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Dynamic simulation of accidental closure of intermediate heat exchanger isolation valve in a pool type LMFBR
Annals of Nuclear Energy 38 (2011) 748–756
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Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Dynamic simulation of accidental closure of intermediate heat exchanger isolation valve in a pool type LMFBR K. Natesan ⇑, N. Kasinathan, K. Velusamy, P. Selvaraj, P. Chellapandi, S.C. Chetal Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, India
a r t i c l e
i n f o
Article history: Received 31 March 2010 Received in revised form 18 November 2010 Accepted 7 December 2010 Available online 7 January 2011 Keywords: Fast reactors Plant dynamics Safety analysis PFBR Sleeve valve
a b s t r a c t In a pool type liquid metal cooled fast breeder reactor (LMFBR), core and other internals such as pumps, heat exchangers are immersed in a pool of sodium. Heat exchange from primary sodium circuit (pool) to secondary sodium circuit (loop) is through four intermediate heat exchangers (IHX) immersed in primary sodium pool. Each IHX is provided with a sleeve valve at its primary sodium inlet window for the purpose of isolating the shell side of IHX from the sodium pool. With such a provision, an inadvertent partial or complete closure of a sleeve valve of one of the IHX during normal operation of the reactor has been considered as a design basis event for the reactor. One dimensional transient thermal hydraulic models of the primary and secondary sodium circuits have been developed to study the thermal hydraulic consequences of such an event. The main areas of concern in the plant and the availability of safety parameters for the detection of the event have been evaluated. Ó 2010 Elsevier Ltd. All rights reserved.
1. Introduction Plant dynamic analysis of various events that can happen in a plant is very important for establishing safety of a plant. This study is also important for planning operating procedures and to design various systems and components of the plant to safely withstand the consequences of various anticipated thermal hydraulic transients. Such studies take a leading role in the licensing of a nuclear power plant. Numerical simulation of the dynamic conditions following an initiating event requires special purpose computer codes to be developed specific to the plant design. Restricted analytical models and associated computer codes such as IANUS (Additon et al., 1973) and DEMO (WARD report, 1975) have been developed to simulate the overall response of US LMFBR plants fast flux test facility (FFTF) and clinch river breeder reactor plant (CRBRP) respectively. IANUS code models primary and secondary circuits of the plant with a dump heat exchanger whereas DEMO code models steam generating modules also. Another code, called NALAP (Martin et al., 1975) is also available for the transient simulation of LMFBR developed from the famous code RELAP-3B (RELAP-3B Manual, 1974) which is the system code for water cooled reactors. This code has been developed by substituting sodium properties in place of water in the RELAP 3D code and is capable of predicting rudimentary decay of core flow following a pipe rupture accident. A more generalised code SSC for both loop and pool type LMFBR has been developed by Brookhaven National ⇑ Corresponding author. Tel.: +91 044 27480500/22109; fax: +91 044 27480104. E-mail address: [email protected] (K. Natesan). 0306-4549/$ - see front matter Ó 2010 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2010.12.005
Laboratory (Guppy, 1983). All these studies reflect the importance of dynamic simulation in the successful design of a nuclear power plant. This is paper discusses the thermal hydraulic modules developed for studying the consequences of one of the postulated initiating events in the 500 MWe Indian prototype fast breeder reactor (PFBR), currently under construction. 1.1. Description of the plant The reactor core of a LMFBR where heat is generated by controlled nuclear fission reaction is immersed inside a large pool of sodium named primary sodium circuit along with other components such as pumps and heat exchangers. Sodium pool and primary circuit components are contained inside a large vessel, known as main vessel. The heat produced in the primary sodium circuit is transferred to an intermediate sodium circuit from where it is transferred to steam water system to produce superheated steam in a steam generator. Schematic layout of the heat transport path of LMFBR is shown in Fig. 1. The primary sodium pool is divided into two parts viz. hot pool and cold pool by an internal structure known as inner vessel. Sodium from the cold pool is circulated through the core by two centrifugal pumps operating in parallel. A schematic of primary sodium circuit is shown in Fig. 2. Sodium from hot pool flows through four intermediate heat exchangers (IHX) to cold pool. IHX transfer heat produced by the core to secondary sodium loop. Primary sodium flows from top to bottom in the shell side and the secondary sodium flows through the tubes from bottom to top. The four IHX are connected to the two secondary sodium loops with two IHX in each loop.
K. Natesan et al. / Annals of Nuclear Energy 38 (2011) 748–756
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Nomenclature CT HP I IS IE IO IPL PDr PE PP PPS Pi RI SDr SG SP SPE ST c
List of symbols A flow inertial coefficient, m1 D pump developed head, Pa I moment of inertia, kg m2 K pressure drop coefficient, Pa/(kg s1)2 P pressure, Pa Q mass flow rate, kg s1 S area of cross section, m2 Z elevation, m g acceleration due to gravity, m s2 n number of SA in a radial group t time, s a normalised pump speed b torque, N m q density, kg m3 h temperature x pump speed, rad/s Subscripts CE core inlet CP cold pool
Schematic flow sheet of primary and secondary sodium circuits are shown in Figs. 3 and 4 respectively. More details of the plant layout are available elsewhere (Chetal et al., 1995). 1.2. Event classification Each IHX is provided with a sleeve valve at its primary sodium inlet window. Purpose of the sleeve valves is to isolate the shell side of the IHX from primary sodium circuit during an anticipated single secondary loop operation of the reactor. With such a provision, an inadvertent partial or complete closure of a sleeve valve of one of the four IHX during normal operation has been considered as a design basis event (DBE) for the reactor. Various DBE that can occur in the plant are classified into four categories based on the frequency of their occurrence, namely, Category 2 (>102/ry), Category 3 (6102/ry but >104/ry) and Category 4 (6104/ry but >106/ry) with Category 1 being all the planned operations. ‘ry’
core top hot pool intermediate heat exchanger (IHX) IHX secondary side IHX inlet IHX outlet inlet plenum primary pump drive primary pump inlet primary pump primary pump suction primary pipe reactor inlet secondary pump drive steam generator secondary pump secondary pump entry surge tank core channel
stands for reactor operation year. General approach that is adopted in the categorisation is either probabilistic or deterministic. Sometimes a combined or balanced approach is also adopted. Categorisation of various events is important to design various plant systems such that the consequences of the event are confined within pre-specified limits. The limiting parameters are arrived at based on the operating life of various systems with tighter limits specified for the more frequently occurring events. IHX sleeve design is not present in many reactors. The event of accidental closure of sleeve has not been reported to occur in any reactors. Thus, this event is a very rare event and has been classified as Category 3 event by deterministic approach. 1.3. Main concerns following the event In case of a sleeve valve closure in one of the IHX, viz. IHX-1, the total sodium flow pumped by the primary pumps has to flow
Fig. 1. Heat transport path of an LMFBR.
