Conceptual Design of Inherently Safe Fast Reactor (ISFR)

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1 GENES4/ANP2003, Sep , 2003, Kyoto, JAPAN Paper 1015 Conceptual Design of Inherently Safe Fast Reactor (ISFR) Yoshiro ASAHI* Japan Atomic Energy Research Institute, Tokai, Ibaraki, Japan ISFR is a boiling heavy water fast reactor of process inherent ultimate safety (PIUS) type. ISFR may breed fuel in the core. Owing to a positive void coefficient, the application of the PIUS concept to ISFR is not straightforward. Thus, the gap conductance is small so that the time constant of the positive void feedback process is sufficiently large, while the initially-closed two-way check valves to be used as passive switches to the pumps are installed at the lower honeycombs. As a result, the passive shutdown mechanisms can come into effect sufficiently soon to suppress the positive feedback reactivity. Both large and the passive switches also help stabilize the system so that ISFR can perform a constant power operation with a simple control logic for the main coolant pump speed. In a steam generator tube rupture, fuel temperature was found to smoothly decrease to the decay heat level with nucleate boiling. The feasibility of ISFR was proved only to some extent. Keywords : fast reactor, fuel breeding, tight lattice, minor actinides, inherent safety, PIUS, positive void coefficient, heavy water, control, stability, gap conductance, the THYDE-NEU code 1. Introduction In this work, we will present a conceptual design of Inherently Safe Fast Reactor (ISFR), which is hoped to help form a viable energy supply system. In discussions of inherent safety 1, 2), there are two key words, namely, fuel integrity and passive safety. In inherently safe reactors, the safety criterion is maintenance of fuel integrity, while, for example, in the present licensing of existing light water reactors (LWRs), the safety criterion is maintenance of coolable fuel geometry and hence a loss of fuel integrity is not prohibited. Passive safety of a reactor as a whole is referred to as its inherent safety, which can be quantified in terms of a walk-away (or grace) period. At the beginning of the eighties, the process Tel , Fax. (F) , y.asahi@popsvr.tokai.jaeri.go.jp inherent ultimate safe (PIUS) reactor 1, 2) was proposed as an inherent safe pressurized water reactor (PWR). But, at that time, the LWR technology had been almost established and hence it was difficult for a new LWR concept such as the PIUS reactor to be accepted as commercial reactors. On the other hand, fast breeder reactors (FBRs) are now under development so that it is not too late to apply the notion to them. ISFR is an FBR with a positive void coefficient. It has been common to prohibit a positive void coefficient, since a reactor with a positive void coefficient could have a prompt criticality in accidents due to an accelerated positive reactivity feedback. In this work, applying the PIUS concept to ISFR, we will try to cope with the problem. Originally, however, the PIUS concept was intended for applications to PWRs 1, 3) with a negative void coefficient. Therefore, its 1

2 application to ISFR is expected not to be straightforward, but to need new contrivances. In this work, the THYDE-NEU code 4, 5) will be applied to dynamical calculations of ISFR. In THYDE-NEU, the boron transport, among other things, is simulated by the equation CB) /t = - G CB) /z where z is the distance along the flow path. Boron reactivity is represented such that is an integrated reactor such that the circulating primary system (CPS) is submerged in the inner pool within the reactor pressure vessel (RPV). The RPV is, in turn, submerged in the outer pool contained in the prestressed reactor containment vessel (RCV) (80 bars). ISFR uses the same RPV and the same steam generator (SG) and the same RCV as Integrated Reactor with Inherent Safety (IRIS) 3). The main coolant pumps (MCPs) (4 - (N - N ). (1) units) are placed at the exit of the SG. The outer Note that pool contains several ISFRs and IRISs to form a N acb modular system. with a = (CA /MB) x 10-9 = 5.57 x The equivalent core diameter and the active core height are 1.91 m and 2.45 m, respectively. ISFR is loaded with mixed oxide (MOX) fuel. The volume ratio of moderator to fuel Vm/Vf is In this work, we use heavy water as the coolant in the primary system of ISFR for the next two reasons. First, heavy water has a smaller slowing down power than light water. Secondly, in heavy water, we can expect photo-delayed neutrons 6, 7) The chemical shim system (CSS) is used to compensate for a reactivity change with burnup. The thermal conductivity and the thickness of the insulator at the outer wall of the CPS shell are assumed to be x 10-3 kw/(m O C) and 5 cm, respectively. In the core and the blanket, the thermal power is 500 MWth and 33 MWth, while the power density is 71.2 kw/l and 7.04 kw/l, respectively. The thermal efficiency of ISFR is expected to be about 30 %. Wcore is 1.40 x10 3 kg/s, while WBLK is small, i.e., 87 kg/s. For WBLK = 87 kg/s, it was found that the form loss coefficient at Fig. 1 Schematic of ISFR with Passive ESFs the blanket inlet is large (~8.69 x10 3 ), while at the blanket exit is Overview of ISFR (b) Engineered Safety Features (ESFs) (a) General Description In a PIUS type reactor, the following conceptual The schematic of ISFR is shown in Fig. 1. ISFR relationship holds at a steady state 2

