Unplanned Shutdown Frequency Prediction of FBR MONJU Using Fault Tree Analysis Method

Transcription

1 Journal of Energy and Power Engineering 8 (2014) D DVID PULISHING Unplanned Shutdown Frequency Prediction of FR ONJU Using Fault Tree nalysis ethod asutake SOTSU Plant Dynamics nalysis Group, FR Safety Technology Center, Japan ntonic Energy gency, 1, Shiraki, Tsuruga, Fukui , Japan Received: February 06, 2014 / ccepted: arch 11, 2014 / Published: July 31, bstract: In order to evaluate the operational reliability of Japanese FR (fast breeder reactor) ONJU, frequencies of important intermediate events and equipment failures resulting during reactor automatic trip are predicted using FT (fault tree analysis) technique for the plant model. The targeted devices are the following: (primary heat transport ), SHTS (secondary heat transport ), WS (water and steam ), PPS (plant protection ) and PCS (plant control ). In this paper, the frequency of automatic reactor trips was estimated by extracting and analyzing the important intermediate events and equipment failures covering all the derived fault trees of these s. The analyses predicted 1.2/RY (reactor year) the value of unplanned shut down frequency by the internal factor of the. The largest contributed event was of SHTS accounting for 42.6% of total events followed by with 40.1%. The contribution factor of WS was only 4.4%. Key words: FR, ONJU, fault tree analysis. 1. Introduction The FT (fault tree analysis) of the cooling with a reactor trip as the top event has been arranged in ONJU to estimate unplanned shut down frequency during rated power operation and also to indicate effective strategy improvement by reference to previous studies [1-5]. This arrangement, after being triggered by an unplanned shut down due to a WS (water and steam ) trouble occurred in 1995, has been conducted by applying FT to the WS to quantitatively evaluate the operational reliability of equipment and also to review the actual operation and maintenance procedures of equipment with high contribution ratios to reactor trip frequency, aiming to assure enhanced operation reliability. ased on experience gained during sodium leakage in ONJU SHTS (secondary heat transport Corresponding author: asutake SOTSU, senior research engineer, research fields: probabilistic risk assessment, advancement of plant operation, maintenance of fast breeder reactor. sotsu.masutake@jaea.go.jp. ) in the same year, some SHTS equipment will be modified. nother FT of the SHTS was then performed to confirm that this modification has no adverse influence on the reactor trip frequency. fterwards, to evaluate other large contribution to the reactor trip, FTs of the (primary heat transport ), PPS (plant protection ), and PCS (plant control ) were performed in series, and the frequency assessment of the reactor trip over the entire plant was completed. This paper describes the analysis result of all FTs (fault trees) developed until now. 2. Outline of ONJU ONJU is a sodium-cooled, loop-type prototype fast breeder reactor which can supply 280 W of electricity. The heat generated in the reactor core is removed by three loops of each of them thermally connected through individual IHX (intermediate heat exchanger) to another coolant circulation loop of SHTS. The turbine generator is driven by steam generated at three

2 Unplanned Shutdown Frequency Prediction of FR ONJU Using Fault Tree nalysis ethod 1287 Primary circulation CS SH T/G EV (R/V) (IHX) Secondary Circulation SHTS WS Fig. 1 System diagram of ONJU. EVs (evaporators) and SHs (super heaters) installed at the SHTS as shown in Fig Evaluation rea Table 1 shows reactor automatic trip signal in ONJU. External events, i.e., earthquake acceleration large signal and electric bus voltage low signal were not considered in the FT. The turbine generator equipment of ONJU is similar to fossil electric power plants and consequently the turbine trip event was not developed to a FT. In addition, because the rated power operation is targeted, the neutron flux high signal generated during the plant starts up is not considered in this evaluation. Regarding the turbine trip signal and the electric bus voltage low signal, their frequency was presumed from the experience of Japanese LWR (light water reactor) in this evaluation. The evaluation conditions in this study are as follows: (1) Failure rates of the dynamic mechanical components ( and motor driven valve, etc.) and electric components are taken into account; relatively Table 1 Reactor trip signal of ONJU. Reactor trip signal Source range neutron flux high Wide range neutron flux high Power range neutron flux high Power range neutron flux variation high R/V sodium level low Sodium temperature high at R/V outlet Sodium temperature high at IHX outlet rpm low rpm high flow rate low SHTS rpm low SHTS flow rate low Sodium temperature high at EV outlet Turbine trip Electric bus voltage low Fuel failure detection CV isolation Horizontal acceleration Earthquake acceleration large Vertical acceleration anual trip smaller rates of failures of the static components, the manual valve and the leakage (heat exchanger tube damage, coolant leakage from the main piping, etc.) are neglected. (2) Reactor trip generated by double failures is targeted.

