Conference Paper SIMULATIONS OF A HEAVY-WATER REACTOR OPERATING ON SELF-SUFFICIENT EQUILIBRIUM THORIUM CYCLES COMPANY WIDE CW-123740-CONF-012 Revision 0
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Atomic Energy of Canada Limited
Énergie Atomique du Canada Limitée
Chalk River, Ontario Canada K0J 1J0
Chalk River (Ontario) Canada K0J 1J0
CW-123740-CONF-012 2014 ANS Winter Meeting and Nuclear Technology Expo November 9-13, 2014, Anaheim, CA, U.S.A. Extended Abstract / Short Paper for ANS Transactions
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Simulations of a Heavy-Water Reactor Operating on Self-Sufficient Equilibrium Thorium Cycles †
Sourena Golesorkhi,* Blair P. Bromley, Matthew H. Kaye* * Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, Oshawa, ON † Atomic Energy of Canada Limited, Chalk River, ON
[email protected];
[email protected];
[email protected] INTRODUCTION The 233U/thorium fuel cycle has been studied as an alternative to the plutonium/uranium cycle since the earliest days of nuclear energy research [1]. Thorium fuel cycles generally fall into one of two categories: “OnceThrough Thorium” (OTT) and “Self-Sufficient Equilibrium Thorium” (SSET) [2]. The former is characterized by a fuel mixture of thorium and plutonium or enriched uranium. The latter depends on mixing 232Th with an initial amount of 233U to sustain criticality. The goal of the SSET cycle is to achieve a cycle where the amount of fissile nuclides present in the fuel when it is discharged is greater than or equal to that which it started with, with fissile inventory ratio (FIR) being ≥ 1.0. Heavy water reactors (HWR) offer greater neutron economy due to reduced neutron absorption by using heavy water as a moderator and coolant. Furthermore, the on-line refueling capability of Pressure-Tube Heavy Water Reactors (PT-HWR) allows more versatility in fuelling schemes, including the use of a more heterogeneous seed/blanket core to enhance breeding. These properties make PT-HWRs well suited for the thorium fuel cycle. Thorium-based fuel cycles in PTHWRs have been investigated extensively in the past, as described in Reference 2, although the majority of practical studies have focused on OTT cycles with homogeneous core configurations. The objective of this work is to perform reactor physics simulations of heterogeneous seed/blanket PT-HWR cores, attempting to obtain net breeding, with the conversion ratio and fissile inventory ratio being greater than unity (CR ≥1.0, FIR ≥ 1.0). This paper is a continuation of previous work described in Reference 3.
DRAGON was used in conjunction with an open-source 69-group nuclear data library provided by the WIMS Library Update Project (WLUP) based on ENDF/B-VII.0 [6]. DONJON was used to simulate a full reactor model of a 700-MWe-class PT-HWR (with 380 fuel channels, 12 fuel bundles per channel and 2,061 MWth core power) [7,8]. This model included all flux detectors and various reactivity control devices. To obtain improved neutron economy, all adjuster rods were initially fully withdrawn from the core. The liquid zone controllers (LZC) are the main devices for reactivity control, and thus were kept empty to allow an additional margin of control. The target discharge burnup for different regions of fuel in the reactor were adjusted iteratively to achieve several operational targets and limits. An effective multiplication constant (keff) of between 1.002 and 1.003 was targeted. As previously mentioned, the major goal of the study was to obtain a self-sufficient fuel cycle (CR ≥1.0, FIR ≥1.0). As a feasibility and safety measure, a maximum bundle power of 935 kW was enforced, which is a typical limit for 37-element natural uranium fuel bundles in PT-HWRs [8]. The nominal reactor power was maintained at 2,061 MWth, but was de-rated if necessary to avoid exceeding channel (7.3 MW) and bundle (935 kW) power limits [9]. The two main configuration types examined in this study were based on those studied in previous work [3]. The first configuration type (named Inner Seed Outer Blanket - ISOB) is composed of 184 inner channels with higher-reactivity, lower-breeding seed fuel. This region is surrounded by 196 channels of lower-reactivity, higherbreeding blanket fuel. The IBOS (Inner Blanket Outer Seed) configuration is the reverse of this, with 184 central blanket channels and 196 peripheral seed channels. These two configurations are shown in Fig. 1.
METHODOLOGY RESULTS AND ANALYSIS Simulations were carried out using the DRAGON and DONJON reactor physics codes, which are open source codes developed at L’École Polytechnique de Montréal. DRAGON is a lattice physics code that solves the multi-group neutron transport equation [4]. DONJON is a finite reactor analysis code that solves the few-group neutron diffusion equation [5]. DRAGON was used to model lattice cells based on the 37-element and 28element fuel bundles used in conventional PT-HWRs.
