The Commercial Demand is Born
During the late fifties and early sixties, nuclear power went from a novel and developing technology to a commercially competitive energy source. The rapidity of this transition was an outgrowth of a number of factors. Possibly the primary contributor was the natural market forces generated by the intense competition between the two leading vendors of nuclear power, General Electric and Westinghouse.
After the successes of Vallecitos and Shippingport, which was followed by Dresden-1 (GE) and Maine Yankee (W), the two major nuclear vendors took and aggressive stance for moving nuclear power technology to the public. General Electric would start this market assault by offering utilities "turnkey" plants at an investment loss ("turnkey" implies that the utility would merely have to turn a key to start operation). Not to be outdone, Westinghouse followed GE's example of offering "turnkey" plants, also at an investment loss.
The early sixties also spawned an new public focus on the environment. Coal fired plants were falling out of favor. New regulation forced utilities to clean up their plants; hence, the utilities became more mindful of the cost of pollution control in fossil-fuel plants. They increasingly viewed nuclear power as a good alternative to paying the expenses of pollution abatement in existing fossil plants.
With the rapid rise in demand for nuclear power, the Atomic Energy Commission was inundated with a flood of licensing applications. Given that the normal licensing review time took nearly a year to complete, a backlog grew and further delays followed. One of the big issues that surfaced was reactor siting. Some of the licensing applications made proposals to site reactors rather close to metropolitan areas. Until 1960, the most projects where developed in remote locations; however, this was not always a practical option for commercial nuclear power.
Reactor Siting spawns New Research
Prior to 1961 reactor siting was assessed on a case-by-case basis during the normal processing of a licensing application. During the late fifties, proposals were received by the AEC for the Shippingport, Dresden-1, Indian Point-1, and Enrico Fermi plants that were all within 25 miles of a major metropolitan areas (Pittsburgh, Chicago, New York and Detroit, respectively). To demonstrate safety of these plants, licensing proposals had to present through technical arguments for the given design that
1) Recognized all possible accidents which could release unsafe amounts of radioactive materials;
2) Operation procedures reduced the probability of accidents to an acceptable minimum;
3) Through the appropriate combination of containment and isolation, the public was protected from the consequences of such an accident, should it occur.
This general approach to plant safety turned out to be a tedious exercise to assess the degree of total designed safety dependent on many parameters - some of which could only be defended by qualitative arguments. The product of each of these assessments were wordy judgements catered for the given license proposal. Viewed from outside the AEC, these judgements were complicated and confusing and lacked consistency between various licensing proposal reviews.
Beginning in 1961, the AEC began defining a standard regulatory prescription to licensing. The reactor siting issue was the first subject addressed with the new approach. In what would become 10 CFR 100, the AEC began debate on a baseline for what would be required elements in a reactor site. ACRS member Dr. Clifford Beck began this debate with the following assumptions:
1) the probability of a major accident was relatively small;
2) that an upper limit of fission product release could be estimated;
3) that reactors were expected to be in inhabited areas;
4) and that the containment building would hold;
>From these basic assumptions, Dr. Beck outlined a reactor siting criteria that focused on quantitative limits. This lead to the beginnings of a draft criteria. In this draft criteria was the first "sample calculation" for determining dose to a population. The calculation included assumptions for the limiting accident, the Source Term (amount of radioactive substances in the core), the Dispersal of the Radioactivity (including weather considerations), and an evaluation of radiation effects on people.
The solution to this type of calculation provided parameters for determining the exclusion area (no population) and a low-population zone. Within the low population zone, consequences of the maximum credible accident would be limited to dose limits of 25 rem whole body and 300 rem to the thyroid.
After succeeding drafts and reviews of the reactor site criteria, the AEC report TID 14844 was released compiling considerable detail on the methodology and parameters to be used in calculating accident doses to meet the requirements of the criteria. On April 12, 1962, the AEC published 10 CFR Part 100, which went into effect one month later.