3 1288 Unplanned Shutdown Frequency Prediction of FR ONJU Using Fault Tree nalysis ethod (3) The plant is assumed to be in rated power (3) aintaining function of coolant sodium level. operation, and the mission time is one year (8,760 h). To remove the heat from the reactor core with (4) Leakage of internal fluid, human error and liquid sodium the sodium liquid level should be kept common cause failure are not considered. appropriate. (5) Data of the failure rate of the equipment were (4) Heat exchange and removal function. introduced from the fast reactor equipment database The intermediate heat exchanger is used for both CREDO/CORDS [6, 7] and Japanese LWR database heat removal from the reactor core and heat transfer to NSR [8] (1997). the SHTS. (5) aintaining function of dynamic mechanical 4. Development of FT component of. In the FT construction first were extracted the To make sodium circulate, the primary main intermediate events leading to a reactor trip, considering circulation is needed. specific functions to the target, and then the (6) Protection function of the. intermediate events were developed down to component The reactor protection function to deal with an failures. abnormal transient during rated power operation is needed. The reactor trip due to the malfunction of 4.1 FT of these protection functions is considered. The reactor trip that originates in equipment failure From the above mentioned functions, 11 intermediate of is defined as the top event, and the events that cause the reactor trip were extracted as intermediate events were extracted from the shown in Fig. 2. function. These events are classified along with the The intermediate events were selected as follows following six functions of the : using a similar procedure. (1) aintaining function of coolant flowrate. In this report, it was assumed that the device The heat, generated in the reactor core, is transported failures can be detected instantaneously with four with sodium to SHTS. values of repair times as listed below: (2) Formation function of coolant flow path. (1) The repair time of the devices in the nitrogen The formation of the coolant flow path from the cell to restrain sodium combustion is assumed to be at reactor core to SHTS is needed. least four months (2,880 h). Fig. 2 Overflow R/V sodium level low Overflow tank L flow rate low F T r gas T Check valve Na Temperature high at R/V outlet Reactor trip on. rpm high/low S ux. cooling G Lubricating oil Reactor trip Na Temperature high at IHX outlet Overflow R/V liquid level r gas low Na Temperature high at R/V outlet rpm high/low -G set Na temperature high at IHX outlet flow Check valve fail rate low close

4 Unplanned Shutdown Frequency Prediction of FR ONJU Using Fault Tree nalysis ethod 1289 (2) The repair time of the devices outside the nitrogen cell is assumed to be 1 week (168 h). (3) The devices partially in the nitrogen cell are handled as equipment that exists in the cell. (4) r-gas related devices are regarded to be in the nitrogen cell. 4.2 FT of SHTS When the FT in the SHTS was made, the intermediate event was selected in the beginning similar to the FT of. Plural functions of SHTS were analyzed, and events to cause loss of these functions were selected as the intermediate events. In total, 21 intermediate events were selected including event potentially leading to reactor trip. 4.3 FT of WS In the WS, ONJU-specific devices, i.e., devices from the deaerator to the SV (main steam stop valve) installed on the inlet piping of the turbine generator were evaluated. The main factor of reactor trip, which directly originates in failure of WS, is turbine trip signal. nother factor is the reactor trip signal caused by the feed water flowrate low signal at EV or by the EV outlet steam temperature high signal. mong the trip triggering anomalies to take place in the EV and the SH, mismatch between feed water and secondary sodium flow rates was the only one evaluated event, because the others such as sodium level low or temperature high were already taken into account in the FT of SHTS. Finally, 18 intermediate events were extracted according to the event sequences as mentioned above. 4.4 FT of PPS Complicated data handling is necessary to build this FT as described below. The reactor trip and the reactor containment vessel isolation signal including engineered safety features by false status signals are evaluated as the intermediate events that originate in the PPS. oreover, the event that originates in the control rod position support was also evaluated as an important event. Thus, 16 intermediate events were extracted. Failures of logical circuits in the PPS were evaluated by the following approach. Operating test intervals are assumed as one month for the 2 out of 3 logical circuits and the reactor trip breaker and as one year for all other equipment. ultiple failures of the logic channel were evaluated in consideration of the rule provided in the safety regulations of ONJU as follows: Continuous operation for six hours is allowed after PPS failure is detected; Reactor has to be manually shutdown within 36 hours when failure is not restored within six hours. The reactor trip frequency Fs, during mission time, due to the malfunction of those equipments was evaluated so that a single malfunction of the 2 out of 3 logical circuits and the breaker may cause the reactor trip directly. Fs = T (1) 2 out of 3 logical circuit consist of channels, and C causes a reactor trip when two channels fail. The event frequency Fm of reactor trip in case of a simultaneous failure of channels and is evaluated by the following method: Fm = T T (2) where, : failure rate of channel (/h); T : mission time (h); : failure rate of channel (/h); : repair time of channel (h); : repair time of channel (h). There are three kinds of combination of the false status signal written by Eq. (2). The frequency of reactor trip as a consequence of another channel malfunction as a single failure was evaluated by applying Eq. (2) with the allowed restoration time of 36 hours after a usual shut down operation.

