ACCIDENT AT THE THREE MILE ISLAND NUCLEAR POWERPLANT THURSDAY, MAY 24, 1979 HOUSE OF REPRESENTATIVES, SUBCOMMITTEE ON ENERGY AND THE ENVIRONMENT, COMMITTEE ON INTERIOR AND INSULAR AFFAIRS, Washington, D.C. The subcommittee met at 9:55 a.m., in room 1324, Longworth House Office Building, Hon. Morris K. Udall, presiding. Present: Representatives Udall, Bingham, Weaver, Carr, Markey, Kostmayer, Rahall, Vento, Huckaby, Lujan, Marriott, Edwards, Cheney, and Bereuter. The CHAIRMAN. The subcommittee will be in session. This is the second of our series of hearings on the Three Mile Island accident and problems relating to national nuclear policy. On Monday we heard the subcommittee task force report, and we heard from individual task force members, including Mr. Weaver, Mr. Cheney, Mr. Vento, and others who presented both their collective and their individual views on the excellent investigations they had made on kind of short notice. The task force observations seemed to emphasize design defects, as opposed to operator error, as having been the most significant contributor to the events that ended with major damage to the reactor core. One of the things that bothered me about the situation that existed over the years was the reluctance of industry and of the NRC to accept and respond to technical criticisms suggesting the existence of design defects that might require correction. I was particularly concerned the other day about an analysis prepared by Carl Michelson, a nuclear engineer at TVA, that was sent to both Babcock and Wilcox, and to the NRC. Few, if any, steps were taken to alert the operators of B. & W. reactors of the existence of the problems stressed by Mr. Michelson. It seemed that it took B. & W. about 8 months to respond to his analysis. Mr. Michelson believes that the response did not address certain of his major concerns. I am curious, as I am certain the subcommittee is, as to how the NRC had responded. We were curious the other day, and Mr. Weaver raised the question on Monday as to whether the NRC staff had received the paper, and whether the staff had considered the paper significant. Yesterday we received from the NRC a memorandum that I am going to put in the record of our hearings today, which shows that by early 1978 the memo is dated January 10, 1978-the NRC had received (61) Mr. Michelson's paper, and had accepted one of his principal theses. That is, that the pressurizer water level was not a reliable indication of the reactor pressure vessel being full. What is disturbing about the NRC staff memo is that, while the staff is recommending possible redesign for plants for which no construction permits have been issued, the staff recommended only modified operating instructions for plants where operating licenses are pending, and it recommends nothing at all for operating the plants. I would like to know what action has been taken in response to the Michelson January 1978 memorandum which was 14 months or so ahead of the TMI incident, since it is pretty clear to me that if its warnings had been heeded, it is likely that the TMI incident might have been just a run-of-the-mill event. Today, gentlemen, we are going to hear from General Public Utilities, Babcock and Wilcox, the Edison Electric Institute, and the Atomic Industrial Forum. I would hope that the witnesses might summarize their statements in order that members would have ample opportunity to ask questions. Of particular concern is how we got to Three Mile Island, and how we are going to insure that this does not happen again. We will proceed first with Mr. Dieckamp, who is president of General Public Utilities. I understand you are to be accompanied by Mr. John Herbein, vice president of Metropolitan Edison, and Gary Miller, superintendent of the TMI plant. [The prepared statements of Herman Dieckamp, John G. Herbein, and Gary Miller may be found in the appendix.] PANEL CONSISTING OF HERMAN M. DIECKAMP, PRESIDENT, GENERAL PUBLIC UTILITIES; JOHN G. HERBEIN, VICE PRESIDENT, METROPOLITAN EDISON CO.; AND GARY MILLER, SUPERINTENDENT, THREE MILE ISLAND NUCLEAR POWERPLANT Mr. DIECKAMP. Thank you, Chairman Udall, members of the subcommittee. We are pleased to have the opportunity to be with you today to discuss the accident at Three Mile Island. We have submitted material for the record on the part of not only myself, but also Mr. Herbein and Mr. Miller. In the interest of conserving time and serving your needs best, I thought that I would attempt to summarize very briefly our view of the accident contained in my testimony in three major areas. The CHAIRMAN. Your prepared statements are very helpful. We are trying to compile a comprehensive record from all sides on this, and we will print the prepared statements in the record, and I really appreciate your trying to summarize the high points for us. Mr. DIECKAMP. Thank you. First let me say that we pledge our sincere support and cooperation in the efforts of this committee to make known and to assess the full meaning of this accident. At the outset, we, the company, would like to emphasize that we do not in any way wish to minimize the significance of this accident, and we seek no excuse from our responsibilities as plant owners and oper ators. We do strongly believe that it is important to understand the factors which contributed to this accident, and the ability of our company, Government agencies, and the affected population to cope with it. If this accident is viewed simply as a matter of management or operator failure, the full significance of this experience will be lost. The accident was the result of a complex combination of equipment malfunctions and human factors. The accident departed from the accepted design basis for current nuclear plants. The response of all organizations was influenced by the fact that it was the first accident of this magnitude in the history of the U.S. commercial nuclear power program. It is our hope that this testimony and these hearings can contribute. to an understanding of this accident, the many complex factors that led to it, and the critical learning that we are obligated to derive from it. With respect to the accident causes, we do not propose to present a detailed description or sequence of events for the accident. We are in general agreement with the NRC testimony on this subject as previous ly presented to the U.S. Senate Subcommittee on Nuclear Regulation. We may, however, differ somewhat on the relative importance of the various ingredients of the accident. While extensive data and information have been made available, Met Ed and GPU have not completed a detailed reconstruction of the accident or verified the relative importance of the many ingredients. The following appear to be the major causes of the severity of this accident. Shortly after 4 a.m. on March 28, a reactor coolant system pressure relief valve opened to relieve the normal pressure excursion, but the valve failed to reclose after the pressure decreased. The operator was unaware the valve had not closed. An order for valve closure was signaled in the control room. The operator monitored