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Fig. 2. Schematic of primary sodium circuit.
through the available opening in IHX-1 and through the other three IHX viz. IHX-2, IHX-3 and IHX-4. IHX-1 primary flow comes down to leak flow value and the other three IHX flows correspondingly increase. Since there are no corresponding changes in the secondary sodium flows through IHX, the IHX-1 primary outlet temperature falls to a low value (close to secondary sodium inlet temperature) and the other three IHX primary sodium outlet temperatures increase. However, since the temperature of majority of the sodium inflow to the cold pool increase, cold pool sodium temperature (hRI) also increases. Since there is automatic plant protection action of reactor trip (SCRAM) based on 10 K increase in cold
pool sodium temperature from its nominal value, reactor trip may happen. The possible reduction in the core flow would be very small due to the small pressure drop offered by the IHX shell side (1.5 mlc) compared to the total primary sodium circuit pressure drop (75 mlc). Even though there is an increase in the reactor inlet temperature, the negative reactivity feedback due to it causes the reactor power to fall. The combined result of these two opposing effects prevents the excessive rise of core assembly temperatures. The reduction in primary sodium flow through IHX-1 from hot pool to cold pool causes increase in the primary sodium flow through IHX2-4. Therefore, there will be an increase in the hot pool
K. Natesan et al. / Annals of Nuclear Energy 38 (2011) 748–756
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Fig. 3. Schematic model of primary sodium circuit.
Fig. 4. Schematic model of a secondary sodium circuit.
sodium level and decrease in the cold pool sodium level in order to sustain the increased pressure drop across IHX. This may be a concern for the buckling of inner vessel and sodium overflow from the hot pool through the inner vessel top. The increased primary sodium flow through IHX-2 and IHX-3 may also cause the flow induced vibration of tubes. Near the IHX top tube sheet-shell weld junction, the outside of the shell will be at a temperature corresponding to hot pool sodium temperature. Since, there is only very small primary sodium leak flow through the IHX-1 whose sleeve valve is closed, it may not wash the bottom of the top tube sheet properly to maintain it hot. Hence, the top tube sheet will be at a temperature close to that of secondary sodium outlet. Because of the reduction in the primary sodium flow, the secondary sodium outlet temperature of IHX-1 falls sharply. Hence, a large DT (temperature difference) will be caused across the top tube sheet-shell weld of IHX-1 (DTIHXSh). The increase in the primary sodium flow through IHX-2 and IHX-3 causes the primary sodium temperature at their outlets to increase. This results in the increase of DT across the cold end tube sheet (DTIHXCE). The reduction in IHX-1 primary sodium flow causes the secondary sodium outlet temperature of IHX-1 to
decrease. The increase in primary sodium flow through the IHX-2 causes the secondary outlet temperature of IHX-2 to increase. Hence, the surge tank of loop-1 will be supplied by sodium through two pipes having different temperatures (DTSuT). This is of concern for the surge tank from structural integrity considerations. The net reduction in the primary sodium flow through the IHX of loop-1 causes the corresponding loop SG sodium inlet temperature and hence the steam temperature to decrease. But, steam temperature of loop-2 increases due to the increased primary sodium flow through its IHX, which results in large DT between the steam produced in two loops (DTSTM). As explained earlier, IHX-1 primary sodium outlet temperature falls and that of other IHX rises during the event. This can cause circumferential DT on the main vessel, inner vessel and other structures in the cold pool. However, IHX-1 primary sodium flow being very small compared to that through other IHX, the circumferential DT may not be of much concern. This paper brings out the details of the mathematical model of primary and secondary sodium circuit hydraulics to quantify the parameters of concern explained in the above paragraphs during an instantaneous closure of one IHX sleeve valve.
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2. Mathematical model The governing equations for the primary sodium circuit hydraulics are obtained by making momentum balances between IHX inlet and IHX outlet, pump inlet and core inlet and core inlet and core outlet, by mass balances for the hot and cold pools and by torque balance on the primary pump shaft. Similarly, the governing equations for secondary sodium circuit hydraulics are obtained by making momentum balances between surge tank outlet to pump tank inlet through the SG modules and pump tank inlet to surge tank inlet through secondary sodium pump (SSP) and IHX, by mass balance for surge tank and by torque balance on secondary pump shaft. Specific features of the model are as follows. Sodium flow is considered as one dimensional and incompressible. Head developed and impeller-consumed torque by the pump are obtained from homologous pump characteristics available from Streeter and Wylie (1967). Friction factors and hence the friction pressure drop coefficients in pipes and rod bundles (Novendstern, 1972) under turbulent flow conditions, are proportional to Re0.25, (Re = Reynolds number). All the pressure drop coefficients in the primary and secondary hydraulics model are considered to vary with flow by this relationship. Further, even under low flow conditions in the primary and secondary sodium circuits, the flows are turbulent. Governing equations used in the analysis are given below: 2.1. Primary sodium circuit hydraulic model The governing equations of the primary sodium circuit hydraulics model are: Momentum balance between IHX inlet and outlet
AI
dQ Il ¼ ðZ HP Z IE ÞqHP g ðZ CP Z IO ÞqCP g K Il Q Il jQ Il j dt Z ZIO ql dZ for l ¼ 1 to 4 g
where f1 and f2 are the functions evaluating the homologous characteristics. The empirical relation used for the friction in the pump drive system bPP (fri) other than the impeller is:
14:77a; jaj 0:005 bPP ðfriÞ ¼ 0:117 8:797a; 0:005 < jaj 0:0117 bPP ðtotÞ 0:023a þ 0:012; jaj > 0:0117 Mass balance for core flow 10 X
nr Q cr ¼
r¼1
KIl is the resistance of the path through the IHX whose sleeve valve is closed. Momentum balance between pump inlet and outlet
dQ PPl ¼ ðZ CP Z IO ÞqCP g ðZ HP Z CT ÞqHP g DPcore dt K P Q PPl jQ PPl j þ DPPl for l ¼ 1; 2
ð2Þ
where KP = KPPS + KPi + KIPL. Momentum balance for individual core zone between core inlet and outlet
Acr
dQ cr ¼ DPcore K cr Q cr jQ cr j g dt
Z
Z CT
qr dZ; for r ¼ 1 to 10 ð3Þ
Z CE
Mass balance for hot pool
" # " # 2 4 X X dZ qHP SHP HP ¼ ðQ PPl Þ ðQ Il Þ dt l¼1 l¼1
ð8Þ
DPcore
" # P dQ RIj þ 10 r¼1 nr ar br ¼ P10 r¼1 nr ar
where dt = tj tj1
ð4Þ
dQ RIj
¼
" 2 X
ð9Þ for r ¼ 1 to 10
#
Q PPlj Q RIj1
l¼1
1 ar ¼ Acr þ K cr jQ crj1 j dt br ¼ K cr jQ crj1 jQ crj1 þ dr Z ZCT dr ¼ g qr dh
and
Z CE
2.2. Secondary sodium circuit hydraulics model The governing equations of the secondary sodium circuit hydraulics model of each loop are: Momentum balance between pump entry and surge tank
AIS
dQ ISl ¼ PSPE P ST þ DSP K ISl Q ISl jQ ISl j g dt
Z
Z ST
Z SPE
ql dZ
for l ¼ 1 to 2
ð10Þ
Momentum balance between surge tank and pump entry
" # " # 4 2 X X dZ qCP SCP CP ¼ ðQ Il Þ ðQ PPl Þ dt l¼1 l¼1
ð5Þ
Torque balance for pump
ASG
dQ SGl ¼ PST PSPE K SGl Q SGl jQ SGl j g dt
for l ¼ 1 to 4
dxPPl ¼ bPDrl bPPl ; for l ¼ 1; 2 dt DPPl ¼ f 1ðQ PPl ; xPPl Þ; for l ¼ 1; 2
Z
Z SPE
Z ST
ql dZ ð11Þ
Mass balance between flow from pump and flow from surge tank
IPP
¼ f 2ðQ PPl ; xPPl Þ þ bPPl ðfriÞ
Q PPl
Numerical solution of the primary circuit hydraulics is obtained in two stages. In the first stage, IHX flows, pump flows, pump speeds and the sodium levels are calculated utilizing a standard ordinary differential equation solver based on the Hamming’s Predictor–Corrector method. Using the total core flow obtained from the first stage as input for the second stage, core zone equations are solved. For this a semi-implicit finite differencing and linearisation technique (Agarwal et al., 1977) is applied on the governing equations for individual core zone flows. Also by utilizing the fact that change in total core flow (dQRI) computed in the first stage described above to be equal to the sum of the changes in the individual zone flows, the current time individual zone flows are obtained as follows:
Mass balance for cold pool
bPPl ðtotÞ ¼ bPPl ðimpellerÞ þ bPPl ðfriÞ;
!
