Tag Archives: Will Davis

Kewaunee: What does the future hold?

Posted on May 9, 2013 by Will Davis| 5 Comments

By Will Davis

Shortly after 11 a.m. on Tuesday, May 7, 2013, the operators at Dominion Resources’ Kewaunee nuclear power plant opened its output breaker, disconnecting the turbine generator from the grid for the last time after just under 40 years of operation. Shutdown of the reactor followed, and the plant entered what for some is an uncertain (even if pre-ordained) future—a long-term storage period, followed eventually after many years by the complete dismantling and removal of the plant.

Prior to the shutdown, Dominion had announced its decision to change the plant’s status (after the shutdown) to what is called SAFSTOR, which, just as it sounds, implies “Safe Storage.” The Nuclear Regulatory Commission’s official definition of SAFSTOR reads as follows: “A method of decommissioning in which a nuclear facility is placed and maintained in a condition that allows the facility to be safely stored and subsequently decontaminated (deferred decontamination) to levels that permit release for unrestricted use.” This definition implies that a long period of time will be allowed to elapse before serious and heavy dismantling and removal of key plant components is performed, and before the many site structures are completely demolished and removed.

While the intensity of radiation around the immediate vicinity of the reactor and steam generators is slight compared with when the plant was in operation (and those areas unoccupied), it is not insignificant. The time period between the final reactor shutdown and the beginning of the disassembly of the ‘heart’ of the plant will help in a major way to reduce the radiation exposure of the people who will be required to perform the work—not a small consideration, even in a relatively small nuclear station such as Kewaunee.

Briefly, in disposing of a shut down nuclear plant, there are three options: Decommissioning immediately, which means relatively quickly launching into demolition; SAFSTOR, as described above; and ENTOMB, wherein a plant and some of its components are sealed and abandoned in place for a long period of time or permanently. (Piqua and the Hallam Nuclear Power Facility are two examples of former commercial nuclear stations in this status.)

Dominion has, under federal law, 60 years to complete the entire complicated and expensive decommissioning process, which will see the nuclear plant site returned to “green field” status (releasable for any use) with the exception of a dry cask type spent fuel storage facility. According to Dominion’s latest 10-K filed with the U.S. Securities and Exchange Commission, decommissioning cost overall will total $680 million; the decommissioning fund presently has roughly $578 million, with the rest expected to be made up by future earnings. Dominion took a $281 million after-tax charge in the third quarter of 2012 as a result of deciding to decommission Kewaunee.

SAFSTOR

Kewaunee is not by any means the only nuclear plant that will be in, or has been in, the SAFSTOR condition. There are a number of other plants that were placed in this condition either to prevent disruption of the operation of other plants on the same site and/or take advantage of economies of decommissioning multiple reactors at once (Dresden Unit 1, Peach Bottom Unit 1, and Millstone Unit 1 all fit in this category, since they are in SAFSTOR and occupy sites that in all cases contain two other operating nuclear plants.) Other plants, such as Dairyland Power Co-Op Genoa No. 2, which was much more commonly known by its Atomic Energy Commission title as the Lacrosse Boiling Water Reactor, was in a state of modified SAFSTOR for many years as most of the heavy work was deferred while some limited disassembly went on in irregular phases.

In the case of Kewaunee, Dominion will relatively soon (in the next months) remove the fuel from the reactor and move it to the spent fuel pool. Dominion will notify the NRC within 30 days, in writing, that it has shut down the reactor for good; after the reactor has been defueled, Dominion will again notify the NRC, which will issue a license amendment rendering the plant “possession only” in regulatory status, wherein Dominion cannot fuel, much less operate, the reactor.

A Post-Shutdown Decommissioning Activities Report (PSDAR) will be submitted to the NRC by Dominion within two years, which lays out expected procedures, timelines, and costs. Ninety days after the NRC receives this report, the plant owner could conceivably begin heavy demolition and component removal if the disposal choice were immediate decommissioning. However, in the case of Kewaunee, the plant will remain in a monitored state, with (very likely) some component removal taking place slowly.

A Dominion spokesman told Platts that the expectations are that Kewaunee’s spent fuel pool contents will be moved entirely to dry cask storage on site by 2020. Much later, in June 2069, heavy dismantling of the plant will begin with completion expected in August 2072.

Decommission

The difficult work will begin when Dominion finally commences the physical dismantling of the plant. Many readers may not be aware that a number of large (and small) nuclear power plants have been not only shut down, but completely demolished and removed. The challenges encountered at each included both expected and unique problems; the work is complex and time consuming, but is proven to be able to release a site completely for other use. A few examples are in order:

Big Rock Point containment under demolition; courtesy Consumers Power

Big Rock Point containment under demolition. (Consumers Power)

Big Rock Point: This plant (designated by the American Nuclear Society in 1991 as a Nuclear Historic Landmark) was an early General Electric boiling water reactor plant in a remote area of Michigan. The plant operated successfully from 1965 through 1997. Over the next nine years, Consumers Power completed major site surveys and engaged in the complete demolition of the plant. Heavy components such as the reactor vessel were shipped to South Carolina for burial. Thirty-two million pounds of concrete were removed; 53 million pounds of material labeled as low-level radioactive waste were transferred to storage facilities in other states. Fifty-nine million more pounds of clean (uncontaminated) building materials were transported to landfills and buried. The entire 560-acre site was returned to “green field” or a natural state in August 2006, except for the independent spent fuel storage facility.

Connecticut Yankee: This plant, when shut down in 1996 after 28 years of operation, was designated for immediate decommissioning with no SAFSTOR period. The project to return the site (except for spent fuel storage) to green field took place over the period 1998–2007, and 525 acres of natural terrain were the result. Small sections of the property have begun to be turned over to other owners, such as the U.S. Fish and Wildlife Service.

Yankee Rowe site as it appears today; courtesy Yankee Atomic Electric

Yankee Rowe site as it appears today. (Yankee Atomic Electric)

Yankee Atomic Electric: The nuclear plant constructed by this company was among the very earliest commercial power stations, yet operated for 30 years. After final shutdown in 1992, the plant began decommissioning the next year. From the official website of the plant: “Since the start of physical decommissioning in 1993, more than 21 miles of piping and tubing, 1071 valves, 8569 pipe hangers, 321 pumps, and 33 miles of conduit and cable tray have been removed. In addition, six large components weighing a total of more than 500 tons were also removed. Some of the material, including the large components, was sent to the Barnwell, S.C. low-level radioactive waste disposal facility for permanent disposal. Some of the metal was sent to a processing facility in Tennessee.” Over 1700 acres have been released by the NRC and are being considered for future use in a scenic, natural environment.

Component and structural removal

Eventually, the most solidly constructed components of Kewaunee will have to be removed; these are the reactor building and the components inside of it. Projects in the past have encountered special problems and considerations in this type of work, but enough ground has been laid in past years to provide ample experience in this project. Here are some interesting reactor plant related project links:

The International Atomic Energy Agency hosts an excellent presentation by Bluegrass on the processes used to remove the reactor vessel at the long-SAFSTOR but now decommissioning Lacrosse BWR in Wisconsin; see it here. Particular problems were encountered with very small clearances around the reactor vessel, especially at its lower head.

Saxton decommissioning; courtesy GTS Technologies

GTS Technologies has an impressive set of web pages showing the work it did to remove the reactor containment building at the former Saxton nuclear reactor in Pennsylvania.

The final result—in 60 years

Kewaunee employees right now aren’t thinking about whether or not someone will, eight or nine decades from now, be having a picnic or plowing a field on the spot where the plant’s turbine building once stood. They’re worried about where they’ll find work—Reuters has reported that 200 of the 630 workers will be laid off at the end of May, 100 more in another month. By the middle of 2014, the plant will have just under 300 permanent workers on site; this number will remain (along with outside contractors) for the duration of the procedures. Dominion has not yet announced whether or not it intends to contract some or all of the work to an outside company such as EnergySolutions, whose ZionSolutions unit is presently decommissioning Zion Nuclear Station.

Long after the memories of the stress of the workers’ movement and breakup of the Kewaunee Station’s family is over, it’s the intent that the plant site will be returned to as completely natural a state as is possible. As we’ve seen, even though this work will provide many challenging days ahead, it’s not only possible but proven—and perhaps, if we’re lucky, some entity will erect a sign at the site to tell future generations that a complete nuclear power station was built and operated here for many years, and then completely removed. It will be proper if a sign is needed in order to be able to tell.

——————–

(For more information on the nuclear plant decommissioning process, you can read the NRC’s excellent pages on the topic by clicking here. In addition, other sites that have decommissioned include Maine Yankee, Rancho Seco, and Trojan. Part of the former Rancho Seco nuclear plant site is now the Rancho Seco Recreational Area.)

_____________________________

Will Davis is a consultant to, and writer for, the American Nuclear Society; he will serve on the ANS Public Information Committee 2013-2016. In addition to this, he is a contributing author for Fuel Cycle Week, and also writes his own popular blog Atomic Power Review. Davis is a former US Navy Reactor Operator, qualified on S8G and S5W plants.

→ 5 Comments

Posted in decommissioning, licensing, News, Nuclear Energy Narrative, Nuclear events, Nuclear History

Tagged nuclear history, nuclear power, Will Davis

The Hook-Ons

Posted on April 18, 2013 by Will Davis| 13 Comments

by Will Davis

This week’s announcement by Babcock & Wilcox that it had signed the long-awaited funding agreement with the Department of Energy has been taken by advocates of small modular reactors (SMRs) as just the latest good news on the inevitable path to construction of at least one prototype nuclear plant using SMR reactor technology in the United States. It is widely hoped that this is the harbinger of the rapid spread of the market for SMR plants.

The chief advantage of SMRs other than cost reduction over large 1000–1600 MWe nuclear plants is that they can be located practically anywhere (assuming proper geologic characteristics and supply of cooling), since a primary design feature is that the major components of the reactor plant itself are to be easy to ship (i.e., by large truck over existing highways). This design asset potentially opens up locations previously considered unworkable (for large plants, with their enormous reactor vessels and other equipment that needs to be shipped intact to site) and may, in some cases, allow siting of SMR-driven power plants nearer to populated areas in order to take advantage of benefits to the grid (by siting source nearer to use) and even, if some have their way, to supply steam for process use to facilities already in existence or built new.

These concepts—siting closer to communities than with large commercial plants, and supply of steam for existing facilities—are, in fact, not new. In the early days of nuclear energy, a number of nuclear plants were built in order to supply steam to facilities already in use. In the cases of these early reactors, the facilities were all commercial electric power stations; the group of reactors came very loosely to be known as “hook-on” reactors. The concept of expanding the use of nuclear energy in such a way was actively pushed by the Atomic Energy Commission; three of the four plants we’re about to explore were (at least partly) funded under the AEC Power Demonstration Reactor Program.

Elk River (Minnesota)

The Elk River Reactor, widely heralded as “Rural America’s First Atomic Power Plant,” was originally contracted to ACF Industries in 1959 for construction behind the Rural Co-Operative Power Association’s Elk River coal-fired plant (seen at far left in the above post card photo.) The reactor plant was a novel natural circulation, indirect cycle boiling water reactor that, while not fitting the modern definition of “small, modular” of today’s SMRs, did have a reactor vessel small enough to be shipped to the site on the smallest standard railroad flat car of the time (said cars measured 40 feet in length overall.) The 58-MWt reactor produced saturated steam at 922 psig and 536 °F, but the existing turbines in the plant required superheated steam. Construction of a coal-fired superheater interposed between the reactor plant and the power plant adjusted the steam conditions to 612 psig but 825 °F; of the total 22 MWe of generating capacity credited this installation, 7 MW was provided by the superheater.

The plant suffered teething pains that, today, seem not too surprising given the facts that the original reactor vendor was small, and that it was actually bought out by Allis-Chalmers while construction of the Elk River Reactor was in progress. Fuel element defects and reactor pressure vessel cladding cracks contributed (among other things) to delays in the start up of the plant, which did not achieve commercial operation until mid-1965, but after which operated with a very fair degree of reliability.

Eventually, further leakage from welds in the primary coolant system caused investigation into the overall condition of all welds in that system in 1968, and the determination was made that major rework would be required to fix the problems—a problem that looked all the worse given that Allis-Chalmers had decided to exit the nuclear power business in 1966. After considerable debate about what to do with the reactor plant (which was still technically AEC owned), the decision was made in March 1971 to decommission the reactor plant and completely remove it from the site. Below, a March 1971 UPI telephoto showing the plant as it looked at the time that the decommissioning decision was made.

Piqua (Ohio)

The Piqua Nuclear Power Facility (PNPF) was built in the early 1960s in the town of Piqua, Ohio, as a part of the second round of the AEC Power Demonstration Reactor Program. The reactor was unique among the world’s commercial power reactors in being an organic-cooled and -moderated design. A commercial terphenyl preparation (marketed widely as Santowax-OMP by Monsanto) was used for this plant that, because of the low pressure of the primary, originally was designed without any containment whatsoever. The Advisory Committee on Reactor Safeguards, however, ordered that a containment be built. The reactor plant was built just across and down the river from the original Piqua municipal generating station, and supplied steam to it at 450 psia and 550 ºF through underground piping and a new bridge structure over the river. The reactor was rated 46 MWt, and the electric generating capacity credited to it was 11.4 MWe.

