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Current status of spent fuels and the development of computer programs for the PWR spent fuel management in Korea
Progress in Nuclear Energy 53 (2011) 290e297
Contents lists available at ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Current status of spent fuels and the development of computer programs for the PWR spent fuel management in Korea Heui-Joo Choi*, Dongkeun Cho, Donghak Kook, Jongwon Choi Korea Atomic Energy Research Institute, 1045 Daeduk-Daero, Yuseong, Daejeon, Republic of Korea
a r t i c l e i n f o
a b s t r a c t
Article history: Received 16 July 2010 Accepted 28 December 2010
Nuclear power has supplied the national electric power demand for three decades in the Republic of Korea, which has resulted in the accumulation of a large amount of spent fuels. The government has a policy on the temporary storage of these at nuclear power plants at present. In order to establish a proper policy for spent fuel management in the near future, the characteristics and amount of spent fuels should be figured out properly. In this paper, the current status of spent fuels in the Republic of Korea is outlined focusing on the major characteristics of spent fuels such as initial enrichment and discharge burnup. According to the current trend, the average burnup of PWR spent fuels will reach 55 GWd/MtU by the middle of 2010s. Three different kinds of computer programs were developed to supply crucial data regarding spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The projection was verified with real spent fuel data. The second Database program was prepared for the analysis of statistics regarding PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of the geological disposal density and cooling time of PWR spent fuels. Disposal area could be reduced by 50% through a proper analysis of the cooling time of PWR spent fuels. Finally, A-SOURCE program was developed to easily calculate source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selected PWR spent fuel assemblies and could calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated. Ó 2011 Elsevier Ltd. All rights reserved.
Keywords: Spent fuel Burnup Database Source-term Disposal
1. Introduction The South Korean government has maintained a consistent national policy for a stable energy supply by fostering nuclear power industries due to the insufficient energy resources in the country. Nuclear power has reached approximately 40% of the total domestic electricity generated. Since the commencement of the first commercial operation of Kori unit 1 in April 1978, 20 units of Nuclear Power Plants (NPPs) were commercially operating as of December 2009. Four units out of the 20 operating NPPs are CANDU reactors (PHWRs) at the Wolsong site. The 16 PWR units are located at the Kori, Yonggwang, and Ulchin sites. The spent fuels generated from these NPPs are stored in spent fuel storage pools at the reactors or at the on-site dry storage facility for PHWR spent fuels. Low and Intermediate-Level Wastes (LILW) generated from the NPPs are stored at the on-site radioactive waste storage facilities and will be
* Corresponding author. E-mail address: [email protected] (H.-J. Choi). 0149-1970/$ e see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2010.12.006
disposed of at the LILW disposal site which is under construction close to the Wolsong site. As the amount of spent fuels substantially increased, systematic approaches for the management of spent fuels were required by the government and utilities. For example, at the stage of the conceptual design of a geological disposal system or pyro-processing system, it was necessary to define the characteristics of reference spent fuels based on the current statistics and future projection of spent fuels (Cho et al., 2008a,b). Also, an accurate projection of spent fuel arising in the Republic of Korea was frequently requested in order to estimate a radioactive waste management fund especially for spent fuels. In the event that pyro-processing of PWR spent fuels is developed in the future, totally different kinds of high-level wastes are expected. This situation makes it necessary to develop a sourceterm estimation program. Under such circumstances three kinds of computer programs are to be developed for such purposes. The main purpose of this paper is to outline the current status of spent fuels in the Republic of Korea. The major characteristics of PWR spent fuels such as assembly design, initial enrichment, and burnup are introduced. Also, three different kinds of computer programs
H.-J. Choi et al. / Progress in Nuclear Energy 53 (2011) 290e297 Table 1 Spent fuels in NPPs (as of Dec. 2009).
291
500
60 55
Location
PWR
Kori Yonggwang Ulchin Wolsong Total
PHWR
Number
4 6 6 4 20
Storage capacity (MTU)
Cumulative amount (MTU)
Saturation year
2253 2686 2328 5980 13,247
1762 1704 1401 5894 10,083
2016 2016 2017 2009 e
were developed to supply crucial data on spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The second database program was prepared for several statistics regarding PWR spent fuels. Finally, the third one was for calculating the source-terms of decay heat and radioactivity from the pyro-processing of PWR spent fuels. The usefulness of the programs was illustrated with a test case for the pyro-processing of several PWR spent fuel assemblies. 2. Current status of spent fuels 2.1. Amount of spent fuels In August 1998, the criteria for radioactive waste classification were amended with a view to emphasizing the disposal safety of radioactive wastes as above, in the IAEA revised classification system of IAEA Safety Series No.111-G-1.1 (1994). The Korean Atomic Energy Act (AEA) defines ‘Radioactive Waste’ as radioactive materials or materials contaminated with radioactive materials which are the object of disposal, including spent fuel. The Enforcement Decree of the AEA defines High-Level Waste (HLW) as radioactive waste with radioactivity and heat generation over the limit value specified by the Ministry of Education, Science, and Technology (MEST). All radioactive wastes other than HLW and exempt radioactive wastes belong to Low and Intermediate-Level Waste in accordance with the AEA definition. For HLW, the limit value on radioactivity and heat generation is specified in the MEST Notice (Criteria for Radiation Protection, etc.) as follows: Radioactivity: 4000 Bq/g for a-emitting nuclide having a half life longer than 20 years Heat Generation: 2 kW/m3 Spent fuel generation rates are 13 and 16 ton/year for Kori unit 1 with 590 MWe and Kori unit 2 with 680 MWe, respectively, 19
400
Production Amount [tU]
Reactor type
50 45
300
Average Burnup
40 35
200
30 25
100
Production Amount
20 15
Burnup of Spent Nuclear Fuel [GWd/tU]
Maximum Burnup
0 80
82
84
86
88
90
92
94
96
98
00
02
04
06
08
Time [year] Fig. 2. Annual production amount and burnup trend of spent fuel.
ton/year for the 14 1000 MW PWR plants, and 95 ton/year for the remaining 4 PHWR plants. The amounts of spent fuels at the NPP sites are summarized in Table 1. As shown in Table 1, the existing storage capacity and cumulative amount of spent fuels as of December 2009 were 13,247 MTU and 10,761 MTU, respectively (Korea Institute of Nuclear Safety, 2010). The spent fuels generated from NPPs are being stored in the spent fuel storage facility at each unit. The storage capacity for spent fuels was expanded as a consequence of the delayed schedule for the construction of an AwayFrom-Reactor (AFR) interim storage facility in accordance with the 249th meeting and the 253rd meeting of the Korean Atomic Energy Committee. As Kori units 1 and 2 encountered a shortage of the spent fuel storage capacity, the spent fuels which were in excess of the storage capacity were transshipped to the neighboring spent fuel pools of Kori units 3 and 4. Besides the transshipment, Kori and Yonggwang units expanded their storage capacities by installing high-density storage racks. Ulchin also expanded its storage capacity 1.5e2 times by installing high-density storage racks. Full re-racking projects using Borated Aluminum for Ulchin units 3 and 4 were initiated in October, 2005 and completed in August, 2007.At the Wolsong site, 400 concrete silos were built for the storage of CANDU spent fuels. In order to expand its storage capacity, MACSTOR/KN-400 CANDU spent fuel dry storage facility was built in 2009. One storage module of this system can accommodate 24,000 CANDU spent fuel bundles, which are almost equivalent to the amount of spent fuels generated from four CANDU reactors for one year.