The reactor siting discussion brought recognition to a change in the perceived maximum credible accident. For reactors approved for research, the maximum credible accident would likely follow the careless addition of reactivity to the core resulting in a reactor excursion that would result in a intense pulse of radiation that would threaten reactor staff and followed by release of radionuclides to the atmosphere, threatening local population. Commercial designs maintained engineered limiting controls on the rate of reactivity that could be introduced; hence, the criticality excursion was not really credible for commercial plants. Instead, the loss-of-coolant-accident (LOCA) - likely the result of a major pipe break - began to dominate AEC and ACRS meetings. Early calculation should that even following a SCRAM, a large break in the reactor coolant system could leave the reactor core fuel vulnerable to failure and melting. The loss of fuel integrity would release radionuclides out the break and the loss of coolant out the break could threaten the containment integrity.
10 CFR Part 100 does not specify a maximum containment leak rate. The licensing applicant must present a credible analysis based on the physical design of the reactor and containment. Since the utilities had to acquire the property that would be the exclusion zone, they had a considerable interest in using sites with a smaller exclusion radius. To minimize the exclusion zone, utilities pressured vendors to design engineered safeguards into containment system.
The goal of reactor safety reduces to limiting radioactive releases by prevention, or at least mitigation, of accidents through the use of engineered safeguards (also referred to as engineered safety systems or reactor safeguards). Conceptual engineered safeguard features related to LWR LOCA include:
Reactor trip to provide positive and continued shutdown of the nuclear chain reaction
Emergency core cooling (ECC) to prevent or limit fuel melting
Post-accident heat removal (PAHR) to prevent containment overpressurization
Post-accident radioactivity removal (PARR) to reduce the radionuclide inventory available for release
Containment integrity to limit radionuclide release
This change in focus was followed by a considerable amount of research. The AEC sponsored research at three main locations. These were the Containment Mockup Facility (CMF), the Containment Research Installation (CRI), and the Nuclear Safety Pilot Plant (NSPP) at Oak Ridge National Laboratory; the Aerosol Development Facility (ADF) and Containment Systems Experiment (CSE) program at the Pacific Northwest Laboratories; and the Contamination-Decontamination Experiment (CDE) facility at the National Reactor Testing Station. In addition the major vendor such as Westinghouse and General Electric performed much of their own research and co-sponsored some of the AEC directed research. The main thrust of these efforts were in the following areas:
Pressure Vessel Integrity, China Syndrome, and the ECCS Controversy
Following the reactor siting debate, attention focused for a brief time on Pressure Vessel Integrity. This was a heated debate. On one side it was very clear that existing safeguards would be unable to contain the consequences of such a failure; while, one the other side, the reliability of pressure vessels had been 100% in production applications. For most of the 1960s, the AEC, along with the nuclear industry, sponsored a
One of the tangential arguments that came out of this debate was the China Syndrome. During a severe accident, it was theorized that molten fuel could melt through the thick steel pressure vessel, then penetrate the concrete floor, and eventual solidify in the ground after penetrating less than 100 feet (China is about 6000 miles as the worm digs; hence, the nomenclature was an exaggeration). The concern was not so much about the "corium" (melted fuel and core support) in the ground - that could be recovered and managed. Instead, the large amount of stored energy in the corium would be released to the containment and could be enough to breach its integrity.
To some members of the ACRS staff, any means that could be imagined and backed-up with a calculation that could show a breach in containment integrity was enough to stop the licensing of plants. This debate had three positions: the China Syndrome is not credible, improved containment mechanism for the molten core should be developed, and improved core cooling to minimize melt potential. While the debate would linger until the 1970s, the eventual conclusion was that with improved emergency core cooling, the full-scale core melt wasn't credible.
This decision put a lot of focus on the design of Emergency Core Cooling Systems (ECCS).
Emergency Core-Cooling Systems
The matter of loss-of-coolant accidents (LOCA) and emergency core-cooling systems (ECCS) has probably been the major topic of public discussion in regard to light water reactor safety. It is perhaps the matter which has received the greatest attention in licensing reviews, and it has certainly obtained the bulk of the resources expended in nuclear reactor safety research. ECCS did not gain particular attention until after the China Syndrome debates in 1965-66. Prior to that time, the ACRS did not consider ECCS to be a vital safeguard. Early plants such as Indian Point 1 (PWR) and the Dresden 1 (BWR), had very limited ECCS capability. As previously mentioned, the intent of early ECCS designs was to provide an engineered safeguard to justify siting decisions.