3 gh=kgr 2r (3) where = IP riser. Suppose that an accident happens and hence, for example, Eq. (3) will break down. Thus, borated water in the inner pool passively enters the CPS to shutdown the reactor. But, since ISFR has a positive void coefficient ( > 0), a direct application of the concept is expected not to work. In fact, in this work, it will be found necessary, for example, to place the two-way check valves (TCVs) at the lower honeycombs. ISFR has other ESFs than the PIUS configuration (Fig. 1). Among them are (1) the heat pipes 3) that transfer decay heat from the inner pool through the outer pool to the atmosphere (an ultimate heat sink), (2) the passive safety shutdown system (PSSS) 8) (95 m 3, 83.6 O C) installed in the secondary system within the outer pool, and (3) the siphon breaker gap. The regulating pump (RP) is placed in the secondary system (Fig. 1) to obtain pressure balance so that a relationship similar to Eq. (3) can hold for the PSSS. Note that PSSS also has lower and upper honeycombs, which are not shown in Fig. 1. As usual, check valves are placed in the main steam (MS) (62 bars) and feedwater (FW) lines, while relief valves are installed in the MS line and at the tops of both the RPV and the RCV (Fig. 1). Looking over Fig. 1, we may think that the RCV looks as if it were a large accumulator. Making use of this fact, we install passive valves whose actuations depend on pcg. The accumulator type isolation valves (ATIVs) shut the flows off when their pressure becomes below pcg, while the check valves at the RPV top will open to introduce the outer pool borated light water into the RPV, if ppr becomes sufficiently lower than pcg. (c) Structural Features of ISFR Fuel The fuel rods are of PWR type. The number of fuel assemblies (rectangular lattice, 24 x 24) is 48. ISFR fuel is characterized by the fact that each fuel assembly has neither a channel box nor an assembly bypass region. This is because we must exclude a possibility of channel blockages, since ISFR has a positive void coefficient. Without assembly bypass regions, the neutron spectrum can be harder. (d)tsub at the Reactor Inlet As Tsub at the reactor inlet decreases, the void volume and hence the breeding ratio in the core increase. In the present design, Tsub at the reactor inlet is 5.4 OC. On the other hand, subcooling at the MCP inlet is 29.0 m in head. (e) Average Specific Enthalpies Let hpr and hsc be the average specific enthalpies of the primary system and the part of the secondary system that is within the RCV, respectively. Then, hpr and hsc are calculated to be kj/kg and kj/kg, whose saturation pressures are 4.7 bars and 2.85 bars, respectively. (f) Scram Valves We install the scram valves at the reactor exits. They are divided into two, namely, the passive scram valves (PSVs) (Fig. 1) and the active scram valves. The PSVs are check valves which introduce the inner pool water into the CPS if ( p) = (pip priser) is sufficiently larger than ( p) = 0.12 MPa. The PSVs back up the primary TCVs. The PSVs is expected to play an important role in an RPV bottom rupture. In Fig. 1, only a PSV among the scram valves is shown. Note that the PSVs also penetrate the flow coming down from the MCPs, but the fact is not shown in Fig. 2 in order to keep the figure from becoming 3