Fig. 1. History of various fuel supplies in the Republic of Korea (YGN: YoungGwang, UCN: Ulchin, S-UCN: new-Ulchin, S-KR: new-Kori).
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Fig. 3. Main screen of spent fuel accumulation projection program.
Whether to directly dispose of or recycle spent fuels has not been decided yet in the Republic of Korea. Since it was stipulated at the 253rd meeting of the Korean Atomic Energy Committee that a national policy for spent fuel management was to be decided later in consideration of domestic and international technology development, spent fuels are being stored at reactor sites under the utilities’ responsibility for the time being. 2.2. Spent fuel species Nuclear power plants in the Republic of Korea have used various kinds of fuels which are specific in fuel types, fuel materials, and fuel suppliers. They are very various taking into account important
factors such as cooling time, discharge burnup, decay heat, radioactive source concentration, initial enrichment, and so on. There are three fuel types of PWR spent fuels in the Republic of Korea: 14 ⅹ 14, 16 ⅹ 16, and 17 ⅹ 17. In the early 1990s, KOFA (Korean Fuel Assembly), which was co-developed with KWU (Kraftwerk Union), was supplied to the Westinghouse type reactors. For 14 ⅹ 14, Kori unit 1 used Westinghouse 14 ⅹ 14 OFA (Optimized Fuel Assembly). For 16 ⅹ 16, there are two species: one is STD (Standard Fuel Assembly), which is the Westinghouse standard fuel assembly used for Kori unit 2, and the other is KSFA which is the Korea standard fuel assembly used for Ulchin units 3e6 and Yonggwang units 3e6. For 17 ⅹ 17, this fuel type is used at Westinghouse reactors (Kori units 3e4, and Ulchin units 1e2, and Yonggwang units 1e2) with
Fig. 4. Result of spent fuel projection at each reactor.
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293
Table 2 Comparison of projection and real accumulation. Nuclear power plant site
Spent fuel projections [MtU]
Spent fuel accumulation [MtU]
Differences (%)
Kori Ulchin YoungGwang Wolsong
1619.9 1156.7 1432.2 5039
1623.02 1213.43 1491.08 5092.33
0.1 4.6 3.9 1.0
Total
9247.8
9419.86
1.8
V5H. V5H (Vantage 5H fuel assembly) was developed to reach 48 GWd/MtU discharge burnup, but the new fuel ACE7 has replaced the V5H with an average 55 GWd/MtU discharge burnup. ACE7 increases thermal power by 10% compared with V5H and has been loaded into power reactors since August, 2008. The Korea standard reactor, OPR (Optimized Power Reactor), has used Guardian 16 ⅹ 16 fuels since 2001 in order to enhance debrisfiltering efficiency. For 3-loop Westinghouse type plants, RFA (Robust Fuel Assembly), which includes key mechanical enhancements to improve structural stability and vibration characteristics, has been provided since 2003. This fuel type has been replaced with PLUS7 since 2007 which increased thermal performance by 10% and burnup to 55 GWd/MtU. Fig. 1 shows the variety of nuclear fuel species in the Republic of Korea with its supply history. Until 2006, 17 ⅹ 17 fuel type amounted to 57% of total spent fuels, and 16 ⅹ 16 fuel type around 33%. KSFA amounted to 22.6% in total. According to the projection of PWR fuels, 16 ⅹ 16 fuel type will dominate spent fuel inventory after 2020. 2.3. Enrichment and burnup trend The initial enrichment trend is analyzed to estimate the spent fuel burnup trend since the initial enrichment of nuclear fuel is closely connected with the spent fuel burnup. For Kori units 1 and 2, fuels with less than 3.8 wt% enrichment have been used since 1996. For Westinghouse type reactors (Kori unit 3e4, and Yonggwang unit 1e2, and Ulchin unit 1e2), fuels with 3.6 wt% enrichment were used in the early 1990s and V5H fuels with 4.1e4.2 wt% enrichment in the late 1990s, and V5H fuels with 4.5 wt% enrichment in early 2000. For the Korea standard power plant (OPR), fuels with 4.2 wt% enrichment were used in the late 1990s, and fuels with 4.5 wt% enrichment, and all OPRs use 4.5 wt% enrichment fuels these days. In summary, initial enrichments were less than 4.0 wt% before 1995, 4.0e4.5 wt% between 1995 and 2000, and 4.5 wt% after 2000. The enrichment trend means that nuclear power plants yield more thermal power, lengthen the operating period by reducing unplanned power plant trips, and increase the fuel burnup. In order to anticipate the future enrichment trend, several assumptions were made, namely that 3.8 wt% value was used for Kori units 1e2, and 4.5 wt% for Kori units 3e4, and all Yonggwang, Ulchin, Sin-Kori, Sin-Wolsung Units. In the case of the 1400 MWe power plant, 4.437 wt% for 68 assemblies, 4.407 wt% for 8 assemblies, and 4.59 wt% for 16 assemblies were assumed for every re-loading time.
Fig. 5. An example shows the discharge burnup distribution from PWR spent fuel database.
