1、1 NASA TECHNICAL MEMORANDUM - - - - - - NASA TM X-1931 1 i LOSS OF FORGED COOLING FLOW IN THE PLUM BROOK REACTOR I i by H. B. Rarkly, JT. I Lewis Research Center IPIWAUTCS AND SPACE ADYI#lSTWATlO# * WASHDHfiTOW., D. C. DECEYBER 1969 7 , . Provided by IHSNot for ResaleNo reproduction or networking pe
2、rmitted without license from IHS-,-,-LOSS OF FORCED COOLING FLOW IN THE PLUM I BROOK REACTOR 9. Performing Organization Name and Address Lewis Research Center Plum Brook Station National Aeronautics and Space Administration Sandusky, Ohio 44870 3. Recipients Catalog No. I 5. Report Date December 196
3、9 6. Performing Orgonizotion Code I 10. Work Unit No. 13. Type of Report and Period Covered I 12. Sponsoring Agency Name ond Address Technical Memorandum National Aeronautics and Space Administration Washington, D. C. 20546 I 15. Supplementary Notes 16. Abstract The Plum Brook Reactor (PBR) experien
4、ced a temporary loss of forced-cooling flow in November, 1966. To provide a framework for understanding the occurrence, a brief description is given of the PBR, its electrical distribution system, and methods for maintaining flow. Then the occurrence is treated in three parts: causes; inspec- tions
5、and analyses conducted to evaluate the situation; and corrective action taken to avoid a recurrence. It is concluded in this case that an undesired rather than an un- safe condition existed. 18. Distribution Statement Unclassified - unlimited Test reactor operation Test reactor safety Unclassified *
6、For sale by the Clearinghouse for Federal Scientific and Technical Information Springfield, Virginia 22151 Provided by IHSNot for ResaleNo reproduction or networking permitted without license from IHS-,-,-LOSS OF FORCED COOLING FLOW IN THE PLUM BROOK REACTOR* by H. B. Barkley, Jr. Lewis Research Cen
7、ter Plum Brook Station SUMMARY Prevention of the “loss of flow“ and “loss of coolantM accidents continues to have major technical and economic influenc on nuclear power plant design and operation. Realistic assessment of the mechanisms and consequences of the accidents is being given much attention
8、(e. g. , the AEC LOFT facilities and analyses). The Plum Brook Reactor (PBR) experienced a temporary loss of forced-cooling flow. The analyses conducted and conclusions reached after conditions previously analyzed were exceeded may have appli- cation to other plants. On November 22, 1966, following
9、about 7 days of operation at full power of 60 mega- watts (th), the Plum Brook Reactor experienced a temporary loss of forced-cooling flow, initiated by interruption of dc control power to the primary main and shutdown coolant pump breakers. The control power breaker was accidenfally opened. An auto
10、matic pump interlock scram occurred within 1 second after the breaker was opened. Previous hydraulic testing demonstrated that forced flow from coastdown persisted for at least 30 seconds. For ced-cooling flow was restored within an additional 45 seconds. The paper first gives a brief background des
11、cription of the Plum Brook Reactor to provide a framework for understanding the occurrence. Then the causes of the occur- rence are presented, followed by a description of the inspections, analyses, and evalua- tions conducted. The presence or extent of any damage was assessed, and safety for re- st
12、art was assured. Other reviews and evaluations were conducted to determine the pro- per follow-up corrective action. The corrective actions taken to prevent a recurrence are summarized. It is concluded in this case that an undesired rather than an unsafe condition existed. The investigation and acti
13、ons taken have avoided recurrence to date, and it is believed are sufficient to avoid recurrence in the future. *Presented at AMS Topical Meeting on Reactor Operating Experience, San Juan, Puerto Rico, Qet. 1-3, 1969. Provided by IHSNot for ResaleNo reproduction or networking permitted without licen
14、se from IHS-,-,-Considerable attention is properly being given to a realistic assessment of the mechanisms and consequences of the “loss of flow“ accident in nuclear power plants. Lack of complete information leads to designs with rather elaborate and redundant pre- cautions against loss of flow, an
15、d conservative protection against the worst conceivable consequences. Thus, the ?loss of floww accident has considerable practical and econo- mic impact on the design and operation of the nuclear power plant. The Plum Brook Reactor (BBR) experienced a temporary loss of forced-cooling flow, and condi
16、tions which exceeded those previously analyzed were reached. To provide a framework for understanding the occurrence, we start with a brief description of the PBR, its electrical distribution system, and methods for maintaining flow. The paper then treats the occurrence in three parts, (1) causes; (
