Question: b. How much will it cost a reactor vendor to develop and obtain NRC certification for a new reactor design?

Answer: The cost of developing and obtaining NRC certification for a new reactor design depends greatly on the status of the technology and the design at the beginning of the program. The typical cost for NRC certification of an advanced light water (ALWR) design, starting from a conceptual design, is $200 million. These designs, as I noted earlier, are based on thirty years of technology development and operating experience. For designs such as the ALMR and MHTGR, for which the technology and operating base is less developed and therefore require prototype demonstrations, the cost of the total program will exceed $ 1 billion.

GE's ABWR Design Certification Program is budgeted at only $25.8 million. This amount is cost shared between GE and DOE. This limited budget was possible because the design effort, including required test and development, was performed by GE and our associates in support of the plants we are about to build in Japan. Only the U.S. licensing effort had to be supported. I believe a similar situation exists for the ABB-CE System 80+ design.

Only the conceptual design of GE's SBWR had been completed prior to commencement of the Design Certification Program. The total program cost through certification is budgeted at $140 million. DOE is providing $48.5 million in funding with the remainder provided by the private sector. Additionally, we estimate we have avoided approximately $75 million of development cost by applying much of the advanced technology we developed for ABWR. I understand the W AP-600 has a similar cost profile.

Following certification of the ALWR designs and prior to commencement of construction, another large effort will be required to meet the ALWR utility requirement for design completeness. The nuclear industry as part of its standardization policy has stated that 90% of the detailed design should be completed prior to placement of first concrete at a new plant. In order to complete this design effort, known as First-Of-AKind (FOAK) Engineering, we estimate that typically an additional $200 million is required for each design, although this number will also vary based on the preprogram status of the design. Discussions are underway in the industry, including DOE and EPRI, to define programs and funding methods to accomplish this effort.

Question: 7 a. What is the problem at NRC?

Answer: The ABWR technical review got off to a good start and initial progress was excellent. In spite of its good start, however, the ABWR review has recently been affected by process difficulties within the NRC. The first setback came in December 1989 when the Commission reexamined its role in the ALWR programs and provided the NRC staff with a new review procedure to be used on these programs. The new procedure is intended to involve the Commission early in the resolution of ALWR policy issues. Unfortunately, the procedure for early Commission involvement came very late in the ABWR certification effort. The impact of this new procedure was a 10 to 15 month delay.

Most recently the Commission has reexamined the level of design detail required for certification. This issue was addressed early in the licensing review bases discussions in 1986 and 1987 by GE, the staff, the ACRS and the Commission and we believed we had agreement on the detail required. Unfortunately, this level of detail issue reemerged much later at a time when the ABWR SSAR had been virtually completed and the NRC staff technical review was nearing completion. While this issue has again apparently been resolved by the Commission policy discussed above, significant delay was experienced by this uncertainty with respect to process.

The NRC is regulating an industry with very long product cycles compared to other industries. The ABWR was conceived in 1978 and will enter operation in Japan in 1996. The licensing process in the U.S. began in 1986 and we trust will result in NRC certification in 1992. Construction of the initial unit will take an additional six years. Operation is planned for an additional sixty years. During this entire period large investments are being made which critically depend upon the stability, predictability and finality of NRC policies and positions. Conversely, the NRC Commission, as well as other regulatory bodies such as Public Utility Commissions, has a relatively short cycle of change. Each new Commissioner brings his or her own perspective to the position and the institutional memory has tended to be very short. Until a reformed NRC licensing process is codified as proposed, in part, in title XIII and a Congressional mandate for a stable licensing environment is given, I fear the industry will face major, perhaps insurmountable, uncertainties and obstacles to the detriment of the nation. Question: 7 b. What can Congress do to get the NRC back on schedule?