The China Syndrome debate introduced the direct correlation between core melt and a loss of containment integrity. Emphasis shifted from containment buildings to the prevention of core melt; and the LOCA received primary attention as the most probable (although still very small) source of an accident which might lead to core melt. The ACRS challenged the major vendors to include improved ECCS designs for utility licensing applications. By the summer of 1966 General Electric responded in support of the Dresden 3 plant. They introduced a redundant core-flooding system and an automatic depressurization system which would reduce the primary system pressure sufficiently to maximize the effectiveness of the low-pressure core spray or core-flooding system. Later that year Westinghouse introduced accumulators which contained large amounts of water under gas pressure.
The ACRS reviewed both vendor design changes and found them acceptable; however, the ACRS did record some misgivings about Westinghouse's accumulator system. They wrote, "It is conceivable that, in the case of a cold leg break, significant flow from the two remaining accumulators could be lead out of the vessel through the inlet plenum." This observation would serve as an omen for future debate on ECCS design.
The ability of an ECCS to maintain core integrity during a LOCA is assessed by mathematical modeling of the phenomena involved. Such models are built from the report results of many experiments. The primary goal of ECCS research has been to quantify human understanding of the processes involved in cooling the reactor core. Specifically, ECCS research is focused on fluid flow and heat transfer mechanisms. Of course, these mechanisms are closely couple with the neutronic behavior of the nuclear reactor which can complicate analysis; however, the primary complication for light water reactors comes from the interactions of both liquid and vapor water coolant. The flow and heat transfer characteristics of a two-phase fluid can vary substantially under accident conditions. For these reasons, a large investment has been made to develop very accurate models of fluid flow and heat transfer processes of the complete spectrum of possible reactor system conditions.
Tracking Core Integrity during a LOCA
The calculation of the expected fuel temperature transient following a LOCA requires a broad understanding of fluid flow, heat transfer, thermodynamics, mechanical components, reactor dynamics, and the geometry design of the reactor. For all but the simplest problems a computer program is required that contains in one package all the physics. The fundamental principles underlying the mathematical models are the physical requirements to conserve mass, momentum, and energy of the whole system.
The development of thermal-hydraulic, safety-related computer codes in the nuclear industry has been closely tied to experimental research and development. For complex systems such as nuclear power plants much of the phenomena present can only be quantified in terms of empirical correlations. Empirical correlations can be as simple curve fits to a set of data or sophisticated expression incorporate some physical basis. Regardless, experiments are needed to derived the new models and then to validate the accuracy of the models under varying conditions.
The earliest computer code was developed by Westinghouse. This was the FLASH code. FLASH used the very simple "node and branch" approach to modeling. This approach assumes that the spatial elements (nodes) are capable of energy and mass storage only. The nodes are connected by branches the model resistances to flow. Thus, the conservation-of-momentum is written for each branch, while the conservation-of-energy and -mass are written for each node. The FLASH code, as well as other early PWR codes, were based on single fluid models (liquid water coolant). By the 1960s there was already a very good understanding of the properties of liquid water; however, for BWRs and for accurate accident evaluations, a two-phase (water/steam) mixture exists. At that time there was relatively little understanding of two-phase properties of water coolant.
General Electric had a vested interest in understanding two-phase flow phenomenon. Fortunately for GE, many of the research and development organizations around the world also recognized this need. From 1955-1975 an aggressive investigation of boiling heat transfer and two-phase flow generated a large body of knowledge. This huge effort provided the necessary data for modeling boiling water phenomenon in a nuclear plant under a large range of conditions.
Probably the biggest development in two-phase flow research was performed by Novak Zuber. His work in the early to mid-1960s focused on recognizing how steam and water separate in a two-phase flow condition and quantifying the volumetric separation from measurable fluid properties. This work is referred to as the Drift Flux model and is still used today in the major industry thermal-hydraulic computer codes.
The understanding of the phenomena related to the LOCA was advanced by the efforts of many people working all around the world. Significant advances were made in many areas including:
The End of a Golden Age
By the end of the 1960s, there was a lot of excitement for the future of nuclear power. Great strides had been made in thirty short years. Unfortunately and despite the large body of research that advanced the worlds understanding of the nuclear power plant machine, the biggest hurdles were yet to come. While for thirty years investment into the technology of nuclear power plants grew, suspicion of the industry and the governments that supported also grew. These forces would come to a head in the 1970s.
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