The Piqua Nuclear Power Facility is seen on the right, which is the east side of the Miami River; the municipal power plant is on the West side, just upstream.

PNPF began operation in 1963 and operated with occasional problems largely due to coolant breakdown until 1968 when a serious blockage occurred. The decision was made by the city of Piqua not to take over ownership of the plant, and it entered procedures to shut down and decommission immediately. The disposal method (after defueling) was selected by the AEC was SAFSTOR, in which the plant is left in place to allow decay of radioactivity at the same time guaranteeing no impact to the surroundings. The containment and support buildings are still clearly visible in Piqua to this day.

CVTR (South Carolina)

The Carolinas-Virginia Tube Reactor was built adjacent to an existing coal-fired plant (and hydroelectric dam facility) at Parr, South Carolina, under the third round of the AEC Power Demonstration Reactor Program in order to test out the pressure tube reactor concept. This plant was widely reported and heralded in the early 1960s as “The Southeast’s First Atomic Power Plant.” Westinghouse provided the 65-MWt pressurized (tube type) heavy water cooled and moderated reactor; Stone and Webster acted as architect-engineer. The plant (like Elk River, but unlike Piqua) required external superheating; of the rated electrical 17 MWe, 1.7 MWe was contributed by the superheater. The reactor and superheater provided steam at 415 psia and 725 ºF to the old powerhouse near by.

Carolinas-Virginia Nuclear Power Associates owned this plant; this organization was comprised of Duke Power Company, Carolina Power & Light Company, South Carolina Electric & Gas Company, and Virginia Electric and Power Company (the latter often referred to as VEPCO).

Below, a spectacular original pencil rendering of the CVTR plant facility, including the powerhouse and environs, from my collection. The drawing’s labeling is clear when blown up; it is signed “E.E. Grant 1960.” (Click to enlarge.)

The CVTR started up in 1962, and like the other plants we’ve shown so far, had a very short operating life (five years,) shutting down for good in 1967. The reactor was in SAFSTOR condition for many years, but in much more recent times has completely been decommissioned and removed, and today there is very little sign that the plant was ever there. Of course, the site of the former Parr generating station and the adjacent CVTR installation is quite near the Virgil C. Summer Nuclear Generating Station, which today is seeing construction of two Westinghouse AP1000 plants—so that the area of “The Southeast’s First Atomic Power Plant” is again at the cutting edge of nuclear energy’s advance.

Saxton (Pennsylvania)

A fourth early reactor actually is one that contributed the least to commercial power generation of those we’re visiting here, and is also that which is most commonly found in the literature to have the appellation “hook on”.

The Saxton Generating Station was selected to host construction of a nuclear reactor whose primary purpose was developmental testing of fuels, and which was to be officially known as the Saxton Experimental Nuclear Reactor. Owner of the reactor was Saxton Nuclear Experimental Corporation, a non-profit entity formed by Pennsylvania Electric Company, Metropolitan Edison Company, New Jersey Power and Light Company, and Jersey Central Power and Light Company—all of which were subsidiary companies of GPU or the General Public Utilities System. The diminutive pressurized water reactor, rated originally 20 MWt, had only a single loop (and thus one coolant pump and one steam generator) and provided steam to the center of Saxton Generating Station’s three turbine generators. While the containment was clearly visible beside the coal-fired plant, for safety reasons (considering the surrounding community) the reactor vessel was actually located some 15 feet below grade.

According to the February 1959 Atomic Industrial Forum “Forum Memo” magazine, in which the contract for the reactor was revealed, GPU had actually announced that it was considering a “hook on” at Saxton back in 1957 after terminating an investigation into building a pressurized water reactor in the Philippines (another GPU subsidiary was Manila Electric Company.) At that time, the rating of the Saxton plant was given as a very modest 5000 ekw (which we would now write as 5 MWe), although in point of fact later testing was planned at far above the original rated figures; the turbine to which the reactor piped steam was actually rated nominally at 13 MWe, allowing considerable room for uprating for temporary testing.

In the March 1959 issue of the Forum Memo, Elmer L. Lindseth, president of Cleveland Electric Illuminating Company and chairman of the Edison Electric Institute’s Committee on Atomic Power, was quoted as saying that Westinghouse would build the Saxton reactor plant at a fixed price of $6.25 million. GPU would under the same agreement provide the site, use of the No. 2 turbine, and bear operating and maintenance costs—all of which figured to roughly $2 million. Westinghouse also had exclusive fuel production rights for five years.

With Gilbert Associates serving as architect-engineer, construction of this unique “hook on” began in February 1960 (with AEC Construction Permit CPPR-6.) A provisional operating license was issued in November 1961, and the reactor fueled in early April 1962, with criticality achieved at 1:40 AM on April 13, 1962. (Below, a view of the Saxton Experimental Nuclear Reactor next to the Saxton Generating Station.)

As has been mentioned, this plant was not entirely intended as a commercial power reactor; rather, its focus was the development of technology for further, future reactors. Quoting GPU in a Saxton advertising brochure of the day, “Investor owned utilities, dedicated to serving consumers in all walks of life, have invested $8,500,000 of private funds in the nation’s newest operational nuclear reactor so that ‘unknowns’ can be converted into ‘knowns’ and personnel can acquire valuable operating experience for use in designing and manning larger reactors in the future.”

Among other concepts, Saxton experimented with chemical control of reactivity (“chemical shim,” or use of boron in the primary coolant to control reactivity instead of just control rods) and also conducted extended operations with plutonium fuel (MOX or “Mixed OXide” fuel, containing both natural uranium dioxide and plutonium dioxide) beginning in the mid-late 1960s.

As a result of the nature of the program, it appears in retrospect that the plant spent as much of its life operating as not. From the 1964 AEC Report to Congress: “The Saxton Nuclear Experimental Corp.’s pressurized light water reactor near Altoona, Pa., was returned to power operation on January 30, after having been shut down since the previous November for modifications. The reactor, while producing small amounts of electric power, is primarily used for experiments to determine ways in which more heat energy can be obtained from specified amounts of fuel.” It would thus in hindsight be appropriate to consider that the waste heat from the Saxton reactor was not entirely wasted, if we simply view it as a byproduct of advanced fuels testing, by way of connecting the plant to the Saxton Generating Station.

Saxton was finally shut down in May 1972, and after a prolonged period of decommissioning, there is nothing visible at the site to hint that a power station of any sort once existed there. The entire power plant and reactor facility has been removed down to several feet below grade, and the area has been backfilled.

In closing, it’s interesting to consider the notion that today’s concept of placing lower output, transportable nuclear reactors at a now-expanded range of possible locations actually had a roughly correlative precedent early in the construction of nuclear power stations in this country. In the siting of plants nearer to populated areas, and in the use of small plants on grids that could not handle extremely large single generating sources, the early experience was perhaps a herald of things to come, even if it did take another roughly half century and the development of truly integrated, truck transportable, and inherently safe SMRs in order to realize the dream held up for these early small plants. The wide design disparity and newness of the technology associated with these early plants seemed to hint at troubles, which surely were encountered, but today nuclear technology is a half century further down the road so that the question of operability is quite far removed from consideration. As it turns out, everything old is new again—but today, with far better promise of success.

(All illustrations – Will Davis collection. Please do not reproduce without permission.)

(“Atomic Industrial Forum” was a trade group formed in 1953; it is a lineal predecessor of today’s Nuclear Energy Institute.)

______________________________

Will Davis is a consultant to, and writer for, the American Nuclear Society. In addition to this, he is a contributing author for Fuel Cycle Week, and also writes his own blog Atomic Power Review. Davis is a former US Navy Reactor Operator, qualified on S8G and S5W plants.

→ 13 Comments

Posted in News, Nuclear Energy Narrative, Nuclear History, Nuclear pioneers, Small modular reactors

Tagged nuclear history, Will Davis

Inherent and engineered safety: Did Weinberg predict today’s reactors a quarter century ago?

Posted on March 28, 2013 by pbowersox| 6 Comments

By Will Davis

Following the Three Mile Island (TMI) accident on March 28, 1979, it seemed to many as if a slowing nuclear energy industry in the United States had been dealt a death blow. It had not, but the public’s confidence was shaken, and this blow to public opinion built upon a decade’s worth of intensive, focused anti-nuclear effort on the part of a number of large well-funded special interest groups.

Weinberg

Once the causes of the TMI accident were well understood, the task was taken up to predict what would be desirable for increased public support for new reactor construction. Alvin M. Weinberg headed a group that performed such a study under a 1981 request by the Institute for Energy Analysis; the published result was the book The Second Nuclear Era—A New Start for Nuclear Power (1985).

The conclusions reached were numerous in terms of specific recommendations, and the determination as to reactor technology was clear: Contemporary light water reactor (LWR) plants at the time, given their previous safety record, were acceptable to the public—and future designs should be improved and be either inherently or passively safe, certainly in terms of cooling, and perhaps even in terms of shutdown. The group believed that the future of nuclear energy in the United States would initially be based on proven technologies, either already in wide use (LWR plants, specifically pressurized water reactors/PWRs), or already developed to the point of commercial application (high temperature gas-cooled reactor (HTGR) plants, such as Peach Bottom-1 and Fort St. Vrain.)

Contemporary designs (1980s) and development

The earliest nuclear reactor plants were designed with basic water injection systems intended mostly to handle “makeup”—because of the early emphasis in design basis accident analysis for rapid-loss-of-coolant accidents, most had some way to rapidly make up water should a large primary coolant pipe break. This essentially covers most designs through the mid-1960s.

In the middle of 1966, ongoing work by the Atomic Energy Commission (AEC) and the Advisory Committee on Reactor Safeguards, in the processing of applications for comparatively very large reactor plants, began to “force the issue” of increased emergency core cooling systems (ECCS) to the forefront of discussion. The radioactive release possible with larger cores had not been considered in previous standardized siting criteria, or accident analysis. Dr. William E. Ergen was appointed by the AEC’s director of regulation to form a task force to study this problem; the major result was the determination that a relatively much larger, newer core, if uncooled, could cause melt-through of the reactor vessel (because larger power output plants did not have proportionately larger total area for heat dissipation, without added forced dynamic cooling; whereas earlier reactor cores could survive being uncooled.)

Indian Point-1

The Ergen Report made it clear that greatly enhanced ECCS capability would be needed to continue to prove safety, and AEC ordered that plants had to fit or backfit new, higher-capacity equipment meeting revised ECCS requirements by 1974. Plants that were unable to comply had to be shut down; Indian Point-1 shut down permanently in 1974 for this reason.

This improvement in ECCS focus led indirectly to the ability to build nuclear plants in locations previously not considered possible by then-used siting criteria. A letter from the ACRS to the chairman of the AEC in 1964, when early consideration of improved safeguards was underway, stated in part:

“It is the opinion of the Advisory Committee on Reactor Safeguards that the including of properly engineered safeguards in reactor plants can permit the reduction of distances required for protection of the public and that engineered safeguards of selected type should make feasible the siting of power reactors at many locations not otherwise considered as suitable.”

Post-TMI: Cancellations and public opinion

The causes of the TMI accident were many, varied, and in many ways intertwined. The complexity of the problems facing the industry became clearer as months of reviews and rulemaking dragged into years, and many nuclear plants under construction began to experience incredible delays—first, when all licensing was held up; and then, when plant owners and operators attempted to determine how to backfit or modify existing designs to bring new, but not-yet-started, reactors up to the present specification (which was itself a moving target). For example, in 1983, Detroit Edison stated that costs for its yet-to-start Fermi-2 had skyrocketed due to, among other things, $138 million in TMI-related backfits and modifications.

The effect of TMI on public opinion is commonly stated today in the press as something of a “death blow,” but this is inaccurate. First, public opinion about nuclear energy was starting to move since about 1970, with the first Earth Day and the passage of the National Environmental Policy Act (which later would be used to force nuclear plants to consider environmental impact as a stand-alone topic, which was not done originally). According to a compilation of public opinion research and analysis entitled Public Opinion and Nuclear Energy (1983), public opinion in the United States was already shifting in the mid-1970s away from mostly supporting nuclear power, and public beliefs about reactor safety “changed somewhat from 1975 through 1980.” Public opinion was beginning to change before TMI happened.

Also, public discourse over cost, delays, and cancellations of nuclear plants was increasing. Over 30 nuclear plants had been cancelled, and a number of plants under construction had been pushed back, prior to TMI. This trend increased after TMI.

However, according to this study, opinions on nuclear energy in the United States still did not swing wholly anti-nuclear by any means as a result of the TMI accident. In this study’s summary of post-TMI surveys, it is concluded that

“Although a majority of the general public and most leadership groups believed that there is no guarantee against a catastrophic nuclear accident and that fundamental regulatory changes are necessary to keep risks within tolerable limits, a majority of the public and leadership groups favored the continued use and expansion of nuclear power.”

Weinberg and the direction to a second nuclear era

We have covered a “snapshot” of the development of the nuclear industry in terms of safety engineering (by no means complete; a detailed study would require a career or two) and a “snapshot” of public opinion when Weinberg and his group were tasked to imagine a “way out” for the nuclear industry and nuclear power. As it turns out, Weinberg’s general predictions (detailed earlier) were exactly correct; however, a shortcoming in the study’s conclusion was a dependence on either wholly new plant designs or the use of already-sidelined designs in pursuit of the stated goals.

