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Boiling Water Reactors

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Published in: Mechanical
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Plant engineering 

Sandeep K / Kolkata

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Qualification: M.Tech. (Production Engineering)

Teaches: Chemistry, English, Hindi, Physics, Drawing, Mechanical

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  1. BOILING WATER REACTOR Boiling Water Reactor: The boiling water reactor (BWR) is a type of light water nuclear reactor used for the generation of electrical power. It is the second most common type of electricity-generating nuclear reactor after the pressurized water reactor (PW R), also a type of light water nuclear reactor. The BWR was developed by the Idaho National Laboratory and General Electric in the mid-1950s. The main present manufacturer is GE Hitachi Nuclear Energy, which specializes in the design and construction of this type of reactor. Early concepts The B WR concept was developed slightly later than the P WR concept. Development of the BWR started in the early 1950s, and was a collaboration between GE and several US national laboratories. Research into nuclear power in the US was led by the 3 military services. The Navy, seeing the possibility of turning submarines into full-time underwater vehicles, and ships that could steam around the world without refueling, sent their man in engineering, Captain Hyman Rickover to run their nuclear power program. Rickover decided on the P WR route for the Navy, as the early researchers in the field of nuclear power feared that the direct production of steam within a reactor would cause instability, while they knew that the use of pressurized water would definitively work as a means of heat transfer. This concern led to the US's first research effort in nuclear power being devoted to the PWR, which was highly suited for naval vessels (submarines, especially), as space was at a premium, and PWRs could be made compact and high-power enough to fit in such, in any event. But other researchers wanted to investigate whether the supposed instability caused by boiling water in a reactor core would really cause instability. In particular, Samuel Untermyer Il, a researcher at Idaho National Laboratory (INL), proposed and oversaw a series of experiments: the BORAX experiments—to see if a boiling water reactor would be feasible for use in energy production. He found that it was, after subjecting his reactors to quite strenuous tests, proving the safety principles of the BWR. ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 1
  2. BOILING WATER REACTOR Following this series of tests, GE got involved and collaborated with INL to bring this technology to market. Larger-scale tests were conducted through the late 1950s/early/mid- 1960s that only partially used directly-generated (primary) nuclear boiler system steam to feed the turbine and incorporated heat exchangers for the generation of secondary steam to drive separate parts of the turbines. The literature does not indicate why this was the case, but it was eliminated on production models of the BWR. First series of production BWRs (BWR/1—BWR/6) The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR: the torus (used to quench steam in the event of a transient requiring the quenching of steam), as well as the drywell, the elimination of the heat exchanger, the steam dryer, the distinctive general layout of the reactor building, and the standardization of reactor control and safety systems. The first, General Electric, series of production BWRs evolved through 6 iterative design phases, each termed BWR/I through BWR/6. (BWR/4s, BWR/5s, and BWR/6s are the most common types in service today.) The vast majority of BWRs in service throughout the world belong to one of these design