temperature near the valve to indicate valve position. However, the temperature did not clearly confirm the continuing coolant flow through the valve. When system pressure decreased to 1,600 pounds per square inch about 2 minutes into the accident, the high pressure injection, HPI, safety system was automatically initiated; 4 to 5 minutes into the accident, the operator reduced injection of water from the HPI system when pressurized level indicated that the system was full. Operator training and experience had emphasized the retention of a steam vapor space in the pressurizer. The operator apparently did not anticipate that continued depressurization could lead to steam void formation in hot regions of the system other than the pressurizer and that under these conditions his level of fullness indication was ambiguous and misleading. Because of the presence of steam voids in the primary system, indicted primary coolant flow decreased. The operator turned off the main coolant pumps, after about 100 minutes, in order to prevent damage to the pumps. The plant staff expected cooling by natural circulation and prevented immediate reestablishment of pumping. Voiding prevented natural circulation and prevented reestablishment of pumping. An emergency feed system, designed to provide cooling to the steam generators in case of loss of the normal feed water system, was blocked because of two closed valves. This system would have been available to provide secondary cooling. The operator discovered this condition and initiated secondary system emergency cooling by opening the closed valves 8 minutes after the start of the plant transient. The plant safety system surveillance program had called for the placing of these valves into the closed position six times during the first 3 months of 1979. The last test of the emergency feed system was conducted on the morning of March 26, about 42 hours before the March 28 accident. We have to assume that there was a failure to reopen those valves at the completion of that test. Primary coolant initially vented through the pressurizer relief was pumped into the auxiliary building because the containment design did not require isolation until building pressure reached 4 psi. Continued plant operation required some transfer of fission products to the auxiliary building. The accident differed from the common perception of accidents because of the extended time necessary to achieve a full definition of its scope. In this case, the time required to develop a reasonably complete understanding of the accident and its result was approximately 2 to 3 days. It should be stressed that, while the full impact of the accident was not fully evaluated initially, there was sufficient understanding of system conditions to maintain plant cooling stability during this period. There were three key areas in which full evaluation required time. First, assessment of the degree of core damage. Second, the generation of hydrogen gas during the accident and (a) its potential impact on system heat transfer and (b) its implications relative to core damage. Third, the impact of continued operations on the potential for release of radioactive material from the plant. A preliminary sequence of events was being extracted from the various plant records by the afternoon of March 28. The data for the 16-hour accident period became available in summary graphical form on the morning of March 29. The probable occurrence of a zirconium-water reaction and the presence of hydrogen gas in the reactor containment building was deduced during the evening of March 29 from containment pressure records that indicated a pressure spike during the accident. The size of the hydrogen gas bubble in the reactor coolant system was first measured from system data just after midnight March 30. The concentration of hydrogen gas in the containment building was determined from analysis of the first containment gas sample taken about 4 a.m. on March 31. The first quantitative data with respect to fission product release and degree of reactor fuel damage became available via analysis of a primary coolant sample taken at 5 p.m. on March 29. The data on hydrogen and fission product releases provided the basis for the next level of core damage evaluation. The point of this enumeration is simply to indicate the time necessary to gain insight into the scope of the accident and, in turn, to provide the basis for a meaningful analysis. In any review of the timeliness of the accident assessment, it must be remembered that the plant management and staff faced immediate, continuing, and firstpriority demands to maintain the damaged plant in a controlled and safe state. I would like to also comment on the relationships between the company and the NRC during the incident and the immediately following period. The NRC and the relationship between the company-by the way, this is on page 17 of the prepared statement. The CHAIRMAN. It would be helpful if you designated the pages. The role of the NRC and the relationship between the company and the NRC has been the source of much speculation in the press. The company's view of the relationship is one of mutual respect and cooperation. The popular perception of the relationship may have been significantly colored by the company's election to reserve comment on plant status and plans. The NRC spokesman adequately covered this aspect of communication. The question of who was in charge was not a critical factor. The company has, from the outset, recognized the role of the NRC in this accident situation. The NRC's access to the control room provided direct and immediate access to plant status from midmorning of March 28 onward. The need for NRC approval of "off normal" actions and procedures has occurred with limited bureaucracy. The company encouraged a moderation of the normal regulator/regulatee relationship and invited the NRC to participate directly in the twice-daily technical and progress review meetings at the site. There were tense moments but we must emphasize that it is the company's view that the relationship with the NRC has been constructive and effective. We have been able to close ranks in order to effectively employ our joint resources. I would like to complete my remarks at this time simply by turning to the last two sentences on page 21, and say that this has been a very serious accident. It certainly has tested our capacity to respond, but the institutions charged with the responsibility to supply a secure, abundant, and economic source of electrical energy must be able to withstand the impact of an event like the accident at TMI 2. The system must retain the ability to balance the social and economic costs. of energy supply and energy availability. Again, we pledge our support to all who attempt to fully understand this accident, and to derive the full learning from it. Thank you. We would now be happy to take your questions. The CHAIRMAN. Thank you, Mr. Dieckamp. Let me say that I appreciate the openness with which you have confronted what must be a very difficult time for your company and your business. |