PIUS – click to enlarge

Weinberg and his cohorts did in fact admit that contemporary LWR designs (Westinghouse SNUPPS/Sizewell B, GE ABWR, Combustion Engineering System 80) were safe enough for public acceptance, but stressed a look forward to two other designs—the Process Inherent Ultimate Safety (PIUS) reactor, and a form of HTGR. The PIUS was a radically different type of light-water-cooled reactor, developed conceptually by ASEA-ATOM (Sweden), that used a gigantic prestressed concrete vessel, no control rods (reactivity control by boron and temperature only), and was said to have a “hands off” time of one week, in which no operator action was required after any potentially damaging failure. The core would remain covered and cooled at all times in this unusual, and never-built, design. The other design that Weinberg’s team selected was a General Atomics HTGR, helium cooled and graphite moderated, with inherent safety features and guaranteed core cooling capability by virtue of basic design—also never built.

What is significant about the selected designs is their “walk away” capability, wherein no operator action was required after potentially damaging incidents (such as loss of all electrical power.) Weinberg was essentially correct in believing that this would be required to gain public acceptance on a wide scale; what he did not envision was a way to mate existing, developed reactor plant design (hardware) with his vision of inherent or “walk away” safety to arrive at a workable, licenseable, affordable, and realistic nuclear power plant. The industry had already, by that time, become wary of any design that was not a light-water-cooled reactor, either PWR or BWR, and the post-TMI licensing logjam practically guaranteed that no radically new design would be licensed in any realistic or desirable time frame (and a reduction in estimated electricity demand guaranteed that no utility would try.)

The future from the past—AP600 to AP1000

In 1992, the National Academy of Sciences (NAS) conducted a study that, among other things, developed a list of promising reactor designs for future application. While the PIUS and another gas-cooled reactor still figured in the NAS report, the bulk of the recommended designs were LWR plants grouped into two categories—”Large evolutionary LWR” plants such as the ABB- Combustion System 80, the GE ABWR, and the Westinghouse APWR (later to become the Mitsubishi APWR and eventually the US-APWR designs) and also, interestingly, “Mid-size passive LWRs” which included a GE SBWR or “Simplified Boiling Water Reactor,” and a Westinghouse design known as the AP600, for “Advanced Passive 600.”

The AP600 design was originally developed with support from the US Department of Energy and the Electric Power Research Institute as a simpler, less complicated, and less expensive proposition than large commercial nuclear stations with net outputs over 1000 MWe. At the time the AP600 was conceived, modular construction was incorporated in the design (as it is with today’s familiar AP1000) and the innovative passive cooling features seen in today’s AP1000 were also incorporated—including the core makeup tanks, accumulators, and the IRWST or in-containment refueling water storage tank. After exhaustive review, the AP600 was given design certification by the Nuclear Regulatory Commission in December, 1999.

The AP600 was not large enough to attract utilities in the United States, but a much larger 1000-MWe direct descendant—the AP1000—was; Westinghouse filed an application for design certification for this large, advanced passive-cooling plant in 2002, and the design was certified in December 2011.

©2013 Westinghouse Electric Company LLC. All right reserved. Image reproduced with Westinghouse’s permission.

Westinghouse AP1000
©2013 Westinghouse Electric Company LLC. All rights reserved. Image reproduced with Westinghouse permission.

In the requirements for passive safety—ECCS requirements that didn’t involve large offsite or onsite AC power supply, and didn’t require operator action—Weinberg, et al. were fully correct in their conception of what a continuous drive for safety, and thus public acceptance, demanded. Public misinformation about nuclear energy had so badly eroded realistic perceptions that, after TMI, many in the public actually believed that nuclear reactors could explode like nuclear weapons—which drove home the need for both a major shift in public perception and a major push in the industry for truly passive, and truly credible, core safety.

Weinberg and his team, however, did not anticipate that developments originally intended for intermediate-sized, less expensive plants for remote siting would be successfully applied to commercial (1000 MWe+) sized plants, giving both the safety required and the necessary dependence on the rugged engineering of decades of previous LWR experience. The selection of the recommended PIUS design, for example, was made in part because it could build on previous LWR experience; the text is quoted as saying “since PIUS is a modified PWR, much technology already in commercial use could be applied.”

What really happened was that passive features were eventually applied external to the core, and external to the containment, which along with rugged (and in some ways traditional) construction of the primary plant worked together to assure safety. There was no need for a radical departure at highest possible speed from most or all of conventional LWR technology; the best (and the ultimate) solution was to apply passive cooling principles to developed PWR design—a vision targeted not specifically by Weinberg and his team, but targeted perfectly in effect.

SELECTED BIBLIOGRAPHY:

Bodansky, D.; Nuclear Energy – Principles, Practices and Prospects. New York. Springer-Verlag 1986.

Detroit Edison Company; A History of Enrico Fermi Atomic Power Plant Unit 2. August 1983.

Nealey, S. M.; Melber, B. D.; Rankin, W. L.; Public Opinion and Nuclear Energy. Lexington, Mass. D. C. Heath and Company 1983.

US Atomic Energy Commission—WASH 1082, Civilian Nuclear Power—Current Status & Future Technical & Economic Potential of Light Water Reactors. March 1968.

US Atomic Energy Commission—WASH 1250, The Safety of Nuclear Power Reactors (Light Water-Cooled) and Related Facilities. July 1973.

Weinberg, A. M.; Spiewak, I.; Barkenbus, J. N.; Livingston, R. S.; Phung, Doan L.; The Second Nuclear Era—A New Start for Nuclear Power. New York. Praeger Publishers 1985.

______________________

→ 6 Comments

Posted in Nuclear History, reactor designs

Tagged alvin weinberg, weinberg, Will Davis

San Onofre debate now more public – and more technical

Posted on March 15, 2013 by Will Davis| 3 Comments

By Will Davis

The debate over the continuing investigations into steam generator U-tube problems at San Onofre Nuclear Generating Station (SONGS) last week entered a new phase of heightened publicity and public scrutiny as the Nuclear Regulatory Commission (NRC) released Mitsubishi documents which detailed that company’s investigations into the root causes of the problems.

Friday, March 8, saw the release of a pair of documents which had been redacted by Mitsubishi Heavy Industries (MHI) (redaction here means that sensitive corporate information that competitors could use to advantage had been removed). This followed the revelation within the previous weeks that an original of this document had somehow fallen into the hands of US Senator Barbara Boxer and US Representative Ed Markey, who then touted the documents as a “smoking gun” showing that plant operator Southern California Edison (SCE) had deliberately installed steam generators already known to be bad. Allegations circulating the internet pointed to a “flawed design by Southern California Edison” and revealed a lack of clarity in the design process for such equipment. SCE quickly and strongly responded to the allegations.

Allegations in this matter made by Friends of the Earth (FOE) turned out to be, in fact, complete falsehoods. So it might be best to examine some of the facts surrounding this case and, as one recent San Diego Union Tribune op-ed piece hinted, “let the experts figure it out.”

RSGs and the Process of Replacement

RSG stands for “Replacement Steam Generator,” and the mystery in the public eye surrounding this process seems only to be growing.

In 2004, the owners of SONGS signed a contract with Mitsubishi to build four RSG’s for the two reactor plants on site. The San Onofre nuclear plants were originally built by Combustion Engineering (CE), which was merged out of existence some years back (Westinghouse is now essentially the lineal descendant). SCE chose to contract with Mitsubishi, which had been manufacturing steam generators of various types since 1970, to fabricate steam generators for the plants.

In this process, SCE provided to Mitsubishi a set of specifications—design standards to which the equipment had to adhere—for the steam generators. The specifications address not just size and weight, but a number of more involved details, such as desired materials. Mitsubishi then began work on a custom design for these plants based on the specifications. Mitsubishi used as a reference design steam generators it had built as RSGs for Fort Calhoun Nuclear Generating Station—also a Combustion Engineering plant, but smaller than San Onofre. A typical steam generator from a CE plant is seen below.

In the original conception of pressurized water reactor plants, the replacement of steam generators was not intended. In these old designs, however, deficiencies became apparent after some time in operation (which varied widely depending on the plant and particular design), so replacement of these massive pieces of equipment had to be considered. In some cases, such as Trojan Nuclear Power Plant in Oregon, replacement was required, but instead the plant shut down permanently and was dismantled when the cost structure and public opinion went against them. This example has not been the norm; and in fact many plants have replaced steam generators.

The original reactor vendors are not using the same facilities or contracts they did when the plants were newly built. The downsizing of the nuclear manufacturing complex after a new construction sales dropoff in the late 1970s led toward an almost wholesale outsourcing of RSG construction today. For example, since Westinghouse ended fabricating RSGs in the USA, it has used ENSA (Spain), Ansaldo (Italy) and Doosan (South Korea) as subcontractors for RSGs, while other RSGs have been supplied to US utilities by AREVA and Mitsubishi. A counter example to this trend is Babcock & Wilcox, which has a contract to replace Davis-Besse’s steam generators this year, as well as a contract for OEM replacements at TVA’s uncompleted Bellefonte units.

In the earliest steam generator replacements, only parts of the steam generators were replaced, but eventually entire units began to be fabricated. Eventually, as with any technology, improvements were made in design, and RSGs began to be fabricated with the same new, improved materials—such as Inconel-690 tubes—and techniques that were being employed in steam generators being fabricated for entirely brand-new reactor plants. Replacing steam generators gave operators an opportunity to incorporate both better materials and better designs; the possibility of uprating could also be realized if more heat transfer area were available in the RSGs. The NRC, recognizing the need to ensure safety with this as with every other practice in the industry, requires that replacement steam generators comply with a strict code that dictates what can, and cannot, be changed—and requires license amendments be applied for and approved when needed.

The above process, as described, is fully what occurred at San Onofre: SCE provided specifications to MHI, which then completed detailed design and fabrication of the steam generators.

Design Problems

In October 2012, after discovery of the issues leading to San Onofre’s RSG failure, MHI revealed it had made errors in computer analysis of the steam generator design. An SCE release provided to this author last October contains the following statement:

The Nuclear Regulatory Commission (NRC) determined that computer modeling used during the design phase by the manufacturer, Mitsubishi Heavy Industries, underpredicted the thermal hydraulic conditions in the steam generators which contributed to the unstable tube vibration. The unstable tube vibration caused the unexpected wear in the steam generators.

As we are now aware, this is only a part of the story. The phenomenon behind the vibration is called Fluid Elastic Instability (FEI). The real problem that allowed FEI to cause vibration serious enough to wear through tubes has to do much more with fundamental design assumptions and then, later, actual fabrication.

Reading of the linked MHI documents reveals clearly that the problem is partly theoretical, partly physical. On the one hand, an assumption in force in steam generator design industry-wide has held that “if out of plane FEI is prevented by design, in-plane FEI can not occur.” This has been proven wrong—at least in the San Onofre steam generators—although it must be stated clearly that this event at San Onofre is the first confirmed occurrence of in-plane FEI known in the industry.

We also see in the report (again, quite clearly) that the design of the Anti-Vibration Bars, which restrain the U-tubes, was slightly modified—and was thought to be improved—in Unit 3. What actually happened was that making the parts to finer (closer) tolerances reduced their contact force—and thus their ability to restrain the U-tubes—and helped lead to the motion-related impact wear.

Public Relations, and Events Outside Regulatory Action

As might be expected, continuous attention is given this situation by the NRC, which has held numerous meetings, inspections, and public hearings on this issue. The NRC is tasked with ensuring that the plant is safely operated and that it meets all technical requirements. The NRC certainly appears to be solidly on the job, given the sheer number of Requests for Additional Information (RAIs) that it has issued.

Politics has also become an integral part of this story. Senator Boxer sent a letter to the NRC stating that she had proof that MHI and SCE knew that the equipment was flawed. The letter was issued prior to any release, or public analysis, of the MHI documents.

In her letter, Boxer “calls on the NRC to promptly initiate an investigation” in the midst of what surely must be one of the most deeply technical investigations in NRC history—or in the history of the manufacture of steam generators. This clearly reveals a lack of perspective on where the MHI report falls in the path between discovery of the issues and development of a resolution.

In response to this ongoing situation, SCE yesterday issued a press release in which Pete Dietrich, SCE Senior VP and Chief Nuclear Officer, states:

The anti-nuclear activists have called the MHI report a ‘bombshell’ which couldn’t be further from the truth …. In fact, the MHI letter explains that SCE and MHI rejected the proposed design changes referenced in the evaluation because those changes were either unnecessary, didn’t achieve objectives or would have adverse safety consequences.

Our decisions were grounded in our commitment to safety. SCE did not, and would never install steam generators that it believed would impact public safety or impair reliability.

SCE goes on to state, “The MHI letter specifically confirms that at the time the replacement steam generators were designed, MHI and SCE believed that {excerpt from MHI report} ‘the replacement steam generators had greater margin against U-bend tube vibration and wear than other similar steam generators’.”