l¼1
Z IE
APPl
2 X
Q crj ¼ Q crj1 þ DPcore ar ar br ; ð1Þ
ð7Þ
ð6Þ for l ¼ 1; 2
2 4 X X ðQ ISl Þ ¼ ðQ SGl Þ l¼1
l¼1
ð12Þ
K. Natesan et al. / Annals of Nuclear Energy 38 (2011) 748–756
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Mass balance for surge tank
qST SST
" # " # 2 4 X X dZ ST ðQ ISl Þ ðQ SGl Þ ¼ dt l¼1 l¼1
ð13Þ
Torque balance for secondary sodium pump
dxSP ¼ bsDr bSP dt DSP ¼ f 1ðQ SP ; xSP Þ
ISP
ð14Þ
bSP ðtotÞ ¼ bSP ðimpellerÞ þ bSP ðfriÞ ¼ f 2ðQ SP ; xSP Þ þ bSP ðfriÞ where f1 and f2 are similar functions as that of PSP as explained above. This set of simultaneous equations is also solved by Hammigs Predictor Corrector method to obtain the evolution of secondary sodium flow. 2.3. Overall thermal–hydraulics model The hydraulic models for predicting the flow evolution in primary and secondary sodium circuits are interfaced with thermal models of various segments of primary and secondary sodium circuits, viz., core, hot pool, cold pool, IHX, secondary sodium piping and steam generator in the plant dynamics code DYANA-P. This code also contains neutronic models for predicting the fission power evolution in the core considering various reactivity feedback effects. Flow evolutions obtained as above are used as input for estimating the temperature evolutions in various parts of the plant using the other modules of the DYANA-P code. Thermal modeling adopted for core, IHX, piping and steam generators are similar to those adopted in the DYNAM code developed for analysing the plant dynamics of fast breeder test reactor (FBTR) under operation in India. FBTR is a 40 MWt loop type reactor, whereas PFBR is of pool type. Hydraulic models used in the simulation of pool hydraulics of PFBR are elaborated in the previous section. Thermal models are similar in both the types of reactors. Detailed description of thermal models adopted in the DYNAM code is given in Vaidyanathan et al. (2010). These methodologies adopted in the formulation of the DYANA-P code have been validated through the commissioning tests carried out in FBTR (Vaidyanathan et al., 1994).
Fig. 5. Evolution of IHX primary flow.
temperature). The primary and secondary sodium outlet temperatures of other three IHX increase. However, since the temperature of majority of the sodium inflow to cold pool increase, reactor inlet temperature (hRI) increases as seen in Fig. 6. Negative reactivity feedback due to the increase in the reactor inlet temperature causes the reactor power to fall. PFBR operates with manual power control procedure due to the fact that burnup compensation requirement is very less compared to that in thermal reactors. Since, power transient lasts only for a few minutes the operator is not expected to intervene and compensate for fall in reactor power within this short time. Operator identifies the occurrence of the event by alarms before he compensates the power. Increase in reactor inlet temperature and reduction of reactor power are two opposing effects on the fuel elements that prevent excessive rise of fuel clad and coolant temperatures. hRI increases by 10 K above nominal at 62 s and automatic reactor
3. Results and discussion Sleeve valves are large in size and a definite closure time is involved in the valve closure process. However, for the purpose of conservative estimation of thermal hydraulic consequences of the event, valves are considered to close instantaneously. A primary sodium leak flow rate of 1% has been considered through the IHX in which the sleeve valve is closed. In the subsequent discussion this IHX is termed as ‘IHX-1’, while the other IHX of the same loop is termed as ‘IHX-2’ (both in ‘loop-1’). The IHX of the second loop (‘loop-2’) are referred as ‘IHX-3’ and ‘IHX-4’. During this event, the total sodium flow pumped by the primary pumps has to flow through the available opening in IHX-1 and through the other three IHX. IHX-1 primary flow comes down to leak flow value and the other three IHX primary flows correspondingly increase to 132% as shown in Fig. 5. The increased primary sodium flow through IHX-2 to IHX-4 may cause flow induced vibration of the tubes. Reduction in the total core flow is negligible due to the small resistance offered by the IHX shell side compared to the total primary sodium circuit hydraulic resistance. Since there are no corresponding changes in the secondary sodium flows, IHX-1 primary and secondary sodium outlet temperatures fall to low values (close to secondary sodium inlet
Fig. 6. Evolutions of reactor inlet temperature and power.