The largest challenge faced in meeting the objectives of the study was keeping the maximum bundle power (generally in the seed region) below the allowable limit. Although the maximum allowable bundle power (as previously stated) is 935 kW, the operational value is generally kept around 850 kW in existing reactors [9]. A potential solution to raise the bundle power limit is the use of the Internally Cooled Annular Fuel (ICAF) bundle
CW-123740-CONF-012 2014 ANS Winter Meeting and Nuclear Technology Expo November 9-13, 2014, Anaheim, CA, U.S.A. Extended Abstract / Short Paper for ANS Transactions concept currently under development [10]. A variant of the ICAF is based on an existing 28-element PT-HWR bundle design, and features an internal coolant annulus in the center of each fuel pin. The bundle power limit could be increased by 15%, or more, although much work remains to be done to better define the new upper limit.
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Revision 0 maximum fuel channel and fuel bundle power after a single refueling operation and at an instant in time.
Time-Average Model Full-core reactor physics calculations were performed using a time-average model, which gives an approximation of the equilibrium burnup distribution in the core, using time-flux-averaged macroscopic cross sections. The results of the time-average core physics calculations are shown in Table I. Table I. Results of Time-Average Calculations Parameter keff Power (MW) Seed Fuel Seed BU (MWd/kg) Blanket Fuel Blank. BU (MWd/kg) Max. Ch. Power (kW) & Location Maximum Bun. Power (kW) & Location FIR Refuel Rate (Chan/d)
ISOB ISOB IBOS @100%FP @60%FP 1.0021 1.0021 1.0019 2,061 1,443 2,061 ICAF bundle – ThO2 with 1.6% 233UO2 11.96 9.14 28-el. bundle – ThO2 with 1.2% 233UO2 1.4% 233UO2 10 16.65 12,669 7,602 6,740 Chan. M11 Chan. M11 Chan. E11 1,425 856 854 Chan. M12 Chan. M12 Chan. E11 Bun. #5 Bun. #5 Bun.#7 1.011 0.995 0.986 3.28 1.97 2.80
Even with the incorporation of the ICAF bundle for the seed (and a standard 28-element bundle for the blanket), the power in the central region of the ISOB configuration was too high. De-rating of the reactor was therefore necessary. However, the decrease in reactor power also caused a noticeable decrease in the breeding rate in the blanket fuel, to the point where FIR < 1.0. The IBOS configuration had the advantage of a much flatter and more manageable power profile (see Fig. 2); however, self-sufficiency was not achieved. Meeting the bundle power limit was most challenging. For comparison, it was found in previous studies [3] that a homogeneous core of 1.45 at% U-233/(Th+U) fuel in 37element fuel bundles could achieve an exit FIR of 1.001; however, the burnup was quite low (5.9 MWd/kg). Instantaneous Model The reactor physics analysis of the ISOB and IBOS core configurations also included instantaneous models. An instantaneous model allows the prediction of the
Quarter-Core Models
Fig. 1. Irradiation Zones and Target Burnups
Fig. 2. Channel power profile in row L
Fig. 3. Pattern for Refueling Sequence and Age in Core The creation of these models required the development of reactor channel fuelling orders. In each case, an “age map” [5] was generated based on applying a 4×4 pattern to each section of the reactor. An example of this is shown in Fig. 3. This age map essentially
CW-123740-CONF-012 2014 ANS Winter Meeting and Nuclear Technology Expo November 9-13, 2014, Anaheim, CA, U.S.A. Extended Abstract / Short Paper for ANS Transactions describes how far along in its refueling cycle a channel is. Using the generated age maps, two years’ worth of online refueling operations were simulated. The goal of the instantaneous calculations was to find the Channel Power Peaking Factor (CPPF), the maximum value of the ratio of instantaneous and time-averaged channel powers, for each configuration. The CPPF is a quantified indication of the power ripple following a refueling operation. In order to reduce the CPPF, both instantaneous and time-averaged calculations were carried out using a 4-bundle shift fuelling scheme in the seed channels and an 8-bundle shift in the blanket channels. The CPPF can also be used to extrapolate an estimate of the actual maximum bundle power (which must still obey the previous operational limits.) The results of the instantaneous calculation are shown in Table II. Both the IBOS and ISOB@60%FP had instantaneous powers within the operating margin. Table II. Results of Instantaneous Calculations Parameter CPPF Max. Channel Power (kW) & Location Max. Bundle Power (kW)
ISOB @ FP
IBOS
1.20
ISOB @ 60%FP 1.18
13,279 Chan. M13
8,138 Chan. L11
7,270 Chan. M19
1,710
1,010
999
1.17
CONCLUSIONS Previous work on the SSET cycle in PT-HWRs was extended with further full core modeling. A seed and blanket core with the seed located at the center of the reactor runs into problems with high power. The adoption of a new bundle concept allows for higher powers, but derating still is necessary. Reducing the reactor power also reduces the breeding rate in the blanket, making the reactor no longer self-sufficient in the production of 233U. A reverse configuration with the blanket in the center of the core has a flatter power distribution. De-rating is not required, but it is not self-sustaining in fuel (FIR 1.0). While both core configurations studied show excellent potential as high-conversion, near-breeder reactors, they fall just short of net breeding. Therefore, more significant design modifications will be required to achieve net breeding with higher seed burnup. To increase the blanket breeding rate, the blanket initial fissile content and burnup before removal could be reduced significantly (