In the release, the Nuclear Energy Institute’s Scott Peterson adds that claims by anti-nuclear activist group Friends of the Earth (whose anti-nuclear creed is clearly stated on its home web page) are part of a campaign of moving “from plant to plant with the goal of shutting them down.” Pointing out the cherry-picked statements that both Senator Boxer and FOE are trying to posit as the ‘proof’ of wrongdoing of SCE, Peterson says: “Not providing proper context for these statements incorrectly changes the meaning and intent of engineering and industry practices cited in the report, and it misleads the public and policymakers.”

What’s Next?

This author spoke to SCE’s Jennifer Manfre yesterday about where this continuously evolving situation is headed. SCE would like to test operate Unit 2 at a 70% power limit for five months, followed by another complete RSG inspection, to assess if the calculational determination that FEI will be avoided here is demonstrated in operation. Manfre stated that this 70% limit is “very conservative—we set a limit for avoiding FEI, and then set a new arbitrary limit below that to ensure safety, as is always our priority.”

NRC has raised some questions regarding the limit and has asked SCE to be able to demonstrate that the plant is actually safe at 100% power during any of this 70% testing which, as Manfre points out, “goes to the technical specifications for the plant.” Manfre relates that SCE is preparing to submit, shortly, to NRC its Operational Assessment showing that the plant is indeed safe at 70% and also at 100% for this testing, saying “we essentially did both, to satisfy NRC and technical specifications.”

Manfre also clearly pointed out that the role of SCE in the RSG process is essentially that of being a customer with a required set of specifications, to which a detailed design is completed by a vendor (in this case, Mitsubishi). SCE did take part in some of the design process (for example, the design of the AVBs) but is not responsible for the overall design of the RSGs. Mitsubishi, who is responsible, has already begun warranty payments to SCE.

When Manfre was asked to speculate as to what a final resolution to this problem might look like—and was offered examples of a new operating license at a lower power rating to avoid FEI, or physical repairs to the steam generators to allow the full presently-rated power rating—she said “we’re not even close to that yet; we need to get through this period of testing.” Anyone in the nuclear industry (and, it might be added, many other industries) can relate to the need to conduct operational testing and analysis before selecting final operational fixes to a complicated technical and physical problem which involves public safety. Boeing’s problems with the 787 Dreamliner battery fire problem comes to mind as a timely parallel—as does the FAA’s handling of the situation.

Quite clearly with the voluntary release of the MHI documents, the process of investigation has unparalleled transparency for this sort of highly technical matter. In a February 26 SCE press release, Dietrich says that “this question and answer process is an important part of safety-based technical solutions in the nuclear industry, and it strengthens our ability to communicate to stakeholders the safety principles and proven industry operating experience that the Unit 2 restart plan was built upon,” in reference to the open nature of the NRC Request for Additional Information Process. The latest MHI release builds upon this process.

This open process between plant operator and Federal regulator has now been added to—or, depending on point of view, detracted from—by inclusion in the public domain of releases of sections of the MHI documents taken out of context. Dietrich, from yesterday’s SCE press release:

As with all engineering evaluations, the MHI letter and report describe a technical evaluation process and need to be read in their entirety to understand the conclusions reached …. The activists are taking portions of paragraphs and sentences out of context, and using them as the basis of their allegations that SCE knew of design defects when the generators were installed, but failed to make changes to avoid licensing requirements. That is untrue.

Manfre also relates that another ‘next step’ will be the impending full cost summation of the entire RSG process to the California Public Utilities Commission (PUC). The California PUC is under great pressure politically and must demonstrate that all rate impacts are fair and reasonable. She also points out an upcoming Atomic Safety & Licensing Board hearing covering the scope of the required license amendments.

All of the developing actions and public Federal regulatory hearings can be found on the NRC’s dedicated San Onofre pages. Developments and press releases from Southern California Edison on this situation can be found on its own dedicated SONGS website.

[Illustrations of San Onofre Nuclear Generating Station courtesy Southern California Edison]

_________________________________

→ 3 Comments

Posted in News, Nuclear Regulatory Commission, San Onofre Nuclear Generating Station

Tagged Fluid Elastic Instability, San Onofre, San Onofre Nuclear Generating Station, steam generators, Will Davis

Preparing to restart: Tsunami safety measures at Japanese nuclear power stations

Posted on February 28, 2013 by Will Davis| 2 Comments

By Will Davis

The approach of the second anniversary of the Great East Japan Earthquake of March 2011 finds nuclear energy in Japan at a crossroads. After the quake and resulting tsunami, the nuclear plants in Japan that did not shut down immediately eventually all had to shut down for their required, scheduled outages. Political pressures, for the most part, prevented any near-term chance of any of them restarting, it seemed at the time. When Tomari Unit 3 shut down in May 2012, Japan found itself with not one single operating nuclear power plant for the first time in decades. Since that time, only two nuclear units have restarted—Ohi Units 3 and 4 in July 2012. Other plants, rumored to be “next” to start up, have still not started up, although they may soon. The question that springs to mind is naturally, “When will the majority of the plants be allowed to restart?” The more insightful question, though, is, “What will have to be done in order to allow any plant to restart?” And how can we tell which will start first—is there any clue present now? Yes, there is.

Continued debate rages about the possibility of active faults being located beneath a number of plants—perhaps the most widely discussed being Tsuruga. For the plants experiencing this problem, restart is highly problematic—and highly politically charged. For the informed, it’s also no safe bet.

Other nuclear plants, however, are emerging as the “sure bets” of owner-operators who are pushing massive amounts of time, money, and material into them, preparing for restart whenever the Japanese government and the new Nuclear Regulation Authority (NRA) allows it. The sheer amount of work being put into two of these is our focus today as we look forward to the time when Japan will return to generating a fair portion of its electric power from nuclear energy.

The photograph above was taken in April 2011 by the Japanese Maritime Self Defense Force, and clearly shows the debris and tsunami damage on the sea side of Units 1 through 4 of the Fukushima Daiichi nuclear powers station. This damage—physical derangement of installed equipment, and water inundation of facilities—was the direct cause of the accident. (The tsunami was preceded by a massive earthquake that caused enormous power outages due to transmission line damage and reactor plant shutdowns, but did not lead to unusual events at the plant in and of itself.) This photo makes fairly obvious the damage, but perhaps not as obvious the height of the water to be defended against.

Kashiwazaki-Kariwa

At right, we see Tokyo Electric Power Company’s (TEOCO) Kashiwazaki-Kariwa nuclear power station. This station has for many years been the largest (highest total output) nuclear station in the world, with seven reactor plants on one site. TEPCO (also owner of Fukushima Daiichi and Fukushima Daini) has been pouring money and material into facilities on and around this site in order to prepare it for certification to start up.

It must be said right off that the most important tsunami defense this plant has is its location; it’s on the opposite coast from Fukushima Daiichi and Fukushima Daini, and according to TEPCO the undersea faulting that does exist west of Japan is not thought to be able to generate tsunami at all. Even so, TEPCO has implemented massive works at the site; click on the following link to see a detailed video of the size and scope of the project. (The videos linked in this ANS Nuclear Cafe article are detailed and impressive, and are “must see” to understand the real scope of the efforts being exerted.)

TEPCO Kashiwazaki-Kariwa Tsunami Protection Enhancements

The provision of seawall protection is fully and redundantly backed up by the protection placed around the reactor buildings in TEPCO’s protection scheme; at Units 1 through 4, a large new artificial sea wall defending against even 15-meter tsunami is backed up by protection of the reactor buildings themselves by new added enclosures, also proof against 15 meters of water. All doors on the reactor buildings will be water-tight, and all openings below 15 meters will be shielded with covers to prevent water entry. On the other hand, the three newer units—Units 5, 6, and 7—already sit on higher ground and thus don’t require as high of a new seawall; further, these units were built having no low openings that water may enter through below 15 meters.

Also notable in the video is the installation of fixed structure to allow portable generating and pumping equipment to supply plant cooling needs in case of long-term station blackout (SBO) and even in the event of serious damage to the site. The portable equipment is located at a high elevation near the plant; it includes mobile generating trucks (using gas turbine engines instead of diesel engines), diesel powered skid-mounted fire pumps, fire engines, and mobile units containing water-to-air heat exchangers. According to TEPCO, the SBO/loss of ultimate heat sink survival time for this site after an earthquake and tsunami is said to be 196 days as a result of the additions and enhancements.

Construction of this new protection and provision of the new equipment is proceeding at a rapid pace; it is expected to be completed this year. A further detailed video, also well worth watching, shows more of the construction of the protection and its progress as of the middle of last year.

Tsunami Protection Enhancements at Kashiwazaki-Kariwa: Progress, June 2012

The Kasiwazaki-Kariwa station has undergone a complete stress test at Units 1 and 7 (which should cover most eventualities at other units, generally), although it seems clear now that the NRA might be inclined to develop further requirements; the final result of NRA’s decision making is due mid-year. For what it is worth, TEPCO believes that the plant is also immune, after the implementation of seismic enhancements, even to very large earthquake accelerations (which is supported by the fact that none of the reports concerning Fukushima Daiichi has so far proven out any of the assertions that the quake itself led to crippling or even problematic system damage.) A TEPCO video covering the stress test can be seen here. The video describes the stress test steps clearly for anyone, even with no knowledge of nuclear energy. It is important to add though that the stress test video portion describing the spent fuel pool “cliff edge” for Unit 1 is actually describing the effect should water overflow the new, outer 15 meter tsunami sea wall and get inside the site.

Overall, the safety measures TEPCO is implementing at this plant are impressive, on a grand scale; comparatively, absolutely nothing of the sort has been done at its other undamaged nuclear power station, Fukushima Daini. This most likely reflects the Fukushima prefectural government’s repeated assertions that no nuclear plant will operate in its territory ever again—dooming the four reactor plants at Fukushima Daini and the two undamaged units (Units 5 and 6) at Fukushima Daiichi. Judging all advance indications (including TEPCO’s investments and the political atmosphere) if any of TEPCO’s nuclear stations would ever restart, Kashiwazaki-Kariwa would be first.

Hamaoka

Whereas it’s reported that TEPCO has spent as much as 70 billion Yen on enhancements at Kashiwazaki-Kariwa, Chubu Electric Power Company has spent 100 million yen at its five-reactor Hamaoka nuclear power station, and has increased the estimated total amount required to 140 billion yen. It has also pushed the expected completion of physical construction/equipment acquisition back an entire year from the originally expected date, to July 2013. This nuclear station is located on the same side of the country as Fukushima, but is well to the south.

At right, Hamaoka nuclear power station, courtesy Chubu Electric Power Company. This station has five nuclear reactor plants; Units 1 and 2, nearest the right of the photo, are undergoing decommissioning, while the other three units are expected to operate in the future.

Preparations at Hamaoka, which comprise over 30 different construction projects, mirror those underway at TEPCO’s plant quite closely, through the provision of sea-side protection, backup power generating, and water pumping equipment, and of course all of the training required to implement the new procedures (using new and unfamiliar equipment). As stated by Chubu, the improvements to the site were begun before a full understanding of the experience at Fukushima Daiichi was widely known. The 40-billion-yen increase in cost, to be spread over several years, comes from alterations to the protection plan that were pointed up from real experience at Fukushima. For example, the design of reactor building doors to be fitted at Hamaoka was changed to a swinging design of watertight door to reduce the time required to shut and secure the doors. It has become clear that in emergency and disaster situations, minutes and seconds count.

Chubu Electric has also produced an excellent video (also in English) quite similar to those by TEPCO, showing the enhancements specific to its Hamaoka nuclear power station site. Click here to see it. Chubu offers the public an excellent PDF file report titled “Tsunami Countermeasures at Hamaoka Nuclear Power Station“ on site protection enhancements that is quite minute in detail.

In December 2012, Chubu Electric also announced additional “Severe Accident Countermeasures” to be taken at Hamaoka that are intended to do three things: prevent an uncontrolled radiological discharge (during an accident), prevent damage to the containment vessels of the reactor plants, and provide increased DC power availability. Specific actions called out included installation of filtered PCV vents, installation of water spray lines in the reactor vessel pedestals (to ensure debris retention), enhanced containment spray (to knock down airborne contamination in event of release inside containment), special cooling for the PCV head (through which it is now believed that hydrogen gas escaped into the reactor buildings at Fukushima Daiichi), provision of upgraded storage batteries, and provision of alternative (and mobile) heat exchanger equipment for core cooling. These are all enhancements directly developed as a result of accident sequence events and site complications known to have occurred during the accident progression at Fukushima Daiichi.

The plans, and the future

Parallels between the TEPCO and Chubu Electric plans are fairly obvious—both are spending large amounts of money on presently shut down nuclear stations of large generating capacity in order to help ensure that they are allowed to restart. When they do, the companies will begin to earn revenue to pay for the disaster enhancements (and, in the case of TEPCO, to pay for many other things, including decommissioning Fukushima Daiichi and, in all probability, eventually Fukushima Daini) and in addition will help restart Japan’s economy. Both companies are relying on a complex mix of physical enhancements to site perimeters, reactor plants, and interconnecting infrastructure (such as new remote wires and pipes). Both are investing heavily in mobile equipment of many types. While the training required to integrate all of this new equipment hasn’t specifically been mentioned, we know that it is exceedingly complicated and will be very time-consuming to get right. Both companies continue to conduct drills on the use of this equipment, with site-wide timed ‘disaster scenarios.’