In early 2010, fuels with more than 4.0 wt% enrichment accounted for 55% and fuels with more than 4.5 wt% accounted for 45% of total spent fuel amount. In 2080, fuels with more than 4.0 wt% enrichment will account for 88%, and fuels with more than 4.5 wt% will account for 62% of total spent fuel amount. The discharge burnup of a spent fuel is essential for calculating source-terms of decay heat and radionuclide inventories. Most of the nuclear designs such as criticality, radiation shielding, and thermal analysis are based on the correct estimation of sourceterms. In this paper the burnup distribution of PWR spent fuels in the Republic of Korea was analyzed up to the year 2008. Fig. 2 shows PWR spent fuel amount as a bar chart, annual average spent fuel burnup as a blue dot & dash line (below), and annual maximum spent fuel burnup as a blue triangle & dash line (above). Based on the recent 10 years’ data, a maximum to average burnup ratio of 1.25 was calculated. In the middle of the 1980s, the discharge burnup was 30 GWd/MtU. As the initial enrichment increased from 3.2 wt% to 3.8 wt% gradually in the late 1980s, the discharge burnup became 37 GWd/MtU in the middle of the 1990s (Cho et al., 2007). V5H fuels with 4.1e4.2 wt% enrichment were loaded into reactors in the late 1990s, which resulted in discharge burnup values exceeding 40 GWd/MtU in early 2000. Recently, spent fuels which had 4.2e4.5 wt% enrichment were discharged with 45 GWd/MtU. The maximum discharge burnup increased continuously from 45 GWd/MtU in the late 1990s to 55 GWd/MtU in early 2000. Regression analysis between initial enrichment and discharge burnup was performed. According to this analysis, the average discharge burnup of spent fuel with 4.0 wt% initial enrichment is 41 GWd/MtU, and 46 GWd/MtU for 4.5 wt% initial enrichment spent fuel. It is expected that the average discharge burnup will be 55 GWd/MtU from 2015 because discharge burnup shows an increasing trend, spent fuels with 4.5 wt% initial will be discharged
Table 3 Structure of Database for PWR spent fuels. Class-1
Class-2
Class-3
Class-4
Class-5
Class-6
Class-7
Output list
Site name (KR, YG, UC)
Unit number (KR #1w#4, YG #1w#6, UC #1w#6)
Initial enrichment (0e5%)
Discharge burnup (0e70 GWd/tU)
Cooling time (0e50 years)
Amount (tU, assembly)
Date (loading, discharge)
Spent fuel type Fuel rod array Amount Initial enrichment Discharge burnup Cooling time
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a discharge burnup model, and an annual average discharge model (Cha et al., 2009). For example, Eq. (1) is for the discharge burnup model:
Cfk ¼ Cck þ 365
n X i¼1
Pk
Lk 1 þ Lkc 3k B k
(1)
where Cfk is the amount of spent fuel to be accumulated, Pk is the electric power of kth nuclear reactor [MWe], Lk is the capacity factor of kth nuclear reactor [%], 3k is the thermal efficiency [%], Bk spent fuel discharge burnup [MWd/tU]. Fig. 3 shows an example of the main screen of the SFAPP program. An example of the spent fuel projection was given in Fig. 4. A user can define and predict any combination of spent fuel accumulation regardless of the site or nuclear reactors. Table 2 shows the comparison result of the projection and real accumulation of PWR spent fuels. As shown in Table 2, the projection shows a good agreement with the real accumulation. Fig. 6. Annual change of discharge burnup of spent fuels at Kori site.
3.2. Spent fuel database program 4000 44000 42000
3500
Number of assemblies
3000
38000 36000
2500
34000 2000
32000 30000
1500
28000 1000
26000
Average burnup (MW d/MtU)
40000
24000
500
22000 0
20000 0~5
5~10
10~15
15~20
20~25
25~30
Cooling Time (year) Fig. 7. Average burnup and number of assembly distribution of PWR spent fuels.
from 2015, and new fuels which have a 55 GWd/MtU burnup target will have been developed and applied. 3. Development of computer programs 3.1. Spent fuel arising projection program Projections for future spent fuel accumulation have frequently been requested. A correct projection for the future accumulation of spent fuels would help establish proper planning for the management of spent fuels. Due to a wide spectrum of nuclear reactors in the Republic of Korea, it was necessary to prepare a computer program to estimate future accumulation. The computer program, SFAPP (Spent Fuel Arising Projection Program), was developed for this purpose. Three different models were used for the projection of PWR spent fuels in the SFAPP, namely, a cycle length model,
By the end of the year 2008, around 11,300 PWR spent fuel assemblies were accumulated at nuclear power plant sites. A Database program was prepared to analyze the statistics related to the characteristics of the spent fuels based on each PWR spent fuel assembly. The data classification is given in Table 3. The site name and unit number indicate the specific location and serial numbering of the nuclear power plant. Cooling time is the date from the discharge date to the user-defined date. Fuel type is the fuel assembly manufacturer’s name and Rod Array signifies the fuel rod configuration array in an assembly. That is, each assembly was specified with 18 items of data, and its data were recorded in the Database program. Fig. 5 and Fig. 6 show examples of the output from the Database. As shown in Fig. 5, the discharge burnup distribution for all the PWR spent fuels is easily obtained. If necessary, the discharge burnup distribution of spent fuels at each site can be analyzed with ease. Fig. 6 shows the discharge burnup distribution of spent fuels accumulated at the Kori site. The usefulness of the database program is illustrated using an example case. According to a previous study on the development of a geological disposal system (Lee et al., 2007), the Korean Reference disposal System for PWR spent fuels was designed on the assumption of a reference burnup of 45,000 MWd/MtU and a 40 year cooling time for a reference PWR spent fuel. Due to the large decay heat from the PWR spent fuels, the spacings between the disposal tunnels and between the disposal holes were determined to be 40 m and 6 m, respectively. However, subsequent research on PWR spent fuel cooling time showed that the disposal density, which means the amount of spent fuels per unit disposal area, could be increased by 50% when the cooling time increased from 40 years to 70 years (Lee et al., 2008). Thus, an analysis of PWR spent fuel cooling time was carried out with the Database program. As shown in Fig. 7, the average burnup and number of assemblies were calculated depending on the cooling time as of Jan. 2009. The calculation was carried out on the basis of a 5-year cooling time.
Table 4 Analysis of average burnup distribution versus cooling time with Database program. Year
0w5
6 w 10
11 w 15
16 w 20
21 w 25
26 w
Total
Number of assemblies Percentage of assemblies Average cooling time (years) Average burnup (Gd/MtU)
3751 33% 2.8 41.90
2662 24% 7.3 39.02
1979 18% 12.3 35.94
1902 17% 17.3 31.29
826 7% 21.7 24.61
156 1% 26.9 26.78
11,276 100% 9.72 36.939
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295
Fig. 8. An example of A-SOURCE input screen showing the selection of spent fuels of interest.
Ass. ID
Assembly type
KK1A01 KK1A24 KK3F15
WH-STD WH-STD WH-KOFA
Initial U loading 400 400 367
enrichment (wt.%) 1.80 3.21 3.32
discharge burnup 12250 31500 33000
downtime (days) 2325 1825 2555
PcP & Waste characterization
Temporary storage Make ORIGEN input for PcP preparation Restart Composition ftKK2F15
Checking irradiation/decay condition, make ORIGEN-S for PcP preparation
WC #1
Make Batch input & decay cal. up to PcP
WC #2 WC #3
Fig. 9. A test case for A-SOURCE calculation.
The result of the analysis is given in Table 4. As given in Table 4, the average cooling time of PWR spent fuels produced by the end of 2008 was 9.7 years as of January 1, 2009, which means more than 10 years in 2010. Even though no policy for spent fuel management has yet been fixed in the Republic of Korea, a final disposal facility for PWR spent fuels will be in operation in the late 2060s. This situation implies around 70 years of cooling time for existing PWR spent fuels. In such a case, the disposal density could be increased by 50% for the above-mentioned PWR spent fuels. Also, it is believed that a more accurate analysis of the average burnup and cooling time will be possible in the future. 3.3. Source term assessment program Generally, the source-terms such as radioactivity and decay heat of spent fuels are calculated using a well-known computer program, ORIGEN series (Gauld et al., 2006). However, given that the pyroprocessing of PWR spent fuels must be taken into account, it is necessary to develop a new computer program to calculate the source-terms of HLW from pyro-processing. The amount of HLW from pyro-processing is expected to be very small, which means several PWR spent fuel assemblies should be processed to make an HLW waste form. In such a case, it is not easy to calculate the decay heat or radioactivity of the waste form from the pyro-processing of spent fuels with the ORIGEN program. To overcome the difficulties,
an interface program called A-SOURCE was developed, which makes it possible to use ORIGEN-S with the material balance from the pyroprocessing of PWR spent fuels. The A-SOURCE program picks up data from the PWR spent fuel assembly information from the abovementioned Database and sends the data to ORIGEN-S to calculate the nuclide concentrations according to the following equation:
X X dNi dji lj Nj þ fki sk Nk F ðli þ si FÞNi ¼ dt j
(2)
k
where Ni is the number density of nuclide [m3] dji is the fraction of nuclide Nj decaying to Ni, fki is the fraction of nuclide Nk transmutated to Ni due to neutron capture,
Table 5 Characteristics of PWR spent fuels for the test case. Assembly ID
Assembly type
Initial U loading per ass.(kg)
Enrichment (wt.%)
Discharge burnup (GWd/tU)
Downtime (days)
KK1A01 KK1A24 KK3F15 KK2F15
WH-STD WH-STD WH-KOFA WH-STD
400 400 367 367
1.8 3.21 3.32 3.32
12,250 31,500 33,000 33,000
2325 1825 2555 2500
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Table 6 Transfer rate of major fission products in pyro-processing. Product #1
Elem.