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trip action (SCRAM) is initiated by it (safety action) due to which, the reactor power falls steeply as seen in Fig. 6. Primary and secondary pumps coast down to 20%, following SCRAM. Secondary sodium temperature at the outlet of IHX responds faster compared to reactor inlet temperature during this event as inertial effects do not attenuate it much. This parameter decreases at the outlet of affected IHX and increases at the outlet of unaffected ones. Decrease in this parameter can be considered as a SCRAM parameter for quick protection during this event. However, in the case of events causing loss of cooling to IHX, secondary sodium temperature at the outlet of IHX increases. Thus, this parameter is effective only for IHX sleeve valve closure event. Considering this parameter as an additional SCRAM parameter will lead to increase in the total number of SCRAM parameters and thereby affecting the plant availability. However, as already discussed there is no concern over core safety during this event and hence quick automatic safety action is not essential for this event. Automatic detection and manual SCRAM are sufficient for this event. Reactor inlet temperature which is mainly envisaged as a SCRAM parameter for protection against loss of cooling in the secondary and tertiary circuits is definite and reliable alarm parameter for this event (threshold at 5 K above nominal). However, in order to improve the reliability of event detection based on difference between the secondary sodium outlet temperatures of IHX in each loop is added as an additional alarm parameter in the plant.
falls sharply. Hence, the top tube sheet-shell weld of IHX-1 is subjected to a large temperature difference (DTIHXSh) as shown in Fig. 8. Increase in the primary sodium flow through IHX-2 to IHX-4 causes primary sodium temperature at their outlets to increase. This in turn results in increase in temperature difference across the cold end tube sheet (DTIHXCE) as shown in Fig. 9. Reduction in the IHX-1 primary sodium flow causes the secondary sodium outlet temperature of IHX-1 to decrease. Increase in the primary sodium flow through the other IHX causes the secondary outlet temperature in them to increase. Hence, the surge tank of
3.1. Consequences on structures The increase in the primary sodium flow through IHX-2 to IHX4 causes an increase in the hot pool sodium level and decrease in the cold pool sodium level in order to sustain the increased pressure drop across IHX for the increased primary sodium flow. The evolutions of sodium free levels in hot and cold pools are shown in Fig. 7. The maximum level difference reached during the event does not cross the limiting value of 4.5 m and hence there is no risk of buckling of inner vessel. Because of the reduction in the primary sodium flow, the secondary sodium outlet temperature of IHX-1
Fig. 8. Evolution of DTIHXSh of IHX-1.
Fig. 7. Evolutions of sodium free levels.
Fig. 9. Evolution of DTIHXCE of IHX.
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Fig. 10. Evolution of DTSuT.
Fig. 11. Evolution of DTSTM.
loop-1 will be supplied by sodium through two pipes having different temperatures (DTSuT). The evolution of DTSuT is shown in Fig. 10. Net reduction in primary sodium flow through of loop-1 IHX causes the corresponding loop SG sodium inlet temperature and hence the steam temperature to decrease. Whereas, loop-2 steam temperature increases due to the increased primary sodium flow through its IHX. This results in large temperature difference between the steam produced in two loops (DTSTM) as shown in Fig. 11. The evolutions of IHX primary outlet temperatures of IHX-1 and IHX-2 are shown in Fig. 12. It can be seen that there is a maximum difference of 108 K between the primary sodium outlet temperatures of IHX-1 and IHX-2. Primary outlet temperature evolutions of IHX-3 and IHX-4 are very close to that of IHX-2. As far as cold pool is concerned, only 1% of sodium flow (IHX-1 primary flow) is coming at low temperature and the rest 99% comes at high temperature. Hence, even though the temperature difference between the streams is high, this will not cause any considerable circumferential temperature variation on the structures in cold pool. The scenario following the event is summarised in Fig. 13. 3.2. Impact of reactor SCRAM and event detection Maximum values of various parameters reached during the event with and without reactor SCRAM action are shown in Table 1. It can be seen that reactor trip is not helpful in reducing the maximum values of sodium level and DTs. In fact, the maximum values of DTs on various structural components increase due to SCRAM. This is due to the differences in flow coasting of primary and secondary sodium pumps following SCRAM. However, level difference between the sodium pools reduces sharply following reactor trip and component DTs start decreasing very slowly after about 100 s. The occurrence of complete closure of one of the four IHX sleeves valve would be indicated by a reduction in the secondary outlet temperature of IHX-1 and increase in the same of IHX-2 to IHX-4 resulting in the evolution of DTISO. This parameter is added
Fig. 12. Evolution of IHX primary outlet temperatures.
to the list of automatic alarm parameters of the plant to improve the reliability of event detection. Apart from this, there will be alarm available from the reactor inlet temperature monitoring system when its value increases by 5 K above nominal. The magnitude to which DTISO rises during the event would give an indication of the extent of valve closure. The increase in IHX cold end DT (DTIHXCE) of IHX-2 to IHX-4 with primary and secondary sodium flows remaining unchanged is also an another unique feature that would happen during the event.
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Fig. 13. Scenario following the event.
Table 1 Maximum values of various parameters reached during the event. Plant condition
Normal operation Max. values reached without safety actions Max. values reached with safety actions
Level difference between hot and cold pools, m
DTIHXSh, DTIHXCE, DTSuT, DTSTM, K K K K
1.5 2.498
38 220
42 75
0 195
0 97
2.498
211
108
201
100
4. Conclusions Detailed investigations of sleeve valve closure event have been carried out and the following conclusions have been arrived at: 1. There is no concern over fuel, clad and primary sodium temperatures even without any safety actions. But there are some implications on the structural integrity of components. 2. IHX tubes may be subjected to flow induced vibration (FIV) risk due to increased primary sodium flow. 3. IHX, surge tank and steam header are to be protected from the thermal loading caused by the event.
4. Level difference between hot and cold pools during the event does not lead to spill over condition for which inner vessel buckling design has been made. The study provided thermal loading on various components and the components have been designed for these loadings. References Additon, S.L., McCall, T.B., Wolfe, C.F., 1973. IANUS – Outline Description, Westingbouse Advanced Reactors Division, Waltz Mill, Pennsulvania, FPC-939. Agarwal, A.K. et al., 1977. Simulation of transients in LMFBR systems. Nucl. Sci. Eng. 64, 480–491. Chetal, S.C. et al., 1995. Conceptual design of heat transport system and components of PFBR – NSSS. In: Proc. Conceptual Design of Advanced Fast Reactors, IAEA Tech. Committee Meeting, IAEA-TECDOC-907, Kalpakkam, pp. 117–132. DEMO, 1975. LMFBR Demonstration Plant Simulation Model, DEMO, Westinghouse Advanced Reactors Division, WARD-D-0005. Guppy, J.G., 1983. Super System Code [SSC] an Advanced Thermohydraulic Simulation Code for Transients in LMFBRS, NUREG/CR-3169. Martin, B.A., Agrawal, A.K., Albright, D.C., Epel, L.G., Maise, G., 1975. NALAP: an LMFBR System Transient Code, BNL-50457. Novendstern, E.A., 1972. Turbulent flow pressure drop model for fuel rod assemblies utilising a helical wire-wrap space system. Nucl. Eng. Des. 22, 19–27. RELAP-3B, 1974. RELAP-3B Manual a Reactor System Transient Code, Brookhaven National Laboratory, RP1035. Streeter, V.L., Wylie, E.B., 1967. Hydraulic Transients. Mc Graw Hill Book Co., New York. Vaidyanathan, G. et al., 1994. Dynamic tests related to undercooling events in FBTR. In: Proc. Int. Top. Meet. Sodium Cooled Fast Reactor Safety, vol. 1, Obninsk, pp. 156–165. Vaidyanathan, G. et al., 2010. Dynamic model of fast breeder test reactor. Ann. Nucl. Energy 37 (4), 450–462.