Answer: The outlook for successful demonstration of the new licensing process is at a critical point. The task at hand is to restore meaningful project schedules and permit expeditious processing and issuance of the final design approval (FDA) followed by design certification through rulemaking for the ABWR. To do so would permit the start of the institutional reactor demonstration project (described above) needed to restore the nuclear option. GE, in a joint program with DOE, remains committed to do its part to successfully demonstrate the workability of the new licensing process. Congress' continued support of funding authorizations is essential.

Congressional oversight is also clearly needed to monitor the process. An active Congressional interest in the causes for institutional delays and in achieving timely remedial action is an important aspect of productive oversight. Constructive interest in NRC Certification delays, such as that expressed in last year's report1 of the Senate Subcommittee on Energy and Water Development, is very helpful in maintaining a clear, goal-oriented course.

1) Report of the Senate Subcommittee on Energy and Water Development, Committee on Appropriations (S.Rep. No. 101-378, 101st Congress, 2d Sess, July 14, 1990 at 199).

Questions

Senator Malcolm Wallop (R-WY)

Nuclear Advanced Reactor and Licensing

provisions of S.341, the Energy Security Act of 1991

March 5, 1991

QUESTIONS FOR BERTRAM WOLFE

Question: You point out that today's nuclear technology is very different from that of 30 years ago when commercial nuclear plants began operating. Could you please state those advances you feel are most significant in relationship to efficiency, economy and enhanced safety features?

Answer: Over the past thirty years, light water reactor technology has matured and been proven. We have encountered difficulties as we gained experience along the way; however, we have responded positively to the challenges and now have a demonstrated capability to design, build and operate even better plants in a safe, economical, and environmentally sound manner. While other advanced technologies are in development, the advanced light water reactor is ready now for large scale deployment.

In viewing the evolution of GE's boiling water reactor (BWR) over the past three decades three trends become evident which I'll discuss in more detail below. These trends are first, the growth in plant electrical output; second, the simplification of the basic reactor and containment designs; and finally, the increase in safety.

General Electric selected the boiling water reactor (BWR) as the most promising commercial nuclear power concept because of its inherent advantages in control and simplicity, and established an atomic power equipment business in 1955 to offer it commercially.

The beginning of the GE BWR product line was the Vallecitos BWR in 1957, which was located in California. This reactor powered a 5-MWe generator and provided power to the Pacific Gas and Electric Company grid through 1963. A major extrapolation from that first test facility is the Dresden 1 plant, located in Morris, Illinois. Order for the 180 MWe plant was placed in 1955, with commercial power production achieved in August of 1960.

The value of the BWR concept has been proven by the designs GE evolved over the subsequent decades. The Oyster Creek plant, located in Toms River, New Jersey, was the first large direct cycle light water reactor. This plant which entered operation in 1969 had an output well over 500 MWe, truly a commercial size, economical power plant. The BWR continued a disciplined, evolutionary, building-block approach, supported at each step by research and testing towards larger size power plants to take full advantage of the economy of scale. Power plants of 1100-1300 MWe were introduced in the mid-seventies. The ABWR, a 1300 MWe power plant, is the most advanced plant design in that series. In response to the need for a mid-size power plant, GE is also developing the 600 MWe SBWR.

The second trend noted is the evolution of the BWR toward simplification in two major areas - the reactor system and the containment design. This evolution resulted from design enhancements and experience gained from operating reactors, abnormal occurrences and test programs. Both the ABWR and SBWR are the results of this progressive simplification.

The earliest BWR were very similar to pressurized water reactors (PWRs) and had primary and secondary loops. However, the first large BWR, Dresden 1, a dual cycle plant, provided steam flow directly using an elevated steam drum and also via a secondary steam generator to the turbine. The first major simplification came with the introduction of internal steam separators and dryers to replace the external steam drum. This also lead to the elimination of the steam generators and to the first direct cycle BWRs in the late 60s. These designs had five external recirculation loops, which were soon replaced by two loops with the introduction of internal jet pumps. This process of continuing BWR simplification culminated in the next logical step for large plants: total elimination of the external loops and jet pumps and use of internal pumps on the ABWR. For smaller plants, like the SBWR, total elimination of all recirculation pumps was made possible by the use of natural circulation.