Frac.
Elem.
Frac.
Elem.
Frac.
Elem.
Cs Sr I Tc
0.95 0.95 0.99 0.99
Cs Sr I Tc
0.05 0.05 0.01 0.01
U Zr
0.99 1.00
U TRU
Waste class #3
30,000
Frac.
Elem.
Frac.
25,000
0.01 0.99
TRU Others
0.01 Rest
Product #2
In Eq. (2) the neutron flux is a space-energy-averaged value, and the cross sections are a flux-weighted average. Then, A-SOURCE receives the nuclide concentrations directly calculated by ORIGENS and calculates the radionuclide concentration in the waste stream based on the material balance of the pyro-processing. In this way, the radionuclide concentration or decay heat from the combination of several PWR assemblies can be calculated with ease by using the A-SOURCE program. Fig. 8 shows an example of the A-SOURCE program application. The A-SOURCE program identifies the PWR spent fuel assemblies which meet the conditions set by the user. A simple test case is illustrated to demonstrate the application of the A-SOURCE program. As shown in Fig. 9, four PWR spent fuels with different histories were pyro-processed, and the focus is on the radionuclide concentration in the waste forms from the pyro-processing of the spent fuels. The characteristics of the selected spent fuels such as initial enrichment and burnup are given in Table 5. Two assemblies were from Kori unit 1, and the others were from Kori units 2 and 3. Radionuclide transfer rates in the pyro-processing are given in Table 6. The decay heat from the major radionuclides in Waste Class-1 was calculated using the A-SOURCE program, and the results are given in Fig. 10. In the A-SOURCE program, burnup calculations are carried out on the basis of one ton of initial uranium loading. However, source-terms like nuclide concentration, radioactivity, decay heat, and hazard index are normalized to the initial uranium loading. Fig. 11 shows the amount of 235U in the spent fuels before and after pyro-processing. In this illustration, a mixing ratio of 0.9, 0.9, 0.9, and 0.8 in pyro-processing was applied to KK1A01, KK1A24, KK3F15, and KK2F15, respectively. As shown in the example calculation, the source-terms such as decay heat and radionuclide
1200
Decay Heat(Watts/batch)
1000 total 800
600
Y Ba
200 Cs Cs Tc/ Cs/ I
0 0
2000
Amoun t of
F is average neutron flux [m2 s1], s is neutron capture cross section including fission [m2], l is a decay constant [s1].
400
U (g/Basis)
Waste class #2
235
Waste class #1
35,000
Pyroprocessing
20,000
15,000
10,000
5,000
Recovered
U
KK1A01 KK1A24
KK3F15 Decayed
KK2F15 0 -3500 -3000 -2500 -2000 -1500 -1000
-500
0
500
1000
1500
Irradiation/Decay Time Prior to Pyroprocessing(days) Fig. 11. Amount of
235
U before pyro-processing and after pyro-processing.
concentration can be calculated easily with the A-SOURCE program for any combination of PWR spent fuel assemblies. 4. Conclusions As the amount of spent fuels has grown at nuclear power plant sites, local communities have urged the government to establish a policy for them. An appropriate policy should be based on correct information regarding the spent fuel status. This paper outlines the current status of PWR spent fuels in the Republic of Korea. For the purpose of proper management of spent fuels, the major characteristics such as initial enrichment, burnup, and discharge date of spent fuels should be figured out properly. In this paper, spent fuel species, initial enrichment, and discharge burnup trends were analyzed for all the PWR spent fuel assemblies generated by the end of 2008. As expected, the initial enrichment and the discharge burnup have increased continuously. The average burnup will reach 55 GWd/MtU by 2015. The maximum to average burnup ratio of the spent fuels produced for the recent ten years is 1.25. Three kinds of computer programs were developed for the purpose of supporting spent fuel management. To project the future accumulation of PWR spent fuels, the SFAPP program was developed based on three different models, and the projection result was verified with real spent fuel data. A Database program was prepared to analyze the statistics of PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of geological disposal density and cooling time of PWR spent fuels. The disposal area can be reduced by 50% through proper analysis of the cooling time of PWR spent fuels. Given that the pyro-processing of PWR spent fuels could reduce the amount of high-level waste remarkably, the calculation of source-terms might require cumbersome tasks. A-SOURCE program was developed to easily calculate the source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selects PWR spent fuel assemblies and can calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated.
Sr
Acknowledgment 4000
6000
8000
Decay Time(days) Fig. 10. Decay heat in Waste Class-1: results from test case.
10000
This study was performed under the long-term nuclear research and development program sponsored by the Ministry of Education, Science and Technology.
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References Cha, Jeong-Hun, Choi, Heui-Joo, Lee, Jong-Youl, Choi, Jongwon, 2009. Development of a computer programs for an analysis of the logistics and transportation costs of the PWR spent fuels in Korea. Journal of Radiation Protection 34, 1e7. Cho, Dong-Keun, Lee, Yang, Lee, Jong-Youl, Choi, Jongwon, 2007. Characteristics of a Geological Disposal System for the increasing burn-up of spent nuclear fuel in Korea. Journal of Nuclear Science and Technology 44, 1306e1316. Cho, Dong-Keun, Lee, Seung Woo, Cha, Jeong Hun, Choi, Jongwon, Lee, Yang, Choi, Heui-Joo, 2008a. Current status and characterization of CANDU spent fuel for Geological Disposal System design. Journal of the Korean Radioactive Waste Society 6, 155e162.
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Cho, Dong-Keun, Yoon, Seok-Kyun, Choi, Heui-Joo, Choi, Jongwon, Ko, Won Il, 2008b. Reference spent nuclear fuel for pyroprocessing facility design. Journal of the Korean Radioactive Waste Society 6, 225e232. Gauld, I.C., Bowman, S.M., Horwedel, J.E., 2006. ORIGEN-ARP: Automatic Rapid Processing for Spent Fuel Depletion, Decay, and Source Term Analysis. ORNL/ TM-2005/39. ORNL, USA. KINS., 2010. 2010 White Paper on Nuclear Safety. Korea Institute of Nuclear Safety, South Korea. p.236. Lee, Jongyoul, Cho, Dongkeun, Choi, Heuijoo, Choi, Jongwon, 2007. Concept of a Korean Reference Disposal System for spent fuels. Journal of Nuclear Science and Technology 44, 1565e1573. Lee, Jongyoul, Cho, Dongkeun, Choi, Heuijoo, Choi, Jongwon, Lee, Yang, 2008. Analysis of the spent fuel cooling time for a Deep geological disposal. Journal of the Korean Radioactive Waste Society 6, 65e72.