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Dynamic simulation of accidental closure of intermediate heat exchanger isolation valve in a pool type LMFBR K. Natesan ⇑, N. Kasinathan, K. Velusamy, P. Selvaraj, P. Chellapandi, S.C. Chetal Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, India
a r t i c l e
i n f o
Article history: Received 31 March 2010 Received in revised form 18 November 2010 Accepted 7 December 2010 Available online 7 January 2011 Keywords: Fast reactors Plant dynamics Safety analysis PFBR Sleeve valve
a b s t r a c t In a pool type liquid metal cooled fast breeder reactor (LMFBR), core and other internals such as pumps, heat exchangers are immersed in a pool of sodium. Heat exchange from primary sodium circuit (pool) to secondary sodium circuit (loop) is through four intermediate heat exchangers (IHX) immersed in primary sodium pool. Each IHX is provided with a sleeve valve at its primary sodium inlet window for the purpose of isolating the shell side of IHX from the sodium pool. With such a provision, an inadvertent partial or complete closure of a sleeve valve of one of the IHX during normal operation of the reactor has been considered as a design basis event for the reactor. One dimensional transient thermal hydraulic models of the primary and secondary sodium circuits have been developed to study the thermal hydraulic consequences of such an event. The main areas of concern in the plant and the availability of safety parameters for the detection of the event have been evaluated. Ó 2010 Elsevier Ltd. All rights reserved.
1. Introduction Plant dynamic analysis of various events that can happen in a plant is very important for establishing safety of a plant. This study is also important for planning operating procedures and to design various systems and components of the plant to safely withstand the consequences of various anticipated thermal hydraulic transients. Such studies take a leading role in the licensing of a nuclear power plant. Numerical simulation of the dynamic conditions following an initiating event requires special purpose computer codes to be developed specific to the plant design. Restricted analytical models and associated computer codes such as IANUS (Additon et al., 1973) and DEMO (WARD report, 1975) have been developed to simulate the overall response of US LMFBR plants fast flux test facility (FFTF) and clinch river breeder reactor plant (CRBRP) respectively. IANUS code models primary and secondary circuits of the plant with a dump heat exchanger whereas DEMO code models steam generating modules also. Another code, called NALAP (Martin et al., 1975) is also available for the transient simulation of LMFBR developed from the famous code RELAP-3B (RELAP-3B Manual, 1974) which is the system code for water cooled reactors. This code has been developed by substituting sodium properties in place of water in the RELAP 3D code and is capable of predicting rudimentary decay of core flow following a pipe rupture accident. A more generalised code SSC for both loop and pool type LMFBR has been developed by Brookhaven National ⇑ Corresponding author. Tel.: +91 044 27480500/22109; fax: +91 044 27480104. E-mail address: [email protected] (K. Natesan). 0306-4549/$ - see front matter Ó 2010 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2010.12.005
Laboratory (Guppy, 1983). All these studies reflect the importance of dynamic simulation in the successful design of a nuclear power plant. This is paper discusses the thermal hydraulic modules developed for studying the consequences of one of the postulated initiating events in the 500 MWe Indian prototype fast breeder reactor (PFBR), currently under construction. 1.1. Description of the plant The reactor core of a LMFBR where heat is generated by controlled nuclear fission reaction is immersed inside a large pool of sodium named primary sodium circuit along with other components such as pumps and heat exchangers. Sodium pool and primary circuit components are contained inside a large vessel, known as main vessel. The heat produced in the primary sodium circuit is transferred to an intermediate sodium circuit from where it is transferred to steam water system to produce superheated steam in a steam generator. Schematic layout of the heat transport path of LMFBR is shown in Fig. 1. The primary sodium pool is divided into two parts viz. hot pool and cold pool by an internal structure known as inner vessel. Sodium from the cold pool is circulated through the core by two centrifugal pumps operating in parallel. A schematic of primary sodium circuit is shown in Fig. 2. Sodium from hot pool flows through four intermediate heat exchangers (IHX) to cold pool. IHX transfer heat produced by the core to secondary sodium loop. Primary sodium flows from top to bottom in the shell side and the secondary sodium flows through the tubes from bottom to top. The four IHX are connected to the two secondary sodium loops with two IHX in each loop.
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Nomenclature CT HP I IS IE IO IPL PDr PE PP PPS Pi RI SDr SG SP SPE ST c
List of symbols A flow inertial coefficient, m1 D pump developed head, Pa I moment of inertia, kg m2 K pressure drop coefficient, Pa/(kg s1)2 P pressure, Pa Q mass flow rate, kg s1 S area of cross section, m2 Z elevation, m g acceleration due to gravity, m s2 n number of SA in a radial group t time, s a normalised pump speed b torque, N m q density, kg m3 h temperature x pump speed, rad/s Subscripts CE core inlet CP cold pool
Schematic flow sheet of primary and secondary sodium circuits are shown in Figs. 3 and 4 respectively. More details of the plant layout are available elsewhere (Chetal et al., 1995). 1.2. Event classification Each IHX is provided with a sleeve valve at its primary sodium inlet window. Purpose of the sleeve valves is to isolate the shell side of the IHX from primary sodium circuit during an anticipated single secondary loop operation of the reactor. With such a provision, an inadvertent partial or complete closure of a sleeve valve of one of the four IHX during normal operation has been considered as a design basis event (DBE) for the reactor. Various DBE that can occur in the plant are classified into four categories based on the frequency of their occurrence, namely, Category 2 (>102/ry), Category 3 (6102/ry but >104/ry) and Category 4 (6104/ry but >106/ry) with Category 1 being all the planned operations. ‘ry’
core top hot pool intermediate heat exchanger (IHX) IHX secondary side IHX inlet IHX outlet inlet plenum primary pump drive primary pump inlet primary pump primary pump suction primary pipe reactor inlet secondary pump drive steam generator secondary pump secondary pump entry surge tank core channel
stands for reactor operation year. General approach that is adopted in the categorisation is either probabilistic or deterministic. Sometimes a combined or balanced approach is also adopted. Categorisation of various events is important to design various plant systems such that the consequences of the event are confined within pre-specified limits. The limiting parameters are arrived at based on the operating life of various systems with tighter limits specified for the more frequently occurring events. IHX sleeve design is not present in many reactors. The event of accidental closure of sleeve has not been reported to occur in any reactors. Thus, this event is a very rare event and has been classified as Category 3 event by deterministic approach. 1.3. Main concerns following the event In case of a sleeve valve closure in one of the IHX, viz. IHX-1, the total sodium flow pumped by the primary pumps has to flow
Fig. 1. Heat transport path of an LMFBR.