Another parallel that is important not to miss is that much of what TEPCO and Chubu Electric are doing is quite similar to the FLEX approach backed by the Nuclear Energy Institute and owner-operators in the United States.

One contrast between the Kashiwazaki-Kariwa and Hamaoka projects is that whereas the TEPCO plant, on the west coast, is being given 15m–high tsunami wall protection, Hamaoka, which is on the opposite coast, is being given 18m tsunami protection. This reflects the seismic environment of Japan, which as previously stated is much more likely to experience large tsunami on the eastern coastline of the nation.

It seems likely, given the Japanese public’s new well-publicized suspicion of nuclear energy (and particularly the Japanese government interrelations with the Japanese nuclear industry), that restarting plants in Japan will only come with a solid yet transparent combination of physical site protections, emergency backup plans, solid regulation and enforcement, and divorce of the regulator from industry interests. All of these are underway now, and as we’ve seen, at least two of the utilities owning nuclear plants are heavily investing on a nuclear future for Japan, even if the face it presents is very largely different to that it presented to a pre-Fukushima world.

_______________________________

→ 2 Comments

Posted in accident, Fukushima

Tagged Will Davis

The Atlantic Generating Station

Posted on January 3, 2013 by Will Davis| 22 Comments

Recent announcements and news stories about a Russian project to build a floating and essentially portable nuclear power plant have been variously tabbed with the heading “new.” The idea of a floating, mobile nuclear plant (which is not self-propelled and not a ship) is indeed not new—the nuclear barge STURGIS, itself a converted Liberty Ship, served as a power source for the Panama Canal for many years, beginning back in 1967. The new Russian plants bring extra excitement because they are classed, properly, in the now-popular small modular reactor plant category, having been based on true seagoing designs. This, of course, hints at the fact that their output will not approach that of any of the large, conventional nuclear plants familiar today.

For the historian, the question might then come to mind as to what the largest nuclear plants ever seriously considered for construction to such a design were. The answer is very simply this: Full size, commercial nuclear plants in the normal (>1000 MWe) range common today. While the plants weren’t to have been fully mobile in the same sense, they were to have been barge–mounted and would have remained floating while in operation.

In the late 1960s, Public Service Electric & Gas (New Jersey) began to invest very heavily in nuclear energy. The company bought a major investment in Philadelphia Electric’s Peach Bottom expansion, and also began to order units of its own. In 1966, PSE&G ordered Salem Unit 1, followed in 1967 by Salem Unit 2 (both from Westinghouse); in 1969, PSE&G awarded a contract to General Electric for its Newbold Island nuclear station (two units), which eventually would be cancelled for siting reasons; however, with that cancellation, simultaneously the project was moved next to Salem to be built as Hope Creek. According to PSE&G literature of the period, because of increasing worry about the thermal effects (waste heat) of nuclear plants, it decided to make its next order for a nuclear plant a bold, radical step; it decided to contract with Westinghouse to construct nuclear plants essentially at sea, in a man-made structure and mounted on floating barges.

The site eventually chosen after some consideration and study was as shown here in an original advertising illustration from a PSE&G brochure on the project. The caption reads: “The proposed offshore site is 2.8 miles out in the ocean, off Little Egg Inlet, and approximately 12 miles north of Atlantic City.” The location chosen kept the nuclear station out of major shipping lanes.

The site itself would have been prepared (with a breakwater surrounding it) including two moored, side-by-side nuclear power plants, separate from each other but identical and which would have been mounted on gigantic rectangular barge structures. According to the PSE&G brochure, “The Atlantic Generating Station,” the construction process would have been as follows:

“The breakwater will be the largest and strongest structure ever built in the ocean. First, concrete caissons will be floated to the site, sunk, and filled with sand and gravel. Next, thousands of tons of rock will be brought by barge to create the artificial reef, within which the plants will be moored. The mound facing of the reef will consist of sand, gravel, and stones topped by an armor of interlocking pre-cast concrete units called ‘dolos.’ A typical large dolos weighs 42 tons and measures 20 by 20 feet. Approximately 70,000 of these dolosse, in various sizes, will be placed on the breakwater.”

The structure and the plants were designed to survive 43 foot waves, sustained (continuous) hurricane winds of 156 miles per hour and tornado winds of 300 MPH.

Above, cross-section view of the installation as planned. (Our apologies for the slight imperfections in some of the illustrations, which are contained in vintage materials, are not always printed perfectly, and which are often printed across the center staple fold.) Below, an artists’ illustration of the plant, whose official name was in fact the Atlantic Generating Station, from the air.

The two plants that were to become the Atlantic Generating Station (AGS) were first announced in 1971, according to WASH 1174-71, but were not named at that time, nor was a location specified. In September 1972, according to the Atomic Industrial Forum (now the Nuclear Energy Insitute) report “Historical Profile of U.S. Nuclear Power Development,” 1985, the two plants were officially ordered from Westinghouse as Atlantic-1 and -2. (The reactor plants were to have been 1150-MWe four-loop PWRs.) The plants were to have been built at a wholly new dedicated facility in Jacksonville, Florida, as a part of a joint Westinghouse–Tenneco operation known as “Offshore Power Systems.”

PSE&G had printed, in its public relations materials of the time, that it intended to rapidly increase its nuclear generating assets. From a 1976 brochure on Hope Creek: “To prepare for the coming ‘electric economy’ when electricity will play an even greater role in our daily lives, PSE&G is relying on nuclear energy. From now until the end of this century, all new major generating units will be nuclear. It is our intention to phase out our oil and coal burning plants and eventually have approximately 50 percent or more of our electric capacity in the nuclear stations we share with other utilities. This nuclear capacity will provide approximately 75% of our energy needs by 1990.”

To that end, in November, 1973, PSE&G ordered two further nuclear units of the same type as ordered previously as Atlantic-1 and -2. While the AIF document previously mentioned does not give a plant name or site for these, a later Energy Information Administration/Department of Energy document identifies these plants as Atlantic-3 and -4.

The nuclear plants would have been mounted on barges approximately 400 feet square. The draft of the nuclear plant barge structures (that is, the depth to which they extended underwater) would have been roughly 30 feet; the breakwater/reef enclosure would have had a further 10 feet of clearance under the plants for water flow. In an interesting nod toward today’s AP1000 plant and its modular construction, PSE&G said about the AGS units that the shipyard fabrication plan “allows for assembly line production techniques, as well as standardization of design and licensing procedures—which will result in reduced costs and planning lead times.” Heavy underwater cables, instead of high tension towers, would have connected the plants to the grid. A shore base would have been built, to shuttle workers to and from the AGS and to station repair parts, consumables, and any other requirements for the nuclear station several miles out to sea.

Above: “Artist’s conception depicts how the plant will appear on a clear day to a person standing on the nearest beach.” The illustration is meant to dispel the fears that the plant would be an eyesore.

Of course, we all know today how this overall plan played out. PSE&G did not experience nearly the expected growth in electric power demand that it had predicted. While a 1976 PSE&G brochure on Hope Creek also prominently features the Atlantic Generating Station, a 1977 brochure on Salem does not mention it at all. In 1978, PSE&G cancelled all four units ordered for its offshore nuclear power station program, and the AGS project died immediately. (Work did continue on the other plants mentioned earlier, but not even all of these were finished; work on Hope Creek-2 lagged, and that plant was finally cancelled in 1981, leaving Hope Creek-1 a single unit.)

As we can see, a large amount of challenging engineering and construction would have been required to complete the Atlantic Generating Station. One wonders if such a project could survive today’s regulatory environment—to say nothing of clearing approval by a utility’s ownership when the extra cost of constructing the artificial reef type breakwater and shore-based support infrastructure is considered. The best guess for both right off the bat is “probably not,” meaning that the Atlantic Generating Station was probably the closest we’ll ever get to a full-scale commercial nuclear plant situated well off shore.

Sources of information and illustrations: Various original PSE&G brochures—”The Atlantic Generating Station” (undated), “Hope Creek Generating Station” (8/76), “The Salem Generating Station” (4/77), “PSE&G: Nuclear Energy” (6/85). Also, “The Nuclear Industry 1971″–WASH 1174-71, U.S. Atomic Energy Commission. “Historical Profile of U.S. Nuclear Power Development,” Atomic Industrial Forum 1985. “Nuclear Plant Cancellations: Causes, Costs and Consequences,” US EIA/DOE 1983. All materials in Will Davis’ library.

For more on this topic, particularly the plant construction end of the project, see Rod Adams’ article from 1996 on Atomic Insights.

_________________________________

Will Davis is a consultant to, and writer for, the American Nuclear Society. In addition to this, Davis is on the Board of Directors of PopAtomic Studios, is a contributing author for Fuel Cycle Week, and also writes his own blog Atomic Power Review. Davis is a former US Navy Reactor Operator, qualified on S8G and S5W plants.

→ 22 Comments

Posted in Nuclear History, nuclear technology, PWRs, Westinghouse

Tagged nuclear history, Will Davis

Clinch River Site will once again lead nuclear development

Posted on November 29, 2012 by Will Davis| 1 Comment

By Will Davis

(Above, Westinghouse artwork depicting the Clinch River Breeder Reactor plant as envisaged in November 1973.)

The Department of Energy announced recently that it would award the first of (potentially) two blocks of grant money for small modular reactor (SMR) development to Babcock & Wilcox, Bechtel Corporation, and the Tennessee Valley Authority. The funds would be used for construction of a new SMR–powered reactor plant at the former Clinch River Breeder Reactor (CRBR) site in Oak Ridge, Tennessee—a site that once shined as the future of nuclear energy in the United States.

Decades ago, the Liquid Metal Fast Breeder Reactor (LMFBR) program, originally begun by the Atomic Energy Commission, turned into a real-world project in 1972 when the AEC signed the first Memorandum of Understanding with TVA, Project Management Corporation, Commonwealth Edison, and Breeder Reactor Corporation – to build what would become known as the CRBR plant. Work quickly advanced to include a number of reactor vendors (Westinghouse as lead reactor manufacturer, along with General Electric and Atomics International) and a giant consortium of 753 utility companies nationwide, as well as many other vendors and consultants. Project costs escalated, and in 1977 the Carter administration decided to terminate the licensing activity and attempted to kill the project. The CRBR project went on in semi-limbo for years, with much hardware being constructed. Finally, after a brief attempt in 1983 to find ways to increase outside funding for the project, it was cancelled—with over 70 percent of the equipment either delivered or ordered, site preparation work underway, licensing activity nearly completed, and environmental hearings completed (DOE-NE-0050, March 1983.)

When the breeder project was launched, the liquid metal–cooled breeder reactor seemed very much the path to the future for nuclear energy, in order to close the fuel cycle. Now, the SMR seems the path to the future, to provide industrial power and process steam, even for off-grid use. It’s supremely fitting that the Clinch River site—just green field now, but where the “old future” of nuclear energy died—will see the launch of the “new future.” In order to help close the historical circle, let’s take a look at some of the hardware actually constructed for the CRBR project—but never used. We’ve already seen the first exterior concept for the plant above; we’ll see the final one later on.

Above, the reactor vessel for the CRBR, pictured at Babcock & Wilcox’s facility in Mount Vernon, Indiana, as seen in a Westinghouse CRBR status report from 1981. The special J-shaped rig or mount was designed to both transport and help erect the vessel at the time of installation. Cost of this piece of equipment with core support structure was about $27.7 million. The core support was fabricated by Allis-Chalmers.

Above, flow diagram for the CRBR–sodium in the primary and intermediate loops (3 double loops total) with steam/water in the conventional manner in the final cycle. The odd-looking shape of the steam generators and superheaters in the diagram is no mistake, as we’re about to see.

Above, CRBR “evaporator” or steam generator delivered from Atomics International for testing. Both the primary loops and intermediate loops were to use very large electric pumps to move the liquid sodium, which we’ll see below.

Above, a primary loop sodium pump under test at the Byron Jackson Division of Borg-Warner Corporation, as seen in a Westinghouse update on the CRBR project from 1981 (the same photo is duplicated in the 1982 report).

The CRBR project had its own internal newsletter; above, the cover of the December 1978 “Clinch River Currents.” Below is the text from the cover:

“The CRBRP’s in and ex-containment primary sodium storage tanks are complete and will be shipped by barge to Oak Ridge when needed. The three tanks have been purged, sandblasted and painted and are now in storage at ITO Corporation of Ameriport, Camden, New Jersey.

These tanks for the CRBRP were built at the Joseph Oat Company, Camden, New Jersey, under a subcontract from Atomics International. The materials used were ASME SA-515 and SA-516 carbon steel plate, and SA-105 for the nozzle forgings. Single piece spun heads were used in fabricating the tanks.

The contract was awarded in October 1976, and fabrication started in February 1977. The 23-foot-long in-containment tank was completed in August 1978 and the two 32-foot-long ex-containment tanks shown here were completed in September 1978. Each of the three tanks is 18 feet in diameter.”

In that same December 1978 issue we find a number of illustrations and details about completion of the in-vessel fuel transfer machine, illustrated below with original caption material included.