17、2) inspections, analyses, and evaluations to assess the presence or extent of any damage and safety for restart, and to determine the proper corrective action to be taken; and (3) the corrective actions taken. BACKGROUND DESCRIPTION Reactor Core and Reactor Tank Assembly The Plum Brook Reactor (PBR)
18、 is a 60 megawatt (th) pressurized water test reac- tor. Figure 1 illustrates the layout of the reactor core and instrument and experiment facilities. Within the core box can be seen the 3x9 array of fuel elements, surrounded on three sides by a single row of beryllium tfL-piecew reflectors, and on
19、the fourth side by four rows of beryllium sR-piece?l reflectors. Normal cooling water flow has two upward paths, through the R reflector and past the experiment facilities; and all water passes downward through the L reflector and fuel elements. The fuel elements are of the MTR curved plate type, wi
20、th 20 mils of aluminum cladding on a uranium-aluminum 6 alloy. At the hottest spot in the fuel the nominal heat flux is 1.4X10 Btu per hour per square foot with a corresponding fuel element temperature of 325 F (less than satura- tion temperature). The steady -state departure from nucleate boiling r
21、atio (DNBR) is greater than 2, and the transient DNBR is greater than 1.3. Figure 2 shows the reactor core located in the reactor tank. This provides a per- spective view of the reactor, and the instrument and experiment facilities. The direc- tion of normal flow can also be seen, upward past the ex
22、periment thimbles and through the R reflector, and downward through the L reflector and fuel elements. Provided by IHSNot for ResaleNo reproduction or networking permitted without license from IHS-,-,-3. 114 I. d. HT-1 Fueled core 9 i.d. 3 124 0.d. east-west 10 0. d. I vertical I Figure 1. - Horizon
23、tal section of reactor core. (Dimensions are in inches.) Provided by IHSNot for ResaleNo reproduction or networking permitted without license from IHS-,-,-Fuel. Figure 2. - Reactor tank assembly. Provided by IHSNot for ResaleNo reproduction or networking permitted without license from IHS-,-,-Electr
24、ical Dist rioution System Figure 3 is a simplkfied schematic of the portion of the electrical distribution sys- tem that concerns reactor coolant flow, at the time of the temporary loss of forced- cooling flow. (Some changes “cat were unrelated. to the occurrence have subsequently been made. ) Comme
25、rcial power comes in to two buses through the “North Linet and the “Seuth Line. ? These two buses are normally separated, but automatically tie to- gether on loss of either the North or the South Line. The Emergency Power bus is normally fed through one of two breakers, from either the North or the
26、South Commer- cial Power bus. A key interlock prevents both breakers from being closed simultaneously and tying the North and South Lines. Diesel generators run partly loaded on the Emer- gency Power bus in parallel with commercial power. On loss of commercial power, the Emergency Power bus is autom
27、atically separated from the Commercial Power bus, and the Diesels power the necessary loads. The Guaranteed Power bus is a dc bus. It is normally fed through inverters-diverters from the Emergency Power bus. Additionally, however, batteries power this bus as a backup for the unlikely loss of emergen
28、cy power. South line North line Shutdown Shutdown I Diesels A A 3n.c. n. o. 1;n.c coolant coolant Inverters pump PUMP Diverters A Guaranteed power (dc) Commercial - - Control Emergency se - power cooling power (ac) a Main 1 Key T lk Main Main b coolant coolant coolant pump I)iotsp pump pump valves F
29、igure 3. - Simplified electrical distribution system. (n.c., normally closed; n.o., normally open. Provided by IHSNot for ResaleNo reproduction or networking permitted without license from IHS-,-,-Flow Systems The PBR Primary Cooling Water System has two flow loops, a main and a shutdown. The primar
30、y main cooling water loop has three pumps, with two sufficient to provide full flow of 17 000 gpm for full reactor power. Two main coolant pumps are powered from the North Commercial Power bus, and one from the South Commercial Power bus. A normal operating mode is to run two pumps, one from each bu
31、s. The third pump is in standby and automatically starts if either of the running pumps is lost. The primary shutdown cooling water loop has two pumps, with either capable of providing sufficient flow (I000 gpm) for cooling after a scram from prolonged full power operation. The shutdown coolant pump
32、s are powered from the Emergency Power bus. One pump is run whenever the reactor is operating, and the other pump automatically starts if the first pump is lost. Control power for both the main and the shutdown coolant pumps comes from the dc Guaranteed Power bus. A third source of flow, emergency f
33、low, is available for the extremely unlikely loss of both main and shutdown cooling flow. This system provides water by gravity feed from an overhead storage tank through the reflector and core and out a drain to hot re- tention tanks. This system is activated by three buttons in the control room wh