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Fig. 2. Schematic of primary sodium circuit.
through the available opening in IHX-1 and through the other three IHX viz. IHX-2, IHX-3 and IHX-4. IHX-1 primary flow comes down to leak flow value and the other three IHX flows correspondingly increase. Since there are no corresponding changes in the secondary sodium flows through IHX, the IHX-1 primary outlet temperature falls to a low value (close to secondary sodium inlet temperature) and the other three IHX primary sodium outlet temperatures increase. However, since the temperature of majority of the sodium inflow to the cold pool increase, cold pool sodium temperature (hRI) also increases. Since there is automatic plant protection action of reactor trip (SCRAM) based on 10 K increase in cold
pool sodium temperature from its nominal value, reactor trip may happen. The possible reduction in the core flow would be very small due to the small pressure drop offered by the IHX shell side (1.5 mlc) compared to the total primary sodium circuit pressure drop (75 mlc). Even though there is an increase in the reactor inlet temperature, the negative reactivity feedback due to it causes the reactor power to fall. The combined result of these two opposing effects prevents the excessive rise of core assembly temperatures. The reduction in primary sodium flow through IHX-1 from hot pool to cold pool causes increase in the primary sodium flow through IHX2-4. Therefore, there will be an increase in the hot pool
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Fig. 3. Schematic model of primary sodium circuit.
Fig. 4. Schematic model of a secondary sodium circuit.
sodium level and decrease in the cold pool sodium level in order to sustain the increased pressure drop across IHX. This may be a concern for the buckling of inner vessel and sodium overflow from the hot pool through the inner vessel top. The increased primary sodium flow through IHX-2 and IHX-3 may also cause the flow induced vibration of tubes. Near the IHX top tube sheet-shell weld junction, the outside of the shell will be at a temperature corresponding to hot pool sodium temperature. Since, there is only very small primary sodium leak flow through the IHX-1 whose sleeve valve is closed, it may not wash the bottom of the top tube sheet properly to maintain it hot. Hence, the top tube sheet will be at a temperature close to that of secondary sodium outlet. Because of the reduction in the primary sodium flow, the secondary sodium outlet temperature of IHX-1 falls sharply. Hence, a large DT (temperature difference) will be caused across the top tube sheet-shell weld of IHX-1 (DTIHXSh). The increase in the primary sodium flow through IHX-2 and IHX-3 causes the primary sodium temperature at their outlets to increase. This results in the increase of DT across the cold end tube sheet (DTIHXCE). The reduction in IHX-1 primary sodium flow causes the secondary sodium outlet temperature of IHX-1 to
decrease. The increase in primary sodium flow through the IHX-2 causes the secondary outlet temperature of IHX-2 to increase. Hence, the surge tank of loop-1 will be supplied by sodium through two pipes having different temperatures (DTSuT). This is of concern for the surge tank from structural integrity considerations. The net reduction in the primary sodium flow through the IHX of loop-1 causes the corresponding loop SG sodium inlet temperature and hence the steam temperature to decrease. But, steam temperature of loop-2 increases due to the increased primary sodium flow through its IHX, which results in large DT between the steam produced in two loops (DTSTM). As explained earlier, IHX-1 primary sodium outlet temperature falls and that of other IHX rises during the event. This can cause circumferential DT on the main vessel, inner vessel and other structures in the cold pool. However, IHX-1 primary sodium flow being very small compared to that through other IHX, the circumferential DT may not be of much concern. This paper brings out the details of the mathematical model of primary and secondary sodium circuit hydraulics to quantify the parameters of concern explained in the above paragraphs during an instantaneous closure of one IHX sleeve valve.
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2. Mathematical model The governing equations for the primary sodium circuit hydraulics are obtained by making momentum balances between IHX inlet and IHX outlet, pump inlet and core inlet and core inlet and core outlet, by mass balances for the hot and cold pools and by torque balance on the primary pump shaft. Similarly, the governing equations for secondary sodium circuit hydraulics are obtained by making momentum balances between surge tank outlet to pump tank inlet through the SG modules and pump tank inlet to surge tank inlet through secondary sodium pump (SSP) and IHX, by mass balance for surge tank and by torque balance on secondary pump shaft. Specific features of the model are as follows. Sodium flow is considered as one dimensional and incompressible. Head developed and impeller-consumed torque by the pump are obtained from homologous pump characteristics available from Streeter and Wylie (1967). Friction factors and hence the friction pressure drop coefficients in pipes and rod bundles (Novendstern, 1972) under turbulent flow conditions, are proportional to Re0.25, (Re = Reynolds number). All the pressure drop coefficients in the primary and secondary hydraulics model are considered to vary with flow by this relationship. Further, even under low flow conditions in the primary and secondary sodium circuits, the flows are turbulent. Governing equations used in the analysis are given below: 2.1. Primary sodium circuit hydraulic model The governing equations of the primary sodium circuit hydraulics model are: Momentum balance between IHX inlet and outlet
AI
dQ Il ¼ ðZ HP Z IE ÞqHP g ðZ CP Z IO ÞqCP g K Il Q Il jQ Il j dt Z ZIO ql dZ for l ¼ 1 to 4 g
where f1 and f2 are the functions evaluating the homologous characteristics. The empirical relation used for the friction in the pump drive system bPP (fri) other than the impeller is:
14:77a; jaj 0:005 bPP ðfriÞ ¼ 0:117 8:797a; 0:005 < jaj 0:0117 bPP ðtotÞ 0:023a þ 0:012; jaj > 0:0117 Mass balance for core flow 10 X
nr Q cr ¼
r¼1
KIl is the resistance of the path through the IHX whose sleeve valve is closed. Momentum balance between pump inlet and outlet
dQ PPl ¼ ðZ CP Z IO ÞqCP g ðZ HP Z CT ÞqHP g DPcore dt K P Q PPl jQ PPl j þ DPPl for l ¼ 1; 2
ð2Þ
where KP = KPPS + KPi + KIPL. Momentum balance for individual core zone between core inlet and outlet
Acr
dQ cr ¼ DPcore K cr Q cr jQ cr j g dt
Z
Z CT
qr dZ; for r ¼ 1 to 10 ð3Þ
Z CE
Mass balance for hot pool
" # " # 2 4 X X dZ qHP SHP HP ¼ ðQ PPl Þ ðQ Il Þ dt l¼1 l¼1
ð8Þ
DPcore
" # P dQ RIj þ 10 r¼1 nr ar br ¼ P10 r¼1 nr ar
where dt = tj tj1
ð4Þ
dQ RIj
¼
" 2 X
ð9Þ for r ¼ 1 to 10
#
Q PPlj Q RIj1
l¼1
1 ar ¼ Acr þ K cr jQ crj1 j dt br ¼ K cr jQ crj1 jQ crj1 þ dr Z ZCT dr ¼ g qr dh
and
Z CE
2.2. Secondary sodium circuit hydraulics model The governing equations of the secondary sodium circuit hydraulics model of each loop are: Momentum balance between pump entry and surge tank
AIS
dQ ISl ¼ PSPE P ST þ DSP K ISl Q ISl jQ ISl j g dt
Z
Z ST
Z SPE
ql dZ
for l ¼ 1 to 2
ð10Þ
Momentum balance between surge tank and pump entry
" # " # 4 2 X X dZ qCP SCP CP ¼ ðQ Il Þ ðQ PPl Þ dt l¼1 l¼1
ð5Þ
Torque balance for pump
ASG
dQ SGl ¼ PST PSPE K SGl Q SGl jQ SGl j g dt
for l ¼ 1 to 4
dxPPl ¼ bPDrl bPPl ; for l ¼ 1; 2 dt DPPl ¼ f 1ðQ PPl ; xPPl Þ; for l ¼ 1; 2
Z
Z SPE
Z ST
ql dZ ð11Þ
Mass balance between flow from pump and flow from surge tank
IPP
¼ f 2ðQ PPl ; xPPl Þ þ bPPl ðfriÞ
Q PPl
Numerical solution of the primary circuit hydraulics is obtained in two stages. In the first stage, IHX flows, pump flows, pump speeds and the sodium levels are calculated utilizing a standard ordinary differential equation solver based on the Hamming’s Predictor–Corrector method. Using the total core flow obtained from the first stage as input for the second stage, core zone equations are solved. For this a semi-implicit finite differencing and linearisation technique (Agarwal et al., 1977) is applied on the governing equations for individual core zone flows. Also by utilizing the fact that change in total core flow (dQRI) computed in the first stage described above to be equal to the sum of the changes in the individual zone flows, the current time individual zone flows are obtained as follows:
Mass balance for cold pool
bPPl ðtotÞ ¼ bPPl ðimpellerÞ þ bPPl ðfriÞ;
!