The containment design evolution has had a similar extensive history with the design always tending towards simplified construction, enhanced safety, and plant protection. Early BWRs were housed in large volume dry containments of a spherical design similar to many PWRS today. However, the pressure suppression containment was developed and became a standard feature of BWRs, because of its many advantages: high heat capacity; lower design pressure; depressurization capability; fission product scrubbing and retention; and provision of a large source of readily available makeup water.

BWR pressure suppression containment design evolved from the early Mk I design with its characteristic torus/lightbulb arrangement, the Mk II where the larger drywell provided more room for the steam and emergency core cooling systems (ECCS) piping, to the Mk III design with its simple cylindrical shape for ease of construction and horizontal vents for reduced dynamic loads. The containment also evolved from the all-steel, free-standing (in many cases) structures to the reinforced concrete containment design.

The next logical step in the containment evolution was to take the best features of the previous designs and develop an optimized design to meet the performance, construction and cost goals of our ALWRs. The cylindrical shaped reinforced concrete containment was chosen for both the ABWR and SBWR. Many of the standard features of the BWR containments enumerated above are, in fact, being adopted in other advanced passive designs.

The final trend I'll discuss is related to increases in safety. These improvements were driven by technology advances, lessons learned and regulation. The first nuclear reactor at Stag Field at the University of Chicago you'll recall was contained only by a grandstand and the safety control rods depended on an axe. We've understandably come a long way.

The first major decision in safety assurance was to require all commercial nuclear power plants to have containments which could protect the public in case of an accident. Today, it seems clearly evident that such containments are prudent However, others, specifically the Soviets, took a different approach. Their reactor designed without containments and with other flaws, led to disastrous results. While light water reactor technology was in its infancy when the decision for containment was made and many risks we have recognized and dealt with over the years were unimagined, the industry made a conservative and in retrospect a wise decision. This single feature lowered the risk to the public by a significant degree.

Emergency core cooling systems (ECCS) were the next major safety feature conceived. These systems, with their emergency backup diesel sources of AC power, greatly reduced the probability of core damage due to loss of coolant. These systems have been greatly improved over the years, as I will discuss later. The ABWR design incorporates three redundant and independent divisions of ECCS and containment heat removal. The SBWR uses gravity and other natural forces to provide these functions.

The final major advancement which I'll discuss is the improvement in design codes and practices utilized to ensure the containment and other vital structures will withstand seismic events without jeopardizing the power plant or the public. Hundreds of millions of dollars have been spent in characterizing through testing, modeling on high speed computers and incorporating in the design the forces which will act upon the plant during an earthquake.

These three features are fundamental to nuclear safety and in combination reduce the risk to the public by five or six orders of magnitude. However, even with these features, our job was not complete. We have continued in numerous ways to improve safety. I'll mention a few of the more significant improvements, but will not attempt to give a complete accounting. Items such as reduced occupational exposure, hydrogen control, post-accident equipment qualification, electrical separation, fire protection, post accident monitoring, improved operator interface, probabilistic risk assessments and failure modes/effects analysis, and provisions for severe accidents are worthy of note.

The current fleet of operating plants has incorporated these improvements and represent an extremely safe, reliable source of electrical power. Their performance has improved significantly in recent years. However, once again the industry was not content. In establishing the requirements for the advanced light water reactors and in completing their design, significant additional improvements have been made. I have included a table of key performance characteristics which illustrates my point. Many of the characteristics listed for GE Operating BWRs are significant improvements over the record of just a few years ago and represent the results of an on-going effort by the industry to improve the performance of the current operating fleet. We are confident, however, that our advanced BWR plants will greatly exceed even the improved performance levels of the operating plants as the key performance characteristics table shows.