4 complicated. (g) Reactor Bypass Flow ISFR does not have assembly bypass flows, but has the reactor bypass flow WBP (=1.39 x 10 3 kg/s). By adjusting WBP, it is possible to satisfy the condition that at the reactor exit is sufficiently large, while Wr (= Wcore + WBLK + WBP ) is large (> 2.85 x 10 3 kg/s) so that instability of ISFR can be small enough to be overcome by the MCP speed control (refer to section 6). 3. Neutronics Design Let keff max be the maximum of keff in the reactor life. As keff max increases, so does (CB)CPS max. (CB)CPS max should be as small as possible for the following two reasons. First, as (CB)CPS max increases, so does the burden of the CSS. Secondly, we must exclude a possibility of prompt criticality that may occur for large (CB)CPS and small (CB)IP. Consider an extreme case, for example, (CB)IP = 0 and (CB)CPS = 1,100 ppm. Then, in an accident, as the inner pool water enters the CPS, NB in Eq. (1) decreases from N. Hence, may become large enough to override (< 0). Due to a resultant accelerated positive reactivity feedback, we may have a prompt criticality. In order to reduce keff max as much as possible, we let ISFR fuel contain MAs such as 241 Am and 237Np. It is known that MAs act not only as fertile materials, but also as burnable poisons and hence help decrease keff max. Preliminary cell calculations were performed with the SRAC code 9). The MOX fuel is assumed to be composed of Pu, MAs and depleted uranium (DU). MAs to be used were assumed to have been cooled three years after a burnup of 33 GWD/t in PWRs (UO2 fuel, 3,410 MWth). Parameter surveys were performed for = 0.0, 0.4 and 0.8 to find a pair of EPu and EMA that give keff sufficiently close to unity at 30 GWD/t. Note that EPu + EMA + EDU = 1.0. The result is shown in Table 1. Looking over Table 1, we note that, in the upper core region, fuel breeding takes place, whereas, in the lower subcooled region, fuel consumption takes place and hence burnable poisons should be used. Table 1. EPu and EMA to give keff ~ 1.0 at Burnup = 30 GWD/t 0.0 EPu EMA keff keff stands for the change of keff during a burnup of 30 GWD/t.) 4. Time Constant of Positive Void Feedback It is generally known that, if a reactor has a positive reactivity feedback process, its time constant should be made sufficiently large. Therefore, of ISFR should be made as large as possible. To this end, we make hgap =gap/gap small, since, as hgap decreases, increases. We use Ar as the gap gas, since, as atomic mass of the gap gas increases, gap decreases, while we keep gap from becoming excessively small by giving at the time of fuel fabrication not only a sufficiently large swelling margin in rgap, but also a sufficiently large pgap (~30 bars) to avoid the creepdown of cladding tubes. Throughout this work, we assume that hgap ~ 1. kw/m 2 O Cas will be required in section 6. Note that hgap is about 5. kw/m 2 O C in existing LWRs. Moreover, note that even if hgap is 1. kw/m 2 O C, the initial peak fuel temperature may be almost the same as 4

5 in existing LWRs, since the linear heat rate of ISFR is 8.35 kw/m in the core. Table 2. Dynamics Parameters of ISFR (15 groups) hgap 1.0 kw/(m 2 O C) (CB)IP 2,000 ppm (CB)CPS 1,100 ppm D $/ O C $/( /) keff monotone decreasing as a function of C 1.01 for n > otherwise loss of on-site n = 0.05 power MCP 10. s for coast down 100. s for reactor control RP 10. s for coast down TCV 0.1 s p)tcv OPN 2 kpa for the TCV PR s 10 kpa for the TCV SC s however, we will show that it is possible to design a control logic that enables ISFR to perform a constant power operation provided that the following three conditions are satisfied. First, Wr is sufficiently large (refer to section 2). Secondly, hgap should be small (~1. kw/(m 2 O C)). Thirdly, the fictitious closed valves should be installed at the lower honeycomb below the reactor. In section 7, the fictitious valves will be upgraded as the TCVs. First, we place fictitious closed valves at the lower honeycombs. Since they are going to be upgraded as the TCVs, only for the sake of convenience we will use the symbol (p) to express the pressure difference across them. It is important to note that (p)tcv O s vanish. 5. THYDE-NEU Representation of ISFR In sections 6 and 8, THYDE-NEU will be applied to ISFR. To this end, ISFR is nodalized by 82 nodes, 73 junctions and 46 heat conductors. The main dynamics parameters are shown in Table 2.The density coefficient shown in Table 2 was obtained for (EPu, EMA) = (0.141, 0.013) corresponding to = 0.40 in Table 1. Note that can be converted to by using relationship - (l - g). A typical set of homologous curves will be used for the pumps. Prior to transient calculations, we perform adjustment calculations 4, 5) with THYDE-NEU to obtain an initial state of ISFR so as to satisfy Eq. (3), for example. 6. Constant Power Operation of ISFR Owing to > 0, ISFR can not perform a constant power operation by itself. In this section, 5 Fig. 2 Reactor Power in Constant Power Operation In order to develop an MCP speed control, we make use of the fact that as increases, the discharging pressure increases and hence decreases. The resultant control logic for is