Contents lists available at ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Current status of spent fuels and the development of computer programs for the PWR spent fuel management in Korea Heui-Joo Choi*, Dongkeun Cho, Donghak Kook, Jongwon Choi Korea Atomic Energy Research Institute, 1045 Daeduk-Daero, Yuseong, Daejeon, Republic of Korea
a r t i c l e i n f o
a b s t r a c t
Article history: Received 16 July 2010 Accepted 28 December 2010
Nuclear power has supplied the national electric power demand for three decades in the Republic of Korea, which has resulted in the accumulation of a large amount of spent fuels. The government has a policy on the temporary storage of these at nuclear power plants at present. In order to establish a proper policy for spent fuel management in the near future, the characteristics and amount of spent fuels should be figured out properly. In this paper, the current status of spent fuels in the Republic of Korea is outlined focusing on the major characteristics of spent fuels such as initial enrichment and discharge burnup. According to the current trend, the average burnup of PWR spent fuels will reach 55 GWd/MtU by the middle of 2010s. Three different kinds of computer programs were developed to supply crucial data regarding spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The projection was verified with real spent fuel data. The second Database program was prepared for the analysis of statistics regarding PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of the geological disposal density and cooling time of PWR spent fuels. Disposal area could be reduced by 50% through a proper analysis of the cooling time of PWR spent fuels. Finally, A-SOURCE program was developed to easily calculate source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selected PWR spent fuel assemblies and could calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated. Ó 2011 Elsevier Ltd. All rights reserved.
Keywords: Spent fuel Burnup Database Source-term Disposal
1. Introduction The South Korean government has maintained a consistent national policy for a stable energy supply by fostering nuclear power industries due to the insufficient energy resources in the country. Nuclear power has reached approximately 40% of the total domestic electricity generated. Since the commencement of the first commercial operation of Kori unit 1 in April 1978, 20 units of Nuclear Power Plants (NPPs) were commercially operating as of December 2009. Four units out of the 20 operating NPPs are CANDU reactors (PHWRs) at the Wolsong site. The 16 PWR units are located at the Kori, Yonggwang, and Ulchin sites. The spent fuels generated from these NPPs are stored in spent fuel storage pools at the reactors or at the on-site dry storage facility for PHWR spent fuels. Low and Intermediate-Level Wastes (LILW) generated from the NPPs are stored at the on-site radioactive waste storage facilities and will be
* Corresponding author. E-mail address: [email protected] (H.-J. Choi). 0149-1970/$ e see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2010.12.006
disposed of at the LILW disposal site which is under construction close to the Wolsong site. As the amount of spent fuels substantially increased, systematic approaches for the management of spent fuels were required by the government and utilities. For example, at the stage of the conceptual design of a geological disposal system or pyro-processing system, it was necessary to define the characteristics of reference spent fuels based on the current statistics and future projection of spent fuels (Cho et al., 2008a,b). Also, an accurate projection of spent fuel arising in the Republic of Korea was frequently requested in order to estimate a radioactive waste management fund especially for spent fuels. In the event that pyro-processing of PWR spent fuels is developed in the future, totally different kinds of high-level wastes are expected. This situation makes it necessary to develop a sourceterm estimation program. Under such circumstances three kinds of computer programs are to be developed for such purposes. The main purpose of this paper is to outline the current status of spent fuels in the Republic of Korea. The major characteristics of PWR spent fuels such as assembly design, initial enrichment, and burnup are introduced. Also, three different kinds of computer programs
H.-J. Choi et al. / Progress in Nuclear Energy 53 (2011) 290e297 Table 1 Spent fuels in NPPs (as of Dec. 2009).
291
500
60 55
Location
PWR
Kori Yonggwang Ulchin Wolsong Total
PHWR
Number
4 6 6 4 20
Storage capacity (MTU)
Cumulative amount (MTU)
Saturation year
2253 2686 2328 5980 13,247
1762 1704 1401 5894 10,083
2016 2016 2017 2009 e
were developed to supply crucial data on spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The second database program was prepared for several statistics regarding PWR spent fuels. Finally, the third one was for calculating the source-terms of decay heat and radioactivity from the pyro-processing of PWR spent fuels. The usefulness of the programs was illustrated with a test case for the pyro-processing of several PWR spent fuel assemblies. 2. Current status of spent fuels 2.1. Amount of spent fuels In August 1998, the criteria for radioactive waste classification were amended with a view to emphasizing the disposal safety of radioactive wastes as above, in the IAEA revised classification system of IAEA Safety Series No.111-G-1.1 (1994). The Korean Atomic Energy Act (AEA) defines ‘Radioactive Waste’ as radioactive materials or materials contaminated with radioactive materials which are the object of disposal, including spent fuel. The Enforcement Decree of the AEA defines High-Level Waste (HLW) as radioactive waste with radioactivity and heat generation over the limit value specified by the Ministry of Education, Science, and Technology (MEST). All radioactive wastes other than HLW and exempt radioactive wastes belong to Low and Intermediate-Level Waste in accordance with the AEA definition. For HLW, the limit value on radioactivity and heat generation is specified in the MEST Notice (Criteria for Radiation Protection, etc.) as follows: Radioactivity: 4000 Bq/g for a-emitting nuclide having a half life longer than 20 years Heat Generation: 2 kW/m3 Spent fuel generation rates are 13 and 16 ton/year for Kori unit 1 with 590 MWe and Kori unit 2 with 680 MWe, respectively, 19
400
Production Amount [tU]
Reactor type
50 45
300
Average Burnup
40 35
200
30 25
100
Production Amount
20 15
Burnup of Spent Nuclear Fuel [GWd/tU]
Maximum Burnup
0 80
82
84
86
88
90
92
94
96
98
00
02
04
06
08
Time [year] Fig. 2. Annual production amount and burnup trend of spent fuel.
ton/year for the 14 1000 MW PWR plants, and 95 ton/year for the remaining 4 PHWR plants. The amounts of spent fuels at the NPP sites are summarized in Table 1. As shown in Table 1, the existing storage capacity and cumulative amount of spent fuels as of December 2009 were 13,247 MTU and 10,761 MTU, respectively (Korea Institute of Nuclear Safety, 2010). The spent fuels generated from NPPs are being stored in the spent fuel storage facility at each unit. The storage capacity for spent fuels was expanded as a consequence of the delayed schedule for the construction of an AwayFrom-Reactor (AFR) interim storage facility in accordance with the 249th meeting and the 253rd meeting of the Korean Atomic Energy Committee. As Kori units 1 and 2 encountered a shortage of the spent fuel storage capacity, the spent fuels which were in excess of the storage capacity were transshipped to the neighboring spent fuel pools of Kori units 3 and 4. Besides the transshipment, Kori and Yonggwang units expanded their storage capacities by installing high-density storage racks. Ulchin also expanded its storage capacity 1.5e2 times by installing high-density storage racks. Full re-racking projects using Borated Aluminum for Ulchin units 3 and 4 were initiated in October, 2005 and completed in August, 2007.At the Wolsong site, 400 concrete silos were built for the storage of CANDU spent fuels. In order to expand its storage capacity, MACSTOR/KN-400 CANDU spent fuel dry storage facility was built in 2009. One storage module of this system can accommodate 24,000 CANDU spent fuel bundles, which are almost equivalent to the amount of spent fuels generated from four CANDU reactors for one year.