“Four years of design work and over a year of fabrication and assembly by Atomics International Division, Rockwell International, Canoga Park, California, have culminated in completion of the two subassemblies of the in-vessel transfer machine. The next step will be final assembly, followed by an integrated checkout of the unit in air in February. Following completion of this phase, the unit will be turned over to the Energy Technology Engineering Center nearby in Santa Susana, California, for testing in sodium. Turnover is scheduled for May 1979.

The $2.3 million apparatus will be used to transfer fuel inside the reactor vessel during refueling. Mounted on the smallest of three eccentric rotating plugs of the reactor vessel head, it will be capable of locating itself over any removable element of the core, picking it up with a straight pull and transferring it to a temporary storage location inside the reactor vessel. It will also pick up replacement elements from the storage location and place them in the proper position in the core. The triple rotating plug locating concept, also used by West Germany in the SNR 300, is the first such head design used in a US designed LMFBR. Prior rotating head concepts in the US were employed on EBR II [Experimental Breeder Reactor II] and FFTF [Fast Flux Test Facility] but consisted of only two heads and a cantilevered in-vessel fuel handling device…”

Below, the reactor vessel head assembled for testing; the eccentric plugs and gears can clearly be made out.

The design layout for the plant changed a number of times as improvements were made. Below, the final layout as found in 1981–1982 Westinghouse status reports, and which was fairly widely released. This was the final plant configuration.

As we have seen, the CRBR was never built. The equipment ordered was laid up or disposed of, and the work force scattered; the site returned to disuse. The promise of a new and different future for nuclear energy never did die, though—it has taken on new faces from time to time since then, none of which has really reached the hardware stage. Now, at last, the Clinch River site will finally see construction and operation of a nuclear power plant, fulfilling its promise. While the design and appearance of the Generation mPower SMR plant will be vastly different from that envisaged for the CRBR, it’s fitting that it is because the look of the future of nuclear energy has also changed that much in the intervening quarter century.

One last illustration; below we see the cover of the January 1979 Clinch River Currents, whose headline announces “First Major CRBRP Hardware Delivered to Oak Ridge”—this was a protected water storage tank manufactured by Process Equipment Company, Brockton, Massachusetts, and three primary sodium system cold leg check valves (inset) from Foster Wheeler in Mountaintop, Pennsylvania.

(Illustrations from Westinghouse, CRBR management reports; Clinch River Currents illustrations and text, and both CRBR plant external illustrations from Will Davis collection.)

____________________

Will Davis is a former US Navy Reactor Operator, qualified on S8G and S5W reactor plants. Davis performs Social Media services for ANS under contract, writes for ANS Nuclear Cafe as well as for Fuel Cycle Week, and also writes his own Atomic Power Review blog.

→ 1 Comment

Posted in News, reactor designs, Small modular reactors, Tennessee Valley Authority

Tagged babcock & wilcox, bechtel, clinch river breeder reactor, smr, TVA, Will Davis

ANS Winter Meeting 2012: Nuclear Technology Expo

Posted on November 12, 2012 by Will Davis| Comments Off

The first of a series on people and events at the 2012 American Nuclear Society Winter Meeting

By Will Davis

The evening of November 11 saw the opening of the latest ANS Nuclear Technology Expo, in the spacious convention facility housing the ANS 2012 Winter Meeting at the Town & Country Resort in San Diego, California. The event did not disappoint.

The Expo opened with this evening’s ANS President’s Reception, with food and beverage of a high caliber provided for attendees. The turnout was shoulder to shoulder for much of the floor space in the exhibit area.

Over 50 groups were represented in the Expo; the majority were vendors, while some were regulatory or governmental bodies (the Nuclear Regulatory Commission, the International Atomic Energy Agency), national laboratories (Argonne, Idaho National Laboratory) and universities. Every kind of information was available either by brochure or via conversation with attendant representatives. The displays were all quite interesting, with many tailored directly to a very focused audience. One display was appealing to all audiences: a remote grappling arm with sensitivity sufficient to delicately manipulate a very thin-stemmed wine glass.

Vendors represented at the Expo included the large reactor vendor companies including Westinghouse, GE-Hitachi, and Areva, and many other companies whose services are more specialized, including I&C (instrumentation and control), measurement equipment, and engineering consulting. Remotely-controlled, rugged equipment used for decommissioning of nuclear power plants (in addition to general demolition) was represented as well.

Face to face conversation and networking are among the most valuable aspects of this expo event. I did not have the chance to thank Mimi Holland Limbach for her fine presentation at the ANS Annual Meeting in June that was so enjoyable—tonight gave me the opportunity to discuss and thank her in person. I conversed with many colleagues who I haven’t seen since June—that is, when they were not engrossed in deep conversation with other colleagues.

This author came away with, literally, a bag full of relevant, up-to-date technical material that will serve well in answering future questions asked by readers. And, yes, some really “cool” souvenirs—you’ve got to have something to bring back home for the family!

The Nuclear Technology Expo is open for two more days—this Monday and Tuesday. On Monday, hours are from 11:30 AM to 5:30 PM (opening with an ANS Attendee Luncheon until 1 PM), while on Tuesday the hours are from 10 AM through 2 PM. If you’re here in San Diego attending the Winter Meeting and didn’t have a ticket to tonight’s event, do find time in the next few days to explore the Expo. It’s well worth it.

Remember to follow events on the ANS Twitter account! Look for @ans_org. Tomorrow—the Opening Plenary Session, with live tweets. Hash tag #ANS12.

___________________________________

Will Davis is a former USN Reactor Operator qualified on S8G and S5W reactor plants; is a writer and social media consultant for ANS; writes for Fuel Cycle Week; and also writes at his own Atomic Power Review blog.

Comments Off

Posted in American Nuclear Society, Meetings

Tagged American Nuclear Society, meetings, technology expo, Will Davis

Spent Fuel Pool at Oyster Creek

Posted on October 29, 2012 by Will Davis| 5 Comments

By Will Davis

As the Eastern half of the United States falls under siege by Hurricane Sandy and combined weather fronts—which together are being termed ”Frankenstorm”—the nuclear community is targeted by nuclear opponents keen on capitalizing on this severe weather event. A recent piece quoting Arnold Gundersen asserts that Oyster Creek Nuclear Generating Station is facing serious problems should it lose offsite power, saying essentially that the plant will be unable to provide cooling for the spent fuel in its spent fuel pool.

This allegation is without merit.

This document—a memorandum from the Nuclear Regulatory Commission staff to the then-operator of Oyster Creek—spells out the spent fuel pool (SFP) cooling arrangements in place back in 2000. It includes the following description of the SFP cooling arrangements:

Make up water to the SFP is normally provided by the condensate system from the condensate storage tank (CST) which has a nominal capacity of 525,000 gallons. The condensate pumps can provide 250 gallons per minute (gpm) with one pump operating or 420 gpm with two pumps. Additional makeup can be provided from the demineralized water storage tank (nominal capacity 30,000 gallons) by connecting the demineralized water transfer pumps to the SFP with hoses. The fire protection system can also provide makeup from the fire pond to the CST using the 2,000 gpm diesel driven fire pumps through a permanent connection.

The SFPCS {Spent Fuel Pool Cooling System} removes decay heat from fuel stored in the SFP through its associated heat exchangers to the reactor building closed cooling water (RBCCW) system. The SFP water is maintained within its TS limits by these systems. The SFPCS consists of two SFP pumps, two SFP shell and tube heat exchangers, two augmented fuel pool pumps, and one augmented fuel pool plate and frame heat exchanger. In addition, the SFPCS also includes interconnections with the condensate demineralizers and the condensate systems which filter and demineralize the SFP water as well as provide makeup water to the SFP. The SFPCS operates continuously to maintain the SFP water temperature at or below the Oyster Creek TS limit (maximum of 125 degrees Fahrenheit (F)).

As we can see, a total loss of offsite power (LOOP) scenario has clearly been considered—otherwise, diesel fire pumps would not have been mentioned.

Oyster Creek Nuclear Energy Facility

Plants designed to handle spent fuel pools during loss of offsite power

Oyster Creek, like all other operating U.S. nuclear plants, was built to design considerations (10 CFR 50 Appendix A) that set limits on design that includes the protection of spent fuel pool from events both man-made (operational) and natural. The plant has been designed to handle the full heat load of the spent fuel placed in the pool—even with a loss of offsite power.

Spent fuel pool cooling has received greater attention since the Fukushima Daiichi accident; during that accident and for some time after, many had wrongly assumed and asserted that the spent fuel pools were in dire condition. In fact, some even claimed that Fukushima Daiichi Unit 4 was going to collapse and that the spent fuel was going to trigger a cataclysm. Those allegations were refuted at the time, multiple times, and have been proven false.

Even though early post-Fukushima assumptions about spent fuel pools were overly unrealistic, the NRC has emphasized SFP cooling and level measurement as a part of its post-Fukushima action plan. Many experts and the Nuclear Energy Institute consider this approach sensible. NEI points out, however, via NEI Nuclear Notes that moving SFP actions to Tier 1 in no way implies that operating U.S. nuclear plants aren’t already safe. Read that post here.

The Safety Evaluation Report related to license renewal of Oyster Creek at the NRC contains the following information about Oyster Creek’s spent fuel cooling system:

The SFPCS (Spent Fuel Pool Cooling System) is designed for both normal and accident conditions of loss of offsite power coincident with a single active component failure. The augmented SFPCS is designed to provide a seismically qualified cooling loop capable of providing cooling during such conditions.

As if that were not enough:

Exelon – Oyster Creek Safety and Emergency Planning Fact Sheet

Clearly, there is provision for SFP cooling at Oyster Creek using two SFP systems—the one that was originally installed and an augmented system installed when the pool capacity was increased—and also it’s a fact that the plant, like all others in the path of the storm, is and has been well aware of the approach of this storm and has even more personnel (and NRC inspectors) on site than usual, making full preparation for any event. “Any event” includes extended loss of offsite power.

Oyster Creek has multiple cooling systems for spent fuel pool

UPDATE: Exelon has re-confirmed to the American Nuclear Society by telephone and e-mail that Oyster Creek does in fact have numerous, redundant cooling systems for the spent fuel including closed-loop and service water systems. Exelon tells us that if required, two locomotive–sized diesel engines are ready and standing by should offsite power be lost, to provide power to those two backup systems during the refueling outage should an extended LOOP scenario arise.

Exelon has, as expected by many, declared an Unusual Event at Oyster Creek due to the rising water levels. Below are excerpts from Exelon’s press release on this declaration (emphasis added):

Oyster Creek Generating Station Declares Unusual Event

Lowest of four NRC emergency action levels reached due to high water levels

Forked River , NJ (October 29, 2012) Exelon Nuclear declared an “Unusual Event” at Oyster Creek Generating Station at 7 p.m. today after water levels in the plant’s intake structure reached higher than normal levels.

This is an anticipated declaration required by procedures and is the result of Hurricane Sandy’s impact on the region. There is no challenge to the safety of the plant. Oyster Creek is currently shut down for planned maintenance and refueling.

Oyster Creek is a robust and fortified facility, capable of withstanding the most severe weather. When the storm was identified, operators performed a host of plant inspections to ensure that all safety systems were operational and that outside equipment and materials were properly secured.

An Unusual Event is the lowest of four emergency classifications established by the U.S. Nuclear Regulatory Commission. There is no danger to the public or plant employees associated with this declaration.

Exelon Nuclear has notified all appropriate federal, state and local emergency response officials of the Unusual Event.

Oyster Creek is about 60 miles east of Philadelphia in Ocean County, New Jersey. The plant produces 636 net megawatts of electricity at full power, enough electricity to supply 600,000 typical homes, the equivalent to all homes in Monmouth and Ocean counties combined.

_____________________________________

For more information

Below is a brief video interview with the Nuclear Energy Institute‘s Everett Redmond, director of Nonproliferation and Fuel Cycle Policy. He breaks down in straightforward language the purpose and design of spent fuel pools to store used fuel at nuclear energy facilities. This is a basic overview that does not address specific nuclear energy facilities.

_____________________________________

Will Davis is a writer and Social Media consultant for ANS, is a Contributing Reporter to Fuel Cycle Week, owns and writes the Atomic Power Review blog, and is a former US Navy Reactor Operator, qualified on S8G and S5W reactor plants.

ANS staff members also contributed to this report and compiled additional resources for readers.

→ 5 Comments

Posted in BWRs, Natural disasters, News, Nuclear events, Nuclear Regulatory Commission, spent fuel

Tagged Arnie Gunderson, Hurricane Sandy, Oyster Creek, spent fuel, spent fuel pools, Will Davis

The MTR—Gone now, but not forgotten

Posted on October 25, 2012 by Will Davis| 1 Comment

by Will Davis

Recently, Dr. Nicole Stricker of the Idaho National Laboratory sent a link for the following video to members of the ANS Social Media list.

INL Waste Video

The entire video is quite interesting, but my interest was tweaked during the time frame 3:23 to 3:28 in the video by what looked like a reactor vessel being tipped over during decommissioning of a nuclear facility; the voice-over at the time is talking about just that. A request for information revealed that the reactor shown at that moment in the video was the Materials Testing Reactor, or MTR.