l¼1
Z IE
APPl
2 X
Q crj ¼ Q crj1 þ DPcore ar ar br ; ð1Þ
ð7Þ
ð6Þ for l ¼ 1; 2
2 4 X X ðQ ISl Þ ¼ ðQ SGl Þ l¼1
l¼1
ð12Þ
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Mass balance for surge tank
qST SST
" # " # 2 4 X X dZ ST ðQ ISl Þ ðQ SGl Þ ¼ dt l¼1 l¼1
ð13Þ
Torque balance for secondary sodium pump
dxSP ¼ bsDr bSP dt DSP ¼ f 1ðQ SP ; xSP Þ
ISP
ð14Þ
bSP ðtotÞ ¼ bSP ðimpellerÞ þ bSP ðfriÞ ¼ f 2ðQ SP ; xSP Þ þ bSP ðfriÞ where f1 and f2 are similar functions as that of PSP as explained above. This set of simultaneous equations is also solved by Hammigs Predictor Corrector method to obtain the evolution of secondary sodium flow. 2.3. Overall thermal–hydraulics model The hydraulic models for predicting the flow evolution in primary and secondary sodium circuits are interfaced with thermal models of various segments of primary and secondary sodium circuits, viz., core, hot pool, cold pool, IHX, secondary sodium piping and steam generator in the plant dynamics code DYANA-P. This code also contains neutronic models for predicting the fission power evolution in the core considering various reactivity feedback effects. Flow evolutions obtained as above are used as input for estimating the temperature evolutions in various parts of the plant using the other modules of the DYANA-P code. Thermal modeling adopted for core, IHX, piping and steam generators are similar to those adopted in the DYNAM code developed for analysing the plant dynamics of fast breeder test reactor (FBTR) under operation in India. FBTR is a 40 MWt loop type reactor, whereas PFBR is of pool type. Hydraulic models used in the simulation of pool hydraulics of PFBR are elaborated in the previous section. Thermal models are similar in both the types of reactors. Detailed description of thermal models adopted in the DYNAM code is given in Vaidyanathan et al. (2010). These methodologies adopted in the formulation of the DYANA-P code have been validated through the commissioning tests carried out in FBTR (Vaidyanathan et al., 1994).
Fig. 5. Evolution of IHX primary flow.
temperature). The primary and secondary sodium outlet temperatures of other three IHX increase. However, since the temperature of majority of the sodium inflow to cold pool increase, reactor inlet temperature (hRI) increases as seen in Fig. 6. Negative reactivity feedback due to the increase in the reactor inlet temperature causes the reactor power to fall. PFBR operates with manual power control procedure due to the fact that burnup compensation requirement is very less compared to that in thermal reactors. Since, power transient lasts only for a few minutes the operator is not expected to intervene and compensate for fall in reactor power within this short time. Operator identifies the occurrence of the event by alarms before he compensates the power. Increase in reactor inlet temperature and reduction of reactor power are two opposing effects on the fuel elements that prevent excessive rise of fuel clad and coolant temperatures. hRI increases by 10 K above nominal at 62 s and automatic reactor
3. Results and discussion Sleeve valves are large in size and a definite closure time is involved in the valve closure process. However, for the purpose of conservative estimation of thermal hydraulic consequences of the event, valves are considered to close instantaneously. A primary sodium leak flow rate of 1% has been considered through the IHX in which the sleeve valve is closed. In the subsequent discussion this IHX is termed as ‘IHX-1’, while the other IHX of the same loop is termed as ‘IHX-2’ (both in ‘loop-1’). The IHX of the second loop (‘loop-2’) are referred as ‘IHX-3’ and ‘IHX-4’. During this event, the total sodium flow pumped by the primary pumps has to flow through the available opening in IHX-1 and through the other three IHX. IHX-1 primary flow comes down to leak flow value and the other three IHX primary flows correspondingly increase to 132% as shown in Fig. 5. The increased primary sodium flow through IHX-2 to IHX-4 may cause flow induced vibration of the tubes. Reduction in the total core flow is negligible due to the small resistance offered by the IHX shell side compared to the total primary sodium circuit hydraulic resistance. Since there are no corresponding changes in the secondary sodium flows, IHX-1 primary and secondary sodium outlet temperatures fall to low values (close to secondary sodium inlet
Fig. 6. Evolutions of reactor inlet temperature and power.