5 1290 Unplanned Shutdown Frequency Prediction of FR ONJU Using Fault Tree nalysis ethod 4.5 FT of PCS The following devices, related to the heat transport of ONJU among the PCS, were evaluated: (1) power demand master (circuit concerning the, SHTS and WS); (2) the flow rate controller; (3) the SHTS flow rate controller; (4) controller of steam temperature at EV outlet; (5) feed water control valve differential pressure controller. The s of these devices were selected as the intermediate events and each of them was examined whether or not was able to cause a reactor automatic trip with consideration of the following points: manual backup by the operator is not considered; Reactor automatic trip is not assumed to be triggered by a failed interlock which is installed to prevent plant trips in case of malfunctions of power set value or of control circuits; Treble failures are not considered. Table 2 shows the obtained intermediate events. 5. Results If the plant utilization rate of ONJU is assumed to be 80%, unplanned shutdown frequency by the internal factor of the is 1.2/Reactor Year as shown in Table 3. The contribution of WS was not Table 2 Intermediate events for targeted s. -G set Check valve failclose Rise of sodium level Fall of sodium level alfunction of alfunction of lubricating oil False state signal of flow rate low False state signal of rpm high False state signal of rpm low False state signal of primary sodium temperature high at IHX outlet False state signal of primary sodium temperature high at R/V outlet SHTS SHTS rpm high SHTS rpm low ain cooling line valve fail close CS outlet valve fail open SHTS protection circuit activation Rise of SH sodium level Fall of SH sodium level Rise of EV sodium level Fall of EV sodium level SHTS sodium inventory decrease SHTS sodium inventory increase False state signal of SHTS flow rate low False state signal of SHTS rpm high False state signal of SHTS rpm low False state signal of sodium temperature high at EV outlet False state signal of sodium temperature high at SH outlet False state signal of sodium flow rate high at CS outlet standing by False state signal of SG tube leak Failure of SHTS lubricating oil Failure of SHTS thyristor inverter cooling False state signal of SG level detector WS ain feed water protection circuit activation Feed water line valve fail close SH inlet release valve fail open Feed water control valve fail open Feed water line drain valve fail open Feed water line SH in/out let valve fail open ain steam line release valve fail open Feed turbine steam line valve fail close Feed turbine steam control valve fail open alfunction of main turbine extract line for feed water heater Rise of main steam pressure False state signal of EV feed flow rate low False state signal of mismatch of feed water/shts flow rate False state signal of feed water temperature low False state signal of EV outlet steam temperature low alfunction of main steam pressure control SV fail close False state signal of EV outlet steam temperature high PPS Function failure of FCR Function failure of CCR Function failure of CR Events related to PPS breaker Events related to fuel failure detection Event related to wide range neutron flux high signal Event related to power range neutron flux high signal Event related to power range neutron flux variation high signal Event related to R/V sodium level low signal Event related to CV isolation Event related to sodium leak from R/V Event related to R/V sodium level low signal Event related to CV pressure high signal Event related to CV radio activity high signal Event related to CV temperature high signal Event related to sodium leak from, IHX PCS False state signal of power demand master False state signal of flow rate controller False state signal of SHTS flow rate controller False state signal of Controller of steam temperature at EV outlet False state signal offered water control valve differential pressure controller