I had known that the MTR had been long shut down, but was really unaware of its present status. The MTR has a place in nuclear history in the United States as the first widely available test reactor; according to The Atomic Energy Deskbook, the MTR was designed jointly by Oak Ridge and Argonne National Laboratories. Blaw-Knox acted as architect-engineer, and the plant was built by the Fluor Corporation.

Let’s let the words of Phillips Petroleum Company, which operated the MTR for the Atomic Energy Commission, describe the facility; they’re found in the booklet (in my collection) whose cover is reproduced below.

“The Materials Testing Reactor is a unique and versatile research tool. It was designed and constructed as a pioneering step in the development of high neutron intensity reactors with the primary purpose of providing facilities to test the effects of neutron bombardment on materials of interest in future reactor construction. It has neutron fluxes 10 to 100 times greater than those in other reactors. As a result, it can provide radiation at a very high dose rate and produce isotopes with higher specific activity than those now available from other sources.

The MTR is a thermal (slow) neutron reactor using uranium enriched in isotope U235 as fuel, ordinary water as both moderator and coolant, and beryllium as the reflector. It is designed to generate the heat equivalent of 30,000 kilowatts. Because of its high specific power, average neutron fluxes of 2 X 10^14 thermal neutrons per square centimeter per second and 5 X 10^13 fast neutrons per square centimeter per second are available. Peak thermal neutron fluxes of 5 X 10^14 neutrons per square centimeter per second exist in certain positions in the reactor.

The enriched uranium fuel is contained in an active core which is inside a lattice region 40 by 70 centimeters in area and 60 centimeters high (16 x 28 x 24 inches). It is surrounded by a 40 inch high reflector of beryllium pieces. Both lattice and reflector are enclosed in a 55 inch diameter aluminum tank which is extended by stainless steel sections above and below to form a 30 foot deep well which is closed top and bottom with heavy lead filled steel plugs. ….The reactor lattice and beryllium reflector are cooled by water flowing at a rate of 20,000 gallons per minute. This water enters near the top of the well at 100F and leaves near the bottom at 111F. The water is fed by gravity from a 170 foot high tank through the reactor tank to a vacuum spray evaporator system for cooling and degassing, then is pumped back to the tank.”

According to contemporary documents from Sylvania-Corning Nuclear Corporation in my collection, fuel elements for the MTR were “93% enriched uranium alloyed with aluminum, clad in aluminum, and formed into curved plates approximately 24″ long, 3″ wide and 1/16″ thick. The fuel element consists of nineteen such plates brazed into aluminum side plates to form a boxlike assembly approximately 3″ x 3″ in cross-section. Aluminum adaptors are welded to the ends of the fuel element. Each element contains 200 grams of U235 and normally 25 such elements fuel the reactor.”

In addition to offering irradiation services directly using the reactor, the MTR also offered gamma irradiation using spent fuel as described below by Phillips Petroleum:

“The gamma field is provided by used MTR fuel elements, which are stored under water until they have cooled sufficiently to be transferred to the chemical processing plant for recovery of U235.” At left, the original MTR canal where gamma irradiation was performed, which offered, according to Phillips, gamma fields up to 10^7 roentgens per hour.

The MTR first began operating in 1952—although, according to the excellent “Proving the Principle” (Susan M. Stacy/Idaho Operations Office of the Department of Energy, 2000), the plans were started for what became the MTR as early as 1944. The MTR, when placed in operation, quickly found itself with a list of experiments to perform and samples to irradiate. According to documentation provided by Erik Simpson, CWI media spokesman, the MTR performed over 15,000 irradiation experiments during its operational lifetime.

The MTR operated successfully as one of the most highly in – demand test reactors for many years. Time caught up to the MTR in 1970; according to “Proving the Principle,” the final experimental plutonium core (nicknamed “Phoenix”) was operated in the reactor through April 23, 1970, when the reactor was shut down. One final experiment in August 1970 saw the MTR go critical again for 48 hours when Aerojet, by then the MTR contractor, started it up for paid research into mercury contamination of wildlife. But that was it. The reactor never operated again.

The reactor was defueled, and parts of the facility were used for other purposes (some functions even going on next to the shutdown reactor itself without involving it) for some years until the DOE made the decision in 2005 to dispose of the facility. Erik Simpson has provided us with a copy of the 2007 Engineering Evaluation/Cost Analysis for the Materials Test Reactor End-State and Vessel Disposal; of the various site solutions described in this document, the one chosen and carried out is the one that called for removal of the above grade structure, the reactor vessel, and below-grade structure with the vessel being stabilized and stored onsite at a dedicated facility.

Erik provides us with two fascinating links that show much more than we saw in the opening video of the decommissioning of the MTR facility. In the first video link, we see a number of activities of the Idaho Cleanup Project; the MTR facility is seen in this video at the time frame 1:15 – 2:30. The second video link gives us a mostly time-lapse view of the demolition of the MTR reactor building (with the large internal shielding and beam tube/sample tube complex, as well as reactor vessel and tank extensions already gone), but slows to real-time to display the explosive demolition of the roof structure.

It goes without saying that in terms of the overall site, many reactor facilities have been remediated, or placed in some level of storage, or will be remediated. Dr. Stricker points out that the former NRTS site, now called the Idaho National Laboratory site, has housed 52 different reactors.

As related in “Proving the Principle,” there were serious last-minute attempts to revitalize the MTR with new projects and new money, but this wasn’t enough to prevent its shutdown; designation of the MTR as a “historical Signature Property as designated by DOE Headquarters Advisory Council on Historic Preservation” (as related in the disposal analysis) wasn’t enough even to keep the building. We’ve at least put a marker for the MTR and all those who worked on, or at, the facility on the ANS Nuclear Cafe blog with this post, and noted its passing.

(Photo at top courtesy Idaho National Laboratory, via Dr. Nicole Stricker. Video links courtesy Erik Simpson. MTR brochure photos, Will Davis collection.)

Additional resources

For more information, please visit Argonne National Laboratory’s Basic and Applied Science Research Reactors website—click HERE to open the the page dedicated to the MTR.

_______________________

→ 1 Comment

Posted in engineering, environmental remediation, Irradiation, National Laboratories, Nuclear pioneers, nuclear technology

Tagged Argonne National Laboratory, Basic and Applied Science Research Reactors, high neutron intensity reactors, Materials Testing Reactor, MTR, Oak Ridge National Laboratory, Phillips Petroleum, Will Davis

NRC Public Meeting on San Onofre: October 9 via Webcast, Twitter

Posted on October 9, 2012 by lscheele| Comments Off

Note: The NRC public meeting on San Onofre steam generator issues has now adjourned. The webcast will soon be available in archived form at http://video.nrc.gov/. The twitter feed featuring participation by groups on all sides of the issue can be viewed HERE (tweets will eventually expire).

WHEN:

Tuesday, October 9
6:00-9:30 P.M. Pacific Time
Click HERE for the U.S. Nuclear Regulatory Commission (NRC) news release with schedule information

VIRTUAL ACCESS

The meeting is now live on webcast at: http://video.nrc.gov.

A phone bridge will be available by calling: 1-888-989-4359 and entering pass code 1369507.

The webcast and phone bridge will be one-way only.

SOCIAL MEDIA ACCESS

ANS will live-tweet the hearing at @ans_org using hashtag #SanOnofre. Please note that the person(s) doing the live-tweeting will be watching via webcast.

FOR MORE INFORMATION

Click HERE for social media coverage by Will Davis of Atomic Power Review of the San Onofre Nuclear Generating Station steam generator issues, including a roundup of helpful links at the end of the entry.

Click HERE for a wealth of information from Southern California Edison regarding the San Onofre steam generator issues.

Comments Off

Posted in News, nuclear opponents, Nuclear Regulatory Commission, San Onofre Nuclear Generating Station

Tagged American Nuclear Society, atomic power review, Laura Scheele, Meredith Angwin, public hearing, public opinion, San Onofre, southern california edison, steam generators, Will Davis

ANS Annual Meeting Dresden Station technical tour

Posted on August 29, 2012 by pbowersox| 5 Comments

by Will Davis

When I was registering for the various events scheduled to take place during the ANS Annual Meeting this past June, I was quite excited to see that one of the three technical tours would be at Exelon’s Dresden Nuclear Station, not too far from downtown Chicago where the meeting was taking place. Luckily, I made the cut for attendance and was issued a ticket for the tour when I checked in at the meeting desk.

The transportation for the tour was a comfortable, air-conditioned motor coach—which was good, because Illinois was in the midst of a heat wave. The ride to Dresden was quicker than I’d expected, given the distance. Those of us who hadn’t been there before were looking out of the windows all the time to see the first hint of the tall stacks, or perhaps the spherical containment for the now shut down Dresden Unit 1.

Permit a digression at this point; the aforementioned structure, which is a steel sphere 180 feet in diameter, is one of those “nuclear relics” of some note that we have in this industry. Spherical containment didn’t last too long in commercial nuclear plant construction before giving way to far easier to build (and less expensive) cylindrical containment buildings. Seeing a photo or drawing of a spherical containment building immediately brings to mind the early days of nuclear energy; structures like this are the pyramids of our field. (Note: They are less permanent, though—a number of them have been completely dismantled and removed at other sites.) This would be my chance to check off No. 2 on my list of viewing such containment buildings; I say No. 2 because I formerly walked daily past the only larger one, at the Kesselring Site in New York. Not far below this in rank in terms of “nuclear archaeology,” if you will, is the fact that Dresden-1 was the nation’s first privately financed commercial nuclear plant. All of these reasons make Dresden Station a top priority for those of us with an interest in preserving a record of our nuclear history. In fact, the American Nuclear Society has designated Dresden Unit 1 as a Nuclear Historic Landmark.

Dresden Unit 1 under construction, April 1958

When the bus arrived at the station, I noticed something immediately that I had not noticed before in photos: The newer Dresden 2 and 3 units were built immediately adjacent to the Dresden-1 turbine building—and in fact the buildings abut and connect. Dresden-2 and -3 are later model GE boiling water reactor/3 reactors in Mark I containment buildings. The sight of the plant is thus a mixture of the old, or should I say original, and the more modern at once. Both units 2 and 3 were running at full power that day; the load on the grid from the heat wave was making the news.

After an orientation and welcome, along with issuance of dosimetry and a few questions, we were divided into groups of not more than five persons each; each group would have one or two escorts for security purposes who also doubled as our tour guides. My “group” as it were had only two members; our escort was Marisa Seloover, a young electrical engineer who acts as the plant’s systems engineer for compressor equipment. Marisa immediately showed her enthusiasm for her job, and was extremely informative and helpful at all times. In fact, everyone at the plant was extremely willing to tell us information and describe operations at the plant, as well as explain equipment.

The tour overall had to be cut a bit short, because the time was cramped and also because the temperature outside was about 102 °F that day. We toured a good portion of the operating plants, although since this was a BWR plant, close access to the turbine generators wasn’t allowed. We looked at control rod drive equipment, the access doors to the drywells, various pumps, and various labyrinthine spaces around the reactor buildings. We got a chance to stand on the refueling floor and look right down into the spent-fuel pool for Unit 3; yellow-clad workers were up on a ladder in the distance. The volume of the space was more impressive than I’d pictured it; the refueling floor level spans both reactors.

A fun moment of the tour occurred when we stopped next to one of the feed pump rooms. One of the escorts managed to yell to us over the din that the pump room held three electric feed pumps, each of roughly 7000 horsepower, and that much of the noise we’d heard outside the plant was actually the cooling air for these pump motors. He indicated that hearing was practically impossible if the access door were open; then he opened the door. He was correct. And yes, of course, we were wearing hearing protection, hard hats, and safety glasses issued by Exelon before leaving the training building outside the plants.

I myself had reserved a special enthusiasm for seeing Dresden-1, and we walked through the turbine hall of Unit 2 to an access door and immediately were in the turbine building of Unit 1. The turbine generator and associated equipment are long gone; the building is now used primarily for tool storage and maintenance work. As we walked along a level that would originally have been well above the turbine generator, I looked down and thought of the old photos I’d seen of when the plant was operating.

Dresden Nuclear Power Station dedication ceremony, October 12, 1960

Then, we went through a door and into the spherical containment itself. One of our escorts immediately yelled—and the echo, which he knew he would get, made all of us laugh. We were on a level above that of the steam generators, which he said were below the flooring (remember that Dresden-1 was a dual-cycle BWR with both direct steam to the turbine generator and four steam generators that fed steam to it as well.) The height of the concrete structure rising block-like in front of us, which formerly contained the steam and water piping and on top of which was the central steam separator drum, was quite impressive. We walked around to our left and could see the opening in the center of this structure that essentially amounted to the refueling space—below the steam drum, and between the steam risers. I quickly imagined that this would have been entirely an exclusion zone when the reactor was operating. The emptiness of the rest of the structure, with few signs of equipment, made the area feel less like a nuclear power plant and more like some sort of test mockup, which of course it was not. As we left the sphere, I recalled that the original design plan included not only the turbine generator but the control room as well inside the sphere; GE and Bechtel eventually changed their minds about that.