Fig. 1. History of various fuel supplies in the Republic of Korea (YGN: YoungGwang, UCN: Ulchin, S-UCN: new-Ulchin, S-KR: new-Kori).
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Fig. 3. Main screen of spent fuel accumulation projection program.
Whether to directly dispose of or recycle spent fuels has not been decided yet in the Republic of Korea. Since it was stipulated at the 253rd meeting of the Korean Atomic Energy Committee that a national policy for spent fuel management was to be decided later in consideration of domestic and international technology development, spent fuels are being stored at reactor sites under the utilities’ responsibility for the time being. 2.2. Spent fuel species Nuclear power plants in the Republic of Korea have used various kinds of fuels which are specific in fuel types, fuel materials, and fuel suppliers. They are very various taking into account important
factors such as cooling time, discharge burnup, decay heat, radioactive source concentration, initial enrichment, and so on. There are three fuel types of PWR spent fuels in the Republic of Korea: 14 ⅹ 14, 16 ⅹ 16, and 17 ⅹ 17. In the early 1990s, KOFA (Korean Fuel Assembly), which was co-developed with KWU (Kraftwerk Union), was supplied to the Westinghouse type reactors. For 14 ⅹ 14, Kori unit 1 used Westinghouse 14 ⅹ 14 OFA (Optimized Fuel Assembly). For 16 ⅹ 16, there are two species: one is STD (Standard Fuel Assembly), which is the Westinghouse standard fuel assembly used for Kori unit 2, and the other is KSFA which is the Korea standard fuel assembly used for Ulchin units 3e6 and Yonggwang units 3e6. For 17 ⅹ 17, this fuel type is used at Westinghouse reactors (Kori units 3e4, and Ulchin units 1e2, and Yonggwang units 1e2) with
Fig. 4. Result of spent fuel projection at each reactor.
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293
Table 2 Comparison of projection and real accumulation. Nuclear power plant site
Spent fuel projections [MtU]
Spent fuel accumulation [MtU]
Differences (%)
Kori Ulchin YoungGwang Wolsong
1619.9 1156.7 1432.2 5039
1623.02 1213.43 1491.08 5092.33
0.1 4.6 3.9 1.0
Total
9247.8
9419.86
1.8
V5H. V5H (Vantage 5H fuel assembly) was developed to reach 48 GWd/MtU discharge burnup, but the new fuel ACE7 has replaced the V5H with an average 55 GWd/MtU discharge burnup. ACE7 increases thermal power by 10% compared with V5H and has been loaded into power reactors since August, 2008. The Korea standard reactor, OPR (Optimized Power Reactor), has used Guardian 16 ⅹ 16 fuels since 2001 in order to enhance debrisfiltering efficiency. For 3-loop Westinghouse type plants, RFA (Robust Fuel Assembly), which includes key mechanical enhancements to improve structural stability and vibration characteristics, has been provided since 2003. This fuel type has been replaced with PLUS7 since 2007 which increased thermal performance by 10% and burnup to 55 GWd/MtU. Fig. 1 shows the variety of nuclear fuel species in the Republic of Korea with its supply history. Until 2006, 17 ⅹ 17 fuel type amounted to 57% of total spent fuels, and 16 ⅹ 16 fuel type around 33%. KSFA amounted to 22.6% in total. According to the projection of PWR fuels, 16 ⅹ 16 fuel type will dominate spent fuel inventory after 2020. 2.3. Enrichment and burnup trend The initial enrichment trend is analyzed to estimate the spent fuel burnup trend since the initial enrichment of nuclear fuel is closely connected with the spent fuel burnup. For Kori units 1 and 2, fuels with less than 3.8 wt% enrichment have been used since 1996. For Westinghouse type reactors (Kori unit 3e4, and Yonggwang unit 1e2, and Ulchin unit 1e2), fuels with 3.6 wt% enrichment were used in the early 1990s and V5H fuels with 4.1e4.2 wt% enrichment in the late 1990s, and V5H fuels with 4.5 wt% enrichment in early 2000. For the Korea standard power plant (OPR), fuels with 4.2 wt% enrichment were used in the late 1990s, and fuels with 4.5 wt% enrichment, and all OPRs use 4.5 wt% enrichment fuels these days. In summary, initial enrichments were less than 4.0 wt% before 1995, 4.0e4.5 wt% between 1995 and 2000, and 4.5 wt% after 2000. The enrichment trend means that nuclear power plants yield more thermal power, lengthen the operating period by reducing unplanned power plant trips, and increase the fuel burnup. In order to anticipate the future enrichment trend, several assumptions were made, namely that 3.8 wt% value was used for Kori units 1e2, and 4.5 wt% for Kori units 3e4, and all Yonggwang, Ulchin, Sin-Kori, Sin-Wolsung Units. In the case of the 1400 MWe power plant, 4.437 wt% for 68 assemblies, 4.407 wt% for 8 assemblies, and 4.59 wt% for 16 assemblies were assumed for every re-loading time.
Fig. 5. An example shows the discharge burnup distribution from PWR spent fuel database.