6 Unplanned Shutdown Frequency Prediction of FR ONJU Using Fault Tree nalysis ethod 1291 large in comparison with the frequency of turbine trip. s shown in Fig. 3, the largest contribution was 42.6% of events that originate in the SHTS and the next was 40.1% of. The events originated in the WS were 4.4% and turbine trip frequency in Table 3 and Fig. 3 is introduced from statistic value of Japanese LWRs. Fig. 4 shows that malfunction of has the largest contribution in the distributions of the intermediate event in. The other important events are: of -G (motor-generator) set, rise of sodium level caused by failure of r-gas compressor and malfunction of lubricating oil. mong the intermediate events of SHTS shown in Fig. 5, the largest contribution is failure of SHTS. The others events are the SHTS sodium inventory in main circuit decrease caused by failure of electromagnetic, other malfunctions and SHTS sodium inventory in main circuit increase. Table 4 shows the events that are more frequent Table 3 Unplanned shutdown frequency. System Frequency (/RY) 0.48 SHTS 0.51 WS 0.05 Turbine trip 0.14 PPS 0.00 PCS 0.01 Total 1.20 Fig. 3 Turbine trip 11.9% WS 4.4% 40.1% Distributions of the. PCS 0.7% PPS 0.2% SHTS 42.6% Fall of sodium level 0.3% Other malfunctions 0.2% Fig. 4 Fig. 5 alfunction of 48.2% Check valve fail close 2.0% -G set function failure 22.4% alfunction of lubricating oil 6.4% Rise of sodium level 20.6% Distributions of the intermediate event in. SHTS sodium inventory in main circuit increase 6.1% Other malfunctions 19.0% CS outlet valve fail open 0.6% ain cooling line valve fail close 0.5% Failure related SHTS 38.2% SHTS sodium inventory in main circuit decrease 35.5% Distributions of the intermediate event in SHTS. Table 4 Intermediate events more than 0.1/RY. (SHTS) SHTS rpm low 0.25 () alfunction of 0.23 (SHTS) SHTS sodium inventory 0.18 decrease () -G set 0.11 () Rise of sodium level 0.10 than 0.1/RY among all of FT results. ll of these events are related to sodium. It is considered that the component failure rates of these s are larger than Japanese LWR s failure rate because of small operating experience of sodium-cooled fast breeder reactors. Reliability improvement of the highly-ranked devices is expected to effectively reduce the unplanned shutdown frequency of ONJU.

7 1292 Unplanned Shutdown Frequency Prediction of FR ONJU Using Fault Tree nalysis ethod 6. Conclusions It was predicted that unplanned shut down frequency, derived from fault tree analysis for the significant equipment of ONJU, is 1.2/RY. The analysis showed that important s are SHTS and regarding the reactor trip and important events in each are the loss of coolant circulation function and the loss of sodium level maintaining function. WS, the nuclear reactor protection, and the plant control were evaluated to have a smaller contribution to the reactor trip. The events of large contribution were mainly related to sodium among all the analyzed s. One major reason of this tendency is that component failure rate of the sodium s with small operating experience is relatively large. The strategy to improve reliability of this equipment will be examined in future. cknowledgments The author would like to recognize the contribution of Yoshiki STOU of NESI for the analytical simulation in this paper and to acknowledge for Shinji YOSHIKW of Japan tomic Energy gency for the editorial suggestions. References [1] USNRC Probabilistic Safety nalysis Procedures Guide. NUREG/CR [2] USNRC Reactor Safety Study, n assessment of ccident Risks in U.S. Commercial Nuclear Power Plants. WSH-1400, NUREG-75/014. [3] USNRC Severe ccident Risks: n ssessment for Five U.S. Nuclear Power Plants. NUREG-1150 vol.1. [4] IE Development and pplication of Level 1 Probabilistic Safety ssessment for Nuclear Power Plants. Specific Safety Guide, No. SSG-3. [5] USNRC Fault Tree Handbook. NUREG [6] Kurisaka, K nalysis of operating experience of LFR components using the CREDO database. Presented at International Topical eeting on Sodium Cooled Fast Reactors Safety, Obninsk, Russia. [7] Kurisaka, K Development of LFR Component Reliability Database. Presented at 4th Japan-Korea PS Workshop, Jeju, Korea. [8] Nuclear Safety Research ssociation. Investigation Concerning Failure Rate Data for PS1997. Japan. (in Japanese)