After we returned to the training center, we were given water to drink because of the heat (we’d skipped the tour of the dry cask storage areas because of that factor) and got the chance to talk to our guides for a few moments. Marisa noted that Unit 2 had been upgraded over the years, and was rated about 960 MWe—although on that day, because of ambient temperatures being so high, the plant was limited to about 950 MWe. Unit 3 was at its full rated power of about 912 MWe; it had not received all of the upgrades yet that Unit 2 had (main generator rewind, new low pressure turbines and turbine casings, new and much more modern recirculation pump drives) but would receive the last of them during the next refueling.

She explained how tight the limits were on water that the plant can discharge to the rivers, and how the plant uses a combination of river water, cooling lake, and added cooling towers to meet thermal discharge limit requirements. Her descriptions of practice drills and events for the plant were very helpful and informative. She, and others like her, struck me as the bright hope for the future in the nuclear energy industry.

After this, it was time to complete the day’s events with brief tours of the control room simulator, and a presentation on the plant, its history, and its operation given by Work Management Director Joe Sipek. The control room tour was extremely informative; the personnel there answered all of our questions fully and clearly. The tour was very thorough, even including the alley behind the main panels and descriptions of how the simulator functions, as well as how the staff rotates through.

After Mr. Sipek’s presentation and some souvenir Dresden Station pens were handed out, it was time to get on the bus and leave for Chicago and our comfortable hotel. This was the final event for the whole annual meeting for me; I’d be flying home the next day. On the way in, I’d looked at the site and wondered about very many things—mostly about what I’d find inside, what I’d see. On the way out, I thought mainly about the people I’d met and how well they’d impressed me. Of course it goes without saying that any plant is primarily its people, and Exelon left this writer with an impression of a well-motivated, well-engaged, dedicated workforce.

All in all, as the bus made the journey down the access road out of the plant, I knew that my lasting impression would be not of things like structures that would someday be dismantled, but instead would be of the people who work and pass their knowledge on to others. It’s no overstatement to say that the time invested in the trip paid me great dividends.

I’d like to thank the following people from Exelon who helped make the trip possible: Natalie Zaczek, Paul Bembnister, Scott Ackerman, Joe Sipek, Marisa Seloover, Dan Murphy, Tom Mohr, Kyle Cook, Samantha Cosenza, Nick Oudin, and Marie Frese. Also, thanks to Robert Osgood of Exelon who handled pre-trip security and communications. Any omissions from this list are my error and I offer my apologies for any such.

__________________________

Davis

Will Davis is the author of the nuclear energy blog “Atomic Power Review,” and is a member of the American Nuclear Society. A former US Navy reactor operator, Davis finds his calling to be presenting the public with information about nuclear energy technology and its history.

→ 5 Comments

Posted in American Nuclear Society, Meetings

Tagged American Nuclear Society, Dresden station, electricity, nuclear power, technical tour, Will Davis

Doel-3 in Belgium reports possible pressure vessel flaw

Posted on August 15, 2012 by pbowersox| 8 Comments

Findings could be significant for other reactors

by Will Davis

Ultrasonic testing of a reactor vessel at the Doel nuclear power station in Belgium has revealed what may be tiny cracks, causing the owner-operator group and Belgian regulatory authority—the Federal Agency for Nuclear Control (FANC)—to commence further testing. Belgian authorities said that they would notify other plants around the world using reactor vessels manufactured by Rotterdam Drydock Company, the company that made the reactor vessel used in Doel-3.

Doel-3 was shut down on June 2 for a 10-year in-service inspection requiring examination of thousands of plant components as per local procedure. During the inspections, testing of the reactor vessel was conducted using a new type of ultrasonic detection equipment. This testing—which is focused on high-stress areas near the pressure vessel belt-line or barrel section welds—has turned up what appear to be a number of cracks, according to public statements made by FANC.

FANC’s website indicates the discovery of a single larger crack, running parallel with the (curved) surface of the pressure vessel, in the lowest ring section of the body (or barrel area of the vessel). This vessel is of forged ring design; see illustration below.

An illustration from FANC showing the original forged ring sections of the reactor vessel separated for clarity

According to FANC, this single crack is between six-tenths and eight-tenths of an inch long. The crack is not in a direction “normally subject to tension,” according to FANC’s site, and is thus theoretically of no risk. Reports have been made fairly widely that the cracking discovered in the vessel is due to radiation embrittlement of the steel, but FANC has reported that this crack is thought to be an original manufacturing defect.

According to FANC, this defect is in the body of the forged ring, and not near one of the circumferential welds. Previously, only the weld areas were inspected under such tests in Belgium, but this test sequence for the first time inspected the body areas of the vessel away from welds.

FANC and the plant’s operator, Electrabel, have commenced a second round of tests with an older, widely used type of ultrasonic testing equipment in hopes of determining whether the anomalies detected in the Doel-3 pressure vessel are real, or are byproducts of some phenomenon associated with the new testing equipment/method.

FANC has reported to the International Atomic Energy Agency that these events constitute a significance of a Level 1 occurrence on the International Nuclear and Radiological Event Scale—Level 1 is an “anomaly,” with no hazard to the public. The scale runs from Level 1 to Level 7, with Level 7 being the most serious threat to the public.

In the meantime, press attention has focused on the fact that the pressure vessel in this plant may not be unique. One other plant in the Netherlands, two in Spain, one in Switzerland, and one in Sweden also have pressure vessels manufactured by Rotterdam Drydock; as many as 21 total plants may be implicated.

There are nine operating plants in the United States that have pressure vessels manufactured by this same firm. Catawba-1, McGuire-2, North Anna -1 and -2, Sequoyah -1 and -2 and Watts Bar-1 have ring forging pressure vessels like that shown above, fabricated of SA-508 Class 2 steel. In addition, Surry-1 and -2 have composite fabrication pressure vessels partly made by Babcock & Wilcox and by Rotterdam Drydock. AEC (now NRC) documents, as well as statements by FANC indicate ten vessels were ordered from this firm, and it is believed that this tenth plant is the unfinished Watts Bar-2.

Rotterdam Drydock entered the field for manufacture of pressure vessels for U.S. nuclear plants in 1969 when a backlog of orders for pressure vessels (among the longest lead time items of the entire plant, and usually second to the turbine-generator as a whole) began to build up. American manufacturers Babcock & Wilcox and Combustion Engineering, the two suppliers of pressurized water reactor pressure vessels for U.S. vendors at the time, were then busily expanding their facilities, but a delay in the startup of Babcock’s then-new facility on the Ohio River at Mount Vernon, Indiana, forced the ordering of two vessels in 1969 from Rotterdam Drydock on Westinghouse vendor contracts as substitutions. Eventually, as many as 10 vessels were ordered for U.S. plants from Rotterdam—all for use in Westinghouse PWR plants.

Rotterdam Drydock went out of business in the mid-1980s.

What we don’t know

Many facts have yet to be reported that will bear mightily on the decisions to be made about Doel-3′s safety. It’s important to consider the alloy with which this vessel is made (1), the total embrittlement of that material over the life of the vessel so far due to neutron exposure, the number of pressurized thermal shock cycles the plant may have endured, the number of times the plant has overcooled or operated outside of its normal parameter bands (if any), and more.

One limiting factor for Doel-3, which is a 3-loop PWR supplied by a consortium known as FRAMACECO (consisting of Framatome, ACEC and Cockerill) and which first achieved criticality in 1982, will be the exact composition of its vessel and how that relates to embrittlement, and reduction in strength by any flaws. Pressure vessel steel specifications for reactors were changing during the time this plant was built, and it is not known if the vessel for this plant conforms to later criteria specifying especially low content of copper and phosphorus. (Standards changed under a supplement to the ASME code prior to the time that the vessel for Doel-3 would have been made, and reduced the trace elements that most seriously contributed to higher rates of embrittlement.)

What we do know: Flaw analysis and embrittlement

All reactor pressure vessels experience embrittlement—that is to say, a change in characteristics over time that makes the material less ductile due to neutron bombardment—which largely determines a plant’s life span, since the reactor vessel itself is considered non-replaceable.

In terms of what cracking the new crack at Doel-3 might imply, and whether or not PWR vessels have sufficient margin to survive such defects if unfound, we might look to a study performed in France over a 10-year time span in the 1970s and 1980s. This study is described in ASTM STP 819, “Radiation Embrittlement and Surveillance of Nuclear Reactor Pressure Vessels: An International Study.” We find the following summation of the French program on page 31 in description of graphed test results:

“The results are compared with the minimum embrittlement as predicted by the U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99. It appears that in practically all cases the measured embrittlement is less than predicted. In Fig. 1 [not included here] all data points above the curve correspond to steels with phosphorus and copper content greater than 0.08 and 0.8% respectively.

The major conclusion of this study is that it is possible to produce steels and welds with relatively low sensitivities to radiation embrittlement if copper and phosphorus contents are maintained at a low level. This is easily achieved for copper content, but seems more difficult for phosphorus.”

The clear implication of this French study (cited from among the many other studies available due to the fact that it was, entirely, conducted by the French nuclear industry that participated in the construction of Doel-3) is that there is often a significant margin between predicted embrittlement and actual embrittlement, both of which are less than the minimum allowed. This will be the “playing room,” so to speak, that the FANC and Electrabel will have in determining the fate of the plant vis a vis the reduction in local material strength from the defects.

Many other studies have been performed worldwide on the behavior of irradiated reactor vessel steels, not only in terms of their embrittlement and reduction in strength due to neutron exposure but also as to the nature of crack propagation and the ability of the material to arrest crack growth. Should the defects in Doel be found to actually exist, it will be up to FANC to determine whether or not to consult such studies and bank on the safety of the vessel (which has already survived three decades with this flaw, if indeed it is an original defect), to perform more tests, or perhaps even to shut down the plant. Replacement of a reactor vessel has never been performed; these are always disposed of when a plant is decommissioned and dismantled and are considered one of the few “lifetime” components of a nuclear power plant.

Basic analysis for reactor safety includes—as a prerequisite of the determination of the vessel’s ability to avoid fracture due to embrittlement and due to temperature changes—a given or assumed initial flaw that is used for calculational analysis. This is not to say that this flaw exists in a (or any) vessel, or that it is allowed to exist; it is simply a basic assumption that provides an extra margin of safety and serves as a launching point for further calculation.

A short while back, this calculational assumption took on a life of its own when activists got hold of a report about the Genkai plant in Japan and assumed that the vessel had a known flaw built in; it did not. The people reporting on this simply did not understand the “pre-existing flaw” assumption made to ensure safety (2).

Possible significance for U.S. plants

How this applies to the ten U.S. plants that have vessels made by Rotterdam Drydock is as follows: What must first be determined is whether or not the defects in the Belgian plant exist or are anomalous. If they do exist, and are determined to be manufacturing defects present since the vessels were made, then it might be logically assumed that such defects might be present in the U.S. built vessels—assuming they were built to exactly the same manufacturing requirements, post-manufacturing quality assurance checks, and handling specifications, and are made of exactly the same alloy with the same trace element content. Further, it would have to be determined that similar inspection for identical flaws has not been carried out in any of the plants. Were all of that to be the case, an inspection of the U.S. vessels might be warranted.

As of this writing, only North Anna-1 and -2, and Surry-1 and -2 in the United States have been publicly identified previously (in Platts, on Friday) as having pressure vessels manufactured by Rotterdam Drydock. Dominion Virginia Power/Dominion North Carolina Power told Platts on August 10 that none of these plants has shown any indication of such cracking in inspections.

In Belgium, another inspection is planned: Tihange-2 also has a pressure vessel manufactured by Rotterdam Drydock that will be inspected in several months.

As should now be clear, the whole situation regarding even the Doel-3 vessel is at this time not yet fully known, and thus any implications for the U.S. plants cannot be accurately made as of yet.

Illustration from FANC of the actual reactor vessel in the Doel 3 nuclear power plant

Notes:

(1) For a discussion of the pressure vessel alloys and trace contents, see this link at Atomic Power Review

(2) For background on the Genkai assumption, see this link at Atomic Power Review

Reporting from the following sources has been used in preparing this report: World Nuclear Association / World Nuclear News, Marketwatch, Platts, The Turkish Weekly, Energy Business Review. Documents consulted from US Department of Energy, US AEC (now US NRC), and American Society for Testing and Materials.

______________________________

Davis

→ 8 Comments

Posted in News, nuclear technology, reactor pressure vessels

Tagged Belgium, Doel-3, embrittlement, nuclear energy, pressure vessels, reactor vessel, Rotterdam Drydock, Will Davis

Tag Archives: Will Davis

Kewaunee: What does the future hold?

The Hook-Ons

Inherent and engineered safety: Did Weinberg predict today’s reactors a quarter century ago?

San Onofre debate now more public – and more technical

Preparing to restart: Tsunami safety measures at Japanese nuclear power stations

The Atlantic Generating Station

Clinch River Site will once again lead nuclear development

ANS Winter Meeting 2012: Nuclear Technology Expo

Spent Fuel Pool at Oyster Creek

The MTR—Gone now, but not forgotten

NRC Public Meeting on San Onofre: October 9 via Webcast, Twitter

ANS Annual Meeting Dresden Station technical tour

Doel-3 in Belgium reports possible pressure vessel flaw

Search

Blogs We Read

15 most recent blog posts

Blog Archives

Categories

Admin

Sign Up for Free Daily Email Summary

ANS Video News Clips

ANS Nuclear Headlines

Social Media

Recent Comments