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trip action (SCRAM) is initiated by it (safety action) due to which, the reactor power falls steeply as seen in Fig. 6. Primary and secondary pumps coast down to 20%, following SCRAM. Secondary sodium temperature at the outlet of IHX responds faster compared to reactor inlet temperature during this event as inertial effects do not attenuate it much. This parameter decreases at the outlet of affected IHX and increases at the outlet of unaffected ones. Decrease in this parameter can be considered as a SCRAM parameter for quick protection during this event. However, in the case of events causing loss of cooling to IHX, secondary sodium temperature at the outlet of IHX increases. Thus, this parameter is effective only for IHX sleeve valve closure event. Considering this parameter as an additional SCRAM parameter will lead to increase in the total number of SCRAM parameters and thereby affecting the plant availability. However, as already discussed there is no concern over core safety during this event and hence quick automatic safety action is not essential for this event. Automatic detection and manual SCRAM are sufficient for this event. Reactor inlet temperature which is mainly envisaged as a SCRAM parameter for protection against loss of cooling in the secondary and tertiary circuits is definite and reliable alarm parameter for this event (threshold at 5 K above nominal). However, in order to improve the reliability of event detection based on difference between the secondary sodium outlet temperatures of IHX in each loop is added as an additional alarm parameter in the plant.
falls sharply. Hence, the top tube sheet-shell weld of IHX-1 is subjected to a large temperature difference (DTIHXSh) as shown in Fig. 8. Increase in the primary sodium flow through IHX-2 to IHX-4 causes primary sodium temperature at their outlets to increase. This in turn results in increase in temperature difference across the cold end tube sheet (DTIHXCE) as shown in Fig. 9. Reduction in the IHX-1 primary sodium flow causes the secondary sodium outlet temperature of IHX-1 to decrease. Increase in the primary sodium flow through the other IHX causes the secondary outlet temperature in them to increase. Hence, the surge tank of
3.1. Consequences on structures The increase in the primary sodium flow through IHX-2 to IHX4 causes an increase in the hot pool sodium level and decrease in the cold pool sodium level in order to sustain the increased pressure drop across IHX for the increased primary sodium flow. The evolutions of sodium free levels in hot and cold pools are shown in Fig. 7. The maximum level difference reached during the event does not cross the limiting value of 4.5 m and hence there is no risk of buckling of inner vessel. Because of the reduction in the primary sodium flow, the secondary sodium outlet temperature of IHX-1
Fig. 8. Evolution of DTIHXSh of IHX-1.
Fig. 7. Evolutions of sodium free levels.
Fig. 9. Evolution of DTIHXCE of IHX.
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Fig. 10. Evolution of DTSuT.
Fig. 11. Evolution of DTSTM.
loop-1 will be supplied by sodium through two pipes having different temperatures (DTSuT). The evolution of DTSuT is shown in Fig. 10. Net reduction in primary sodium flow through of loop-1 IHX causes the corresponding loop SG sodium inlet temperature and hence the steam temperature to decrease. Whereas, loop-2 steam temperature increases due to the increased primary sodium flow through its IHX. This results in large temperature difference between the steam produced in two loops (DTSTM) as shown in Fig. 11. The evolutions of IHX primary outlet temperatures of IHX-1 and IHX-2 are shown in Fig. 12. It can be seen that there is a maximum difference of 108 K between the primary sodium outlet temperatures of IHX-1 and IHX-2. Primary outlet temperature evolutions of IHX-3 and IHX-4 are very close to that of IHX-2. As far as cold pool is concerned, only 1% of sodium flow (IHX-1 primary flow) is coming at low temperature and the rest 99% comes at high temperature. Hence, even though the temperature difference between the streams is high, this will not cause any considerable circumferential temperature variation on the structures in cold pool. The scenario following the event is summarised in Fig. 13. 3.2. Impact of reactor SCRAM and event detection Maximum values of various parameters reached during the event with and without reactor SCRAM action are shown in Table 1. It can be seen that reactor trip is not helpful in reducing the maximum values of sodium level and DTs. In fact, the maximum values of DTs on various structural components increase due to SCRAM. This is due to the differences in flow coasting of primary and secondary sodium pumps following SCRAM. However, level difference between the sodium pools reduces sharply following reactor trip and component DTs start decreasing very slowly after about 100 s. The occurrence of complete closure of one of the four IHX sleeves valve would be indicated by a reduction in the secondary outlet temperature of IHX-1 and increase in the same of IHX-2 to IHX-4 resulting in the evolution of DTISO. This parameter is added
Fig. 12. Evolution of IHX primary outlet temperatures.
to the list of automatic alarm parameters of the plant to improve the reliability of event detection. Apart from this, there will be alarm available from the reactor inlet temperature monitoring system when its value increases by 5 K above nominal. The magnitude to which DTISO rises during the event would give an indication of the extent of valve closure. The increase in IHX cold end DT (DTIHXCE) of IHX-2 to IHX-4 with primary and secondary sodium flows remaining unchanged is also an another unique feature that would happen during the event.
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Fig. 13. Scenario following the event.
Table 1 Maximum values of various parameters reached during the event. Plant condition
Normal operation Max. values reached without safety actions Max. values reached with safety actions
Level difference between hot and cold pools, m
DTIHXSh, DTIHXCE, DTSuT, DTSTM, K K K K
1.5 2.498
38 220
42 75
0 195
0 97
2.498
211
108
201
100
4. Conclusions Detailed investigations of sleeve valve closure event have been carried out and the following conclusions have been arrived at: 1. There is no concern over fuel, clad and primary sodium temperatures even without any safety actions. But there are some implications on the structural integrity of components. 2. IHX tubes may be subjected to flow induced vibration (FIV) risk due to increased primary sodium flow. 3. IHX, surge tank and steam header are to be protected from the thermal loading caused by the event.
4. Level difference between hot and cold pools during the event does not lead to spill over condition for which inner vessel buckling design has been made. The study provided thermal loading on various components and the components have been designed for these loadings. References Additon, S.L., McCall, T.B., Wolfe, C.F., 1973. IANUS – Outline Description, Westingbouse Advanced Reactors Division, Waltz Mill, Pennsulvania, FPC-939. Agarwal, A.K. et al., 1977. Simulation of transients in LMFBR systems. Nucl. Sci. Eng. 64, 480–491. Chetal, S.C. et al., 1995. Conceptual design of heat transport system and components of PFBR – NSSS. In: Proc. Conceptual Design of Advanced Fast Reactors, IAEA Tech. Committee Meeting, IAEA-TECDOC-907, Kalpakkam, pp. 117–132. DEMO, 1975. LMFBR Demonstration Plant Simulation Model, DEMO, Westinghouse Advanced Reactors Division, WARD-D-0005. Guppy, J.G., 1983. Super System Code [SSC] an Advanced Thermohydraulic Simulation Code for Transients in LMFBRS, NUREG/CR-3169. Martin, B.A., Agrawal, A.K., Albright, D.C., Epel, L.G., Maise, G., 1975. NALAP: an LMFBR System Transient Code, BNL-50457. Novendstern, E.A., 1972. Turbulent flow pressure drop model for fuel rod assemblies utilising a helical wire-wrap space system. Nucl. Eng. Des. 22, 19–27. RELAP-3B, 1974. RELAP-3B Manual a Reactor System Transient Code, Brookhaven National Laboratory, RP1035. Streeter, V.L., Wylie, E.B., 1967. Hydraulic Transients. Mc Graw Hill Book Co., New York. Vaidyanathan, G. et al., 1994. Dynamic tests related to undercooling events in FBTR. In: Proc. Int. Top. Meet. Sodium Cooled Fast Reactor Safety, vol. 1, Obninsk, pp. 156–165. Vaidyanathan, G. et al., 2010. Dynamic model of fast breeder test reactor. Ann. Nucl. Energy 37 (4), 450–462.