In early 2010, fuels with more than 4.0 wt% enrichment accounted for 55% and fuels with more than 4.5 wt% accounted for 45% of total spent fuel amount. In 2080, fuels with more than 4.0 wt% enrichment will account for 88%, and fuels with more than 4.5 wt% will account for 62% of total spent fuel amount. The discharge burnup of a spent fuel is essential for calculating source-terms of decay heat and radionuclide inventories. Most of the nuclear designs such as criticality, radiation shielding, and thermal analysis are based on the correct estimation of sourceterms. In this paper the burnup distribution of PWR spent fuels in the Republic of Korea was analyzed up to the year 2008. Fig. 2 shows PWR spent fuel amount as a bar chart, annual average spent fuel burnup as a blue dot & dash line (below), and annual maximum spent fuel burnup as a blue triangle & dash line (above). Based on the recent 10 years’ data, a maximum to average burnup ratio of 1.25 was calculated. In the middle of the 1980s, the discharge burnup was 30 GWd/MtU. As the initial enrichment increased from 3.2 wt% to 3.8 wt% gradually in the late 1980s, the discharge burnup became 37 GWd/MtU in the middle of the 1990s (Cho et al., 2007). V5H fuels with 4.1e4.2 wt% enrichment were loaded into reactors in the late 1990s, which resulted in discharge burnup values exceeding 40 GWd/MtU in early 2000. Recently, spent fuels which had 4.2e4.5 wt% enrichment were discharged with 45 GWd/MtU. The maximum discharge burnup increased continuously from 45 GWd/MtU in the late 1990s to 55 GWd/MtU in early 2000. Regression analysis between initial enrichment and discharge burnup was performed. According to this analysis, the average discharge burnup of spent fuel with 4.0 wt% initial enrichment is 41 GWd/MtU, and 46 GWd/MtU for 4.5 wt% initial enrichment spent fuel. It is expected that the average discharge burnup will be 55 GWd/MtU from 2015 because discharge burnup shows an increasing trend, spent fuels with 4.5 wt% initial will be discharged
Table 3 Structure of Database for PWR spent fuels. Class-1
Class-2
Class-3
Class-4
Class-5
Class-6
Class-7
Output list
Site name (KR, YG, UC)
Unit number (KR #1w#4, YG #1w#6, UC #1w#6)
Initial enrichment (0e5%)
Discharge burnup (0e70 GWd/tU)
Cooling time (0e50 years)
Amount (tU, assembly)
Date (loading, discharge)
Spent fuel type Fuel rod array Amount Initial enrichment Discharge burnup Cooling time
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a discharge burnup model, and an annual average discharge model (Cha et al., 2009). For example, Eq. (1) is for the discharge burnup model:
Cfk ¼ Cck þ 365
n X i¼1
Pk
Lk 1 þ Lkc 3k B k
(1)
where Cfk is the amount of spent fuel to be accumulated, Pk is the electric power of kth nuclear reactor [MWe], Lk is the capacity factor of kth nuclear reactor [%], 3k is the thermal efficiency [%], Bk spent fuel discharge burnup [MWd/tU]. Fig. 3 shows an example of the main screen of the SFAPP program. An example of the spent fuel projection was given in Fig. 4. A user can define and predict any combination of spent fuel accumulation regardless of the site or nuclear reactors. Table 2 shows the comparison result of the projection and real accumulation of PWR spent fuels. As shown in Table 2, the projection shows a good agreement with the real accumulation. Fig. 6. Annual change of discharge burnup of spent fuels at Kori site.
3.2. Spent fuel database program 4000 44000 42000
3500
Number of assemblies
3000
38000 36000
2500
34000 2000
32000 30000
1500
28000 1000
26000
Average burnup (MW d/MtU)
40000
24000
500
22000 0
20000 0~5
5~10
10~15
15~20
20~25
25~30
Cooling Time (year) Fig. 7. Average burnup and number of assembly distribution of PWR spent fuels.
from 2015, and new fuels which have a 55 GWd/MtU burnup target will have been developed and applied. 3. Development of computer programs 3.1. Spent fuel arising projection program Projections for future spent fuel accumulation have frequently been requested. A correct projection for the future accumulation of spent fuels would help establish proper planning for the management of spent fuels. Due to a wide spectrum of nuclear reactors in the Republic of Korea, it was necessary to prepare a computer program to estimate future accumulation. The computer program, SFAPP (Spent Fuel Arising Projection Program), was developed for this purpose. Three different models were used for the projection of PWR spent fuels in the SFAPP, namely, a cycle length model,
By the end of the year 2008, around 11,300 PWR spent fuel assemblies were accumulated at nuclear power plant sites. A Database program was prepared to analyze the statistics related to the characteristics of the spent fuels based on each PWR spent fuel assembly. The data classification is given in Table 3. The site name and unit number indicate the specific location and serial numbering of the nuclear power plant. Cooling time is the date from the discharge date to the user-defined date. Fuel type is the fuel assembly manufacturer’s name and Rod Array signifies the fuel rod configuration array in an assembly. That is, each assembly was specified with 18 items of data, and its data were recorded in the Database program. Fig. 5 and Fig. 6 show examples of the output from the Database. As shown in Fig. 5, the discharge burnup distribution for all the PWR spent fuels is easily obtained. If necessary, the discharge burnup distribution of spent fuels at each site can be analyzed with ease. Fig. 6 shows the discharge burnup distribution of spent fuels accumulated at the Kori site. The usefulness of the database program is illustrated using an example case. According to a previous study on the development of a geological disposal system (Lee et al., 2007), the Korean Reference disposal System for PWR spent fuels was designed on the assumption of a reference burnup of 45,000 MWd/MtU and a 40 year cooling time for a reference PWR spent fuel. Due to the large decay heat from the PWR spent fuels, the spacings between the disposal tunnels and between the disposal holes were determined to be 40 m and 6 m, respectively. However, subsequent research on PWR spent fuel cooling time showed that the disposal density, which means the amount of spent fuels per unit disposal area, could be increased by 50% when the cooling time increased from 40 years to 70 years (Lee et al., 2008). Thus, an analysis of PWR spent fuel cooling time was carried out with the Database program. As shown in Fig. 7, the average burnup and number of assemblies were calculated depending on the cooling time as of Jan. 2009. The calculation was carried out on the basis of a 5-year cooling time.
Table 4 Analysis of average burnup distribution versus cooling time with Database program. Year
0w5
6 w 10
11 w 15
16 w 20
21 w 25
26 w
Total
Number of assemblies Percentage of assemblies Average cooling time (years) Average burnup (Gd/MtU)
3751 33% 2.8 41.90
2662 24% 7.3 39.02
1979 18% 12.3 35.94
1902 17% 17.3 31.29
826 7% 21.7 24.61
156 1% 26.9 26.78
11,276 100% 9.72 36.939
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295
Fig. 8. An example of A-SOURCE input screen showing the selection of spent fuels of interest.
Ass. ID
Assembly type
KK1A01 KK1A24 KK3F15
WH-STD WH-STD WH-KOFA
Initial U loading 400 400 367
enrichment (wt.%) 1.80 3.21 3.32
discharge burnup 12250 31500 33000
downtime (days) 2325 1825 2555
PcP & Waste characterization
Temporary storage Make ORIGEN input for PcP preparation Restart Composition ftKK2F15
Checking irradiation/decay condition, make ORIGEN-S for PcP preparation
WC #1
Make Batch input & decay cal. up to PcP
WC #2 WC #3
Fig. 9. A test case for A-SOURCE calculation.