34、ich actuate motor operated valves powered from the dc Guaranteed Power bus. As mentioned above, the normal reactor operating mode is with two main coolant pumps and one shutdown coolant pump running. The reactor has the customary scrams from low flow and core AP. Additionally, anticipatory protectio
35、n is provided for the reactor for off -normal flow conditions. If one main coolant pump is temporarily lost for any reason, all the control rods are automatically inserted at a fast rate. If both main coolant pumps are lost, a pump interlock reactor scram is actuated in just less than 1 second. The
36、fast rod insertion and the 1 second delay of the scram avoid un- necessary reactor shutdowns for transients or conditions that can be quickly corrected by automatic switching (e. g., tie breaker closure or start of a standby pump). Thus, if for any reason all main cooling flow is lost, the reactor s
37、crams; and shutdown cool- ing flow is provided by a shutdown coolant pump. In the unlikely event shutdown cooling flow is also lost, the reactor operator institutes emergency flow by pressing the three buttons in the control room. Since previous hydraulic testing demonstrated that forced flow from c
38、oastdown persists for at least 30 seconds, the operator has 30 seconds after a scram to verify the presence of flow or to actuate emergency cooling. LOSS OF FORCED COOLING FLOW On November 22, 1966, following about 7 days of operation at full power of 60 mega- watts (th), the Plum Brook Reactor expe
39、rienced a temporary loss of forced cooling flow, Provided by IHSNot for ResaleNo reproduction or networking permitted without license from IHS-,-,-initiated by interruption of dc control power to the primary main and shutdown coolant pump breakers. The control power breaker was accidentally opened.
40、An automatic pump interlock scram occurred within 1 second after the breaker was opened. Previous hydraulic testing demonstrated that forced flow from coastdown persisted for at least 30 seconds. Forced cooling flow was restored within an additional 45 seconds. Causes This potentially serious occurr
41、ence followed the classic pattern, where the conse- quences of an improbable event are compounded by multiple additional errors. The occurrence was caused by a combination of design errors and operator errors. There were three design errors. (1) Too many vital control power circuits on one breaker.
42、Dc control power to all of the main and shutdown coolant pump breakers orig- inated from one breaker on the Guaranteed Power bus. (2) Improper breaker action on loss and restoration of breaker control power. Although the control power breaker was reclosed within a few seconds after its accidental op
43、ening, restoration of the control power did not restart any pumps. The reactor operator had to restart pumps by pres- sing buttons in the control room. (3) Inadequate physical protection for the breaker handles on some vital breakers. This single vital control power breaker handle had no physical pr
44、otection covering it to prevent inadvertent operation. Three operator errors were evidenced. (1) Carelessness by the electrical operator who inadvertently tripped the breaker. (2) Failure of the electrical operator to report the condition to the reactor control room, even though he immediately reclo
45、sed the breaker. (3) Failure of the reactor operators to recognize immediately that a loss of flow condition had occurred, and therefore failure to take immediate proper emergency action. It should be noted that although operator error contributed to the occurrence, it was good operator performance
46、that restored conditions to normal and prevented the occurrence from lasting any longer than it did. Inspections, Analyses, and Evaluations The inspections, analyses, and evaluations of the occurrence can be divided into two phases, (1) the immediate, necessary to assess the presence or extent of an
47、y damage, and the safety for restart, and (2) the subsequent or follow-up, to determine all necessary corrective action to prevent recurrence and any other important corollary changes or actions suggested by the occurrence. Provided by IHSNot for ResaleNo reproduction or networking permitted without
48、 license from IHS-,-,-The first phase of the evaluation included the following. It was first determined that there was no increase in the primary cooling water fission product activity follow- ing the Occurrence. To confirm that no damage had occurred, we performed significant analyses and inspectio
49、ns to systematically consider all of the reactor core components, and instruments and experiments (see figs. 1 and 2). The exact model for convective core cooling was not known. Therefore, we calculated the minimum departure from nucleate boiling ratio (DNBR) assuming free-convective flow. This analysis yielded a DNB ratio
