a more detailed description of aerosol removal in reactor containment, and have been validated by comparison with CONTAIN code or Hanford experiment result. These adjustments make atmospheric source term estimation results more realistic. In addition, passive containment cooling in AP1000 NPP is considered in RASTES, and this makes the system more applicable to AP1000 NPP. 3. Containment radiation levels for specified fuel damage, which are used in reactor core damage assessment, are recalculated, being consistent with adjustments of aerosol removal factors and the specific location of radiation monitor in AP1000 NPP containment. This also enables RASTES to produce more realistic estimation results and be more applicable to AP1000 NPP. In addition to these adjustments, RASTES is used to compare the effect of different aerosol removal processes on atmospheric source term in this paper. The comparison shows that PCC process is able to reduce total release of iodine and aerosol by about 50% compared with natural process, although PCS is mainly used to ensure the

Table 10 Total release of noble gases, iodine and cesium under different conditions. Fission products removal process

Released activity (1014 Bq) Fig. 9. Iodine release under different conditions.

Noble gases Iodine Cesium

Natural process

PCC process

Spray process

2274.26 756.39 64.39

2232.54 408.68 36.95

2188.13 47.68 6.40

Y. Zhao et al. / Progress in Nuclear Energy 81 (2015) 264e275

integration of containment. If spray system is available in the early stage of fission product release process, it is still demonstrated to be the most effective way to reduce aerosol release, despite its nonsafety-related design. Investigation of aerosol removal in containment shows that there are uncertainties in both natural process and spray process. In the future, RASTES may be improved by introducing uncertainty distributions, such as 10th and 90th percentile, in aerosol removal factors, by which atmospheric source terms in different uncertain levels can be calculated. Besides, in addition to LBLOCA scenario, some other important and typical accident scenarios will be integrated into RASTES to enable it to be more flexible in emergency response.

Acknowledgment This work was supported by the National Research and engineering application of technical energy demonstration projects of China (Grant No. NY20120202).

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