The result of the analysis is given in Table 4. As given in Table 4, the average cooling time of PWR spent fuels produced by the end of 2008 was 9.7 years as of January 1, 2009, which means more than 10 years in 2010. Even though no policy for spent fuel management has yet been fixed in the Republic of Korea, a final disposal facility for PWR spent fuels will be in operation in the late 2060s. This situation implies around 70 years of cooling time for existing PWR spent fuels. In such a case, the disposal density could be increased by 50% for the above-mentioned PWR spent fuels. Also, it is believed that a more accurate analysis of the average burnup and cooling time will be possible in the future. 3.3. Source term assessment program Generally, the source-terms such as radioactivity and decay heat of spent fuels are calculated using a well-known computer program, ORIGEN series (Gauld et al., 2006). However, given that the pyroprocessing of PWR spent fuels must be taken into account, it is necessary to develop a new computer program to calculate the source-terms of HLW from pyro-processing. The amount of HLW from pyro-processing is expected to be very small, which means several PWR spent fuel assemblies should be processed to make an HLW waste form. In such a case, it is not easy to calculate the decay heat or radioactivity of the waste form from the pyro-processing of spent fuels with the ORIGEN program. To overcome the difficulties,
an interface program called A-SOURCE was developed, which makes it possible to use ORIGEN-S with the material balance from the pyroprocessing of PWR spent fuels. The A-SOURCE program picks up data from the PWR spent fuel assembly information from the abovementioned Database and sends the data to ORIGEN-S to calculate the nuclide concentrations according to the following equation:
X X dNi dji lj Nj þ fki sk Nk F ðli þ si FÞNi ¼ dt j
(2)
k
where Ni is the number density of nuclide [m3] dji is the fraction of nuclide Nj decaying to Ni, fki is the fraction of nuclide Nk transmutated to Ni due to neutron capture,
Table 5 Characteristics of PWR spent fuels for the test case. Assembly ID
Assembly type
Initial U loading per ass.(kg)
Enrichment (wt.%)
Discharge burnup (GWd/tU)
Downtime (days)
KK1A01 KK1A24 KK3F15 KK2F15
WH-STD WH-STD WH-KOFA WH-STD
400 400 367 367
1.8 3.21 3.32 3.32
12,250 31,500 33,000 33,000
2325 1825 2555 2500
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Table 6 Transfer rate of major fission products in pyro-processing. Product #1
Elem.
Frac.
Elem.
Frac.
Elem.
Frac.
Elem.
Cs Sr I Tc
0.95 0.95 0.99 0.99
Cs Sr I Tc
0.05 0.05 0.01 0.01
U Zr
0.99 1.00
U TRU
Waste class #3
30,000
Frac.
Elem.
Frac.
25,000
0.01 0.99
TRU Others
0.01 Rest
Product #2
In Eq. (2) the neutron flux is a space-energy-averaged value, and the cross sections are a flux-weighted average. Then, A-SOURCE receives the nuclide concentrations directly calculated by ORIGENS and calculates the radionuclide concentration in the waste stream based on the material balance of the pyro-processing. In this way, the radionuclide concentration or decay heat from the combination of several PWR assemblies can be calculated with ease by using the A-SOURCE program. Fig. 8 shows an example of the A-SOURCE program application. The A-SOURCE program identifies the PWR spent fuel assemblies which meet the conditions set by the user. A simple test case is illustrated to demonstrate the application of the A-SOURCE program. As shown in Fig. 9, four PWR spent fuels with different histories were pyro-processed, and the focus is on the radionuclide concentration in the waste forms from the pyro-processing of the spent fuels. The characteristics of the selected spent fuels such as initial enrichment and burnup are given in Table 5. Two assemblies were from Kori unit 1, and the others were from Kori units 2 and 3. Radionuclide transfer rates in the pyro-processing are given in Table 6. The decay heat from the major radionuclides in Waste Class-1 was calculated using the A-SOURCE program, and the results are given in Fig. 10. In the A-SOURCE program, burnup calculations are carried out on the basis of one ton of initial uranium loading. However, source-terms like nuclide concentration, radioactivity, decay heat, and hazard index are normalized to the initial uranium loading. Fig. 11 shows the amount of 235U in the spent fuels before and after pyro-processing. In this illustration, a mixing ratio of 0.9, 0.9, 0.9, and 0.8 in pyro-processing was applied to KK1A01, KK1A24, KK3F15, and KK2F15, respectively. As shown in the example calculation, the source-terms such as decay heat and radionuclide
1200
Decay Heat(Watts/batch)
1000 total 800
600
Y Ba
200 Cs Cs Tc/ Cs/ I
0 0
2000
Amoun t of
F is average neutron flux [m2 s1], s is neutron capture cross section including fission [m2], l is a decay constant [s1].
400
U (g/Basis)
Waste class #2
235
Waste class #1
35,000
Pyroprocessing
20,000
15,000
10,000
5,000
Recovered
U
KK1A01 KK1A24
KK3F15 Decayed
KK2F15 0 -3500 -3000 -2500 -2000 -1500 -1000
-500
0
500
1000
1500
Irradiation/Decay Time Prior to Pyroprocessing(days) Fig. 11. Amount of
235
U before pyro-processing and after pyro-processing.
concentration can be calculated easily with the A-SOURCE program for any combination of PWR spent fuel assemblies. 4. Conclusions As the amount of spent fuels has grown at nuclear power plant sites, local communities have urged the government to establish a policy for them. An appropriate policy should be based on correct information regarding the spent fuel status. This paper outlines the current status of PWR spent fuels in the Republic of Korea. For the purpose of proper management of spent fuels, the major characteristics such as initial enrichment, burnup, and discharge date of spent fuels should be figured out properly. In this paper, spent fuel species, initial enrichment, and discharge burnup trends were analyzed for all the PWR spent fuel assemblies generated by the end of 2008. As expected, the initial enrichment and the discharge burnup have increased continuously. The average burnup will reach 55 GWd/MtU by 2015. The maximum to average burnup ratio of the spent fuels produced for the recent ten years is 1.25. Three kinds of computer programs were developed for the purpose of supporting spent fuel management. To project the future accumulation of PWR spent fuels, the SFAPP program was developed based on three different models, and the projection result was verified with real spent fuel data. A Database program was prepared to analyze the statistics of PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of geological disposal density and cooling time of PWR spent fuels. The disposal area can be reduced by 50% through proper analysis of the cooling time of PWR spent fuels. Given that the pyro-processing of PWR spent fuels could reduce the amount of high-level waste remarkably, the calculation of source-terms might require cumbersome tasks. A-SOURCE program was developed to easily calculate the source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selects PWR spent fuel assemblies and can calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated.
Sr
Acknowledgment 4000
6000
8000
Decay Time(days) Fig. 10. Decay heat in Waste Class-1: results from test case.
10000
This study was performed under the long-term nuclear research and development program sponsored by the Ministry of Education, Science and Technology.
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Cho, Dong-Keun, Yoon, Seok-Kyun, Choi, Heui-Joo, Choi, Jongwon, Ko, Won Il, 2008b. Reference spent nuclear fuel for pyroprocessing facility design. Journal of the Korean Radioactive Waste Society 6, 225e232. Gauld, I.C., Bowman, S.M., Horwedel, J.E., 2006. ORIGEN-ARP: Automatic Rapid Processing for Spent Fuel Depletion, Decay, and Source Term Analysis. ORNL/ TM-2005/39. ORNL, USA. KINS., 2010. 2010 White Paper on Nuclear Safety. Korea Institute of Nuclear Safety, South Korea. p.236. Lee, Jongyoul, Cho, Dongkeun, Choi, Heuijoo, Choi, Jongwon, 2007. Concept of a Korean Reference Disposal System for spent fuels. Journal of Nuclear Science and Technology 44, 1565e1573. Lee, Jongyoul, Cho, Dongkeun, Choi, Heuijoo, Choi, Jongwon, Lee, Yang, 2008. Analysis of the spent fuel cooling time for a Deep geological disposal. Journal of the Korean Radioactive Waste Society 6, 65e72.