6 such that made as small as possible. First of all, we note d/ dt = ( c) / that during the constant power operation shown where cand are given in Table 2. With the above control, THYDE-NEU yields the result as shown in Fig. 2. Note that the amplitude of the oscillation in Fig. 2 is very small and not diverging. Hence, with the control, ISFR will be in section 6, the maximum of (p) is 77.6 Pa and Pa for the primary and secondary closed valves, respectively. In view of this fact, we upgrade the initially-closed valves so that they open if regarded as performing a constant power (p) > (p) OPN, (5) operation. where (p) OPN is 2 kpa and 10 kpa, for the TCV PR s and the TCV SC s, respectively. As the 7. TCVs as Passive Switches In this section, the fictitious closed valves will be upgraded first to passively open in accidents and then to act as passive valves to the pumps. The upgraded valves at the lower honeycombs below the reactor and the PSSS will be referred to as the TCV PR s and the TCV SC s, respectively. Suppose that, in an accident, the TCV PR s and the TCV SC s will open at, say, tpius and tpsss not only to actuate the PIUS mechanism and the PSSS mechanism, but also to bring about natural circulations with time constants, say, PIUS and PSSS in the RPV and the secondary system, respectively. Then, it is required from the viewpoint of safety that not only PIUS and PSSS, but also tpius and tpsss should be made as small as possible. As MCP and RP decrease for coast downs, so do PIUS and PSSS, respectively. name implies, there are two kinds of TCVs. Suppose that if (p) > (p) OPN, then the half of them opens to allow a downward flow through the honeycomb, while if (p) < - (p) OPN, then the other half opens to allow an upward flow. Secondly, we upgrade the TCV PR s and the TCV SC s as passive switches for the power source to all the pumps. Note that an initially-closed check valve acts as if it were a switch. Suppose that, in an accident, Condition (5) is satisfied first by the TCV PR s at tpius not only to open a half of the TCV PR s, but also to trip all the pumps. Hence at tpius, not only the PIUS shutdown mechanism begins coming into effect, but also additional large external disturbances will be introduced into the secondary system to amplify (p) in Condition (5). As a result, tpsss becomes smaller. Especially, PIUS is required to satisfy the following condition ; PIUS <. (4) 8. Safety Analysis Among design basis accidents of ISFR are In Condition (4), has already been decided by setting hgap to be 1. kw/(m 2 O C)) in the previous SGTRs and flow blockages in the core. Since ISFR does not have channel boxes, a large flow section. Note that Condition (4) is rather blockage is unconceivable. In the following, we symbolical, since neither PIUS nor will be explicitly calculated in this work. In the following procedure, the fictitious valves will be upgraded so that tpius and tpsss can be will show the calculated results (up to 180 s) with THYDE-NEU for an SGTR under the conditions shown in Table 2. The break will be assumed to occur with time constant0.1 s. The break area is 6

7 2.01 x 10-2 m 2, which is equivalent to a double-ended rupture of 50 SG tubes. feedback. At tpsss = s, (p)tcv SC becomes (p)tcv OPN = 10.0 kpa and hence the TCV SC s open not only to induce an upward flow through the PSSS, but also to trip all the pumps. Fig. 3 SGTR : Relative Reactor Power In the calculation, the following conservative assumptions will be made. We assume first no active ESFs and secondly no scram valves (and hence no PSV). The third conservative assumption is due to the fact that, in the present ISFR noding, the flows of the FW, the MS and the CSS-feed and -bleed are to be given as boundary conditions. Thus, it will be assumed not at tpsss, but later at the time of a loss of the on-site power that the flows in the FW and CSS-feed lines begin decreasing with time constant 0.1 s, while the flows in the MS and CSS-bleed lines begin decreasing with time constant 5.0 s. Soon after the occurrence of the SGTR, the break flow rapidly increases, while the reactor power rises (Fig. 3) due to the positive void Fig. 4 SGTR : Reactivities in SGTR Since the MCPs have already begun to coast down at tpsss = s, the TCV PR s also open at tpius = 1.28 s when (p)tcv PR becomes ( p)tcv OPN = 2.0 kpa. Hence, the cold inner pool water enters the CPS. As a result, the negative reactivities and (Fig. 4) are introduced and hence the reactor power turns around with the peak value n = 1.10 at 2.12 s (Fig. 3). Thus, the reactor power finally decreases to the decay heat level (Fig. 3). has the small minimum, i.e., 2.12 x 10-2 $ at 1.72 s (Fig. 4) and hence only helps decrease the power peak. In Fig. 4, is + +. An RCV isolation begins conservatively when the condition of the loss of on-site power, that is, n = 0.05 is satisfied at

8 s. Both ppr and psc decrease slowly after 60 s. The natural circulation in the secondary system through the PSSS is established during the RCV isolation which starts at 27.7 s when the loss of the on-site power occurs, while the natural circulation in the RPV is almost established after 128 s, when the break flow has become almost constant. The natural circulation in the RPV partly deviates to form the break flow. The fuel temperatures decrease very smoothly to the decay heat level in spite of the fact that ISFR has the considerably tighter lattice than to monotone increase or to be incinerated, as fuel burns. (3) It has been said that economics of a reactor enhances with its power. Then, it should be investigated how economically competitive the modular system can be as compared to large LWRs. (4) It is expected that, in ISFR, the larger the external disturbance introduced by an accident, the smaller tpius and tpsss become. Hence, it is conjectured that the consequence of an accident such as the initial power spike in Fig. 3 does not existing LWRs. It is expected that the heat pipes necessarily increase in proportion to the finally come into effect to transport the decay heat to the atmosphere. It is to be noticed that neither the bypass valve nor the relief valve is calculated to open. magnitude of the accident. The conjecture should be verified by accident analyses. (5) An extensive neutronics design should be performed, for example, to examine if keff is monotone decreasing as a function of and to 9. Future Works Items yet to be investigated include as follows. (1) In the constant power operation under the MCP speed control, ppr decreases very slowly, i.e., evaluate (CB)CPS max corresponding to keff max. () Suppose that it has been found by safety analyses that α must be less than, say, ( )1. Then, it should be confirmed that ( )1 satisfies 610 Pa in 100s. This trend should be removed. ( )2 ( )1 (6) (2) Consider a modular system formed by ISFRs and and IRISs in the RCV. Within the modular system, ( )3 ( )1, (7) the ISFRs help the IRISs by breeding fuel, while the IRISs help the ISFRs by coping with load variations in the grid and by producing more MAs than ISFRs. Note that ISFR needs MAs to reduce keff max. If MAs are insufficient, we can use 241 Am where ( )2 is the maximum of that allows ISFR to perform a constant power operation with the MCP speed control, while and ( )3 is the maximum of calculated in the neutronics design. Condition (6) states that the MCP speed to be produced by simply transmuting 241 Pu by control should be designed not to jeopardize safety decay. We should decide the number ratio of ISFRs to IRISs so that mass balance of Pu and MAs holds within the modular system. Then, except at the beginning of the site construction, no fissile material, but DU enters the site. It should be examined if Cm isotopes and fission products within the system are to be saturated or of ISFR. Suppose that it has been confirmed that Condition (6) is satisfied. Then, we only have to prove that Condition (6) holds for startups. (7) We should investigate what values are appropriate for (p) OPN and (p) OPN. (8) In the present design with WFW = 500 kg/s and xms = 0.522, the PSSS with the TCV SC s is 8

9 sufficiently stable. WFW should be made as small the magnitude of the external disturbance as possible so that xms can be as large as possible. Note that the heat transfer mode at the secondary side of the SG tube wall is mostly nucleate boiling and hence is independent of WFW. introduced by the accident. One of the design basis accidents of ISFR is SGTRs. In this work, the double-ended rupture of 50 SG tubes was analyzed up to 180 s. The fuel temperatures decreased very smoothly to the 10. Conclusions It was found that, owing to > 0, a direct application of the PIUS concept to ISFR does not work. For example, ISFR is unstable by itself. According test calculations for WFW = 500 kg/s, it was found that, if Wr is larger than 2.85 x 10 3 kg/s, ISFR instability is small enough to be overcome in the following way. We assumed that hgap is 1.0 kw/(m 2 O C) so that can be sufficiently large. Moreover, we installed fictitious closed valves at the lower honeycombs. Then, it was found possible to design the simple MCP speed control that enables ISFR decay heat level. ISFR should be designed such that if (CB)IP is less than 2,000 ppm, neither a startup nor a constant power operation should be possible. Moreover, from the view point of defense-in-depth, we should make keff max and hence (CB)CPS max as small as possible. To this end, MAs will be added to ISFR fuel. There are a number of items yet to be investigated. The most important among them is to obtain the walk-away period of ISFR and to prove that it is sufficiently large. In this work, the objective was achieved only to some extent. to perform a constant power operation. In the resultant control, MCP = 100 s is sufficiently large so that the control can not hinder the ESFs from functioning. Suppose that Condition (6) is unsatisfied withmcp = 100 s. Then, by making MCP larger and hence by degrading the MCP speed control, we can reduce ( )2 to satisfy Condition (6). Next, the fictitious closed valves were upgraded as the initially-closed TCVs which open when Condition (5) is satisfied so that ISFR can be made compatible with both the PIUS shutdown mechanism and the PSSS mechanism. It is important to note that this is possible only if ( p) O s practically vanish. The TCVs were further upgraded as passive switches to all the pumps. As a result, it is expected that the consequence of an accident such as the power peak in Fig. 3 is not necessarily in proportion to Appendix A References 1) D. Babala et al., Pressurized Water Reactor Inherent Core Protection by Primary System Thermohydraulics, Nucl. Sci. Eng., 90, 400 (1985) 2) C. Pind, Secure Heating Reactor, Nucl. Technol., 79, 175 (1987) 3) Y. Asahi et al., Conceptual Design of the Integrated Reactor with Inherent Safety, Nucl. Techno., 91, 28 (1990) 4) Y. Asahi, A Spatial Kinetics Method Ensuring Neutronic Balance with Thermal-Hydraulic Feedback and Its Application to a Main Steam Line Break, Nucl. Sci. Eng., 139, 78 (2001) 5) Y. Asahi, THYDE-NEU : Nuclear Reactor System Analysis Code, JAERI-Data/Code , March ) Berstein et al., "Yield of Photo-Neutrons from 9

10 U235 Fission Products in Heavy Water ", Phys. Rev., 71, 573 (1947) 7) G. R. Keepin et al., "Delayed Neutrons from Fissionable Isotopes of Uranium, Plutonium, and Thorium", Phys. Rev., 107, 1044 (1957) 8) Y. Asahi et al., Improvement of Passive Safety of Reactors, Nucl. Sci. Eng., 96, 73 (1987) 9) M. Tsuchihashi et al., Revised SRAC Code System, JAERI-1302, JAERI (1986) Appendix B Nomenclature CA Avogadro number CB Boron concentration (ppm) E Enrichment or weight fraction g Gravitational acceleration (= 980 m/s 2 ) G Mass flux (kg m -2 s -1 ) h Specific enthalpy (kj/kg) hgap Gap conductance (kj/(m 2 s O C) H Height between the upper and lower hot/cold interfaces (m) Neutron multiplication factor K Form loss coefficient M Atomic mass n Relative reactor power N Atomic number density (cm -3 ) p Pressure (Pa) rgap Gap width (m) xms Mass quality of main steam z Distance along the flow path (m) Void fraction Tsub Subcooling ( O C) Delayed neutron fraction Thermal conductivity (kj/(m s O C)) eactivity ($) Void reactivity coefficient ($ ( /) 1 ) D Doppler reactivity coefficient ($ ( O C) 1 ) Density reactivity coefficient ($ ( /) 1 ) Coolant density (kg m -3 ) Microscopic cross-section (cm -2 ) Time constant (s) Macroscopic cross-section (cm -1 ) Relative MCP speed Sub- and super-scripts a Refers to absorption. B Refers to boron. BLK Refers to the blanket. BP Refers to the reactor bypass. CG Refers to the cover gas. core Refers to the core. D Refers to the Doppler effect. DU Refers to depleted uranium. eff Refers to an effective value. FW Refers to the feedwater. gap Refers to the fuel gap. IP Refers to the inner pool. max Refers to a maximum. MA Refers to minor actinides. MCP Refers to the main coolant pumps. O Refers to an initial or steady state. OPN Refers to a condition to open a valve. p Refers to pressure. PIUS Refers to the PIUS mechanism. PR Refers to the primary system. PSSS Refers to the PSSS. Pu Refers to plutonium. r Refers to the reactor. riser Refers to the riser. SD Refers to a reactor shutdown state. SC Refers to the secondary system within the RCV. sub Refers to subcooling. tot Refers to total reactivity. Refers to void. Refers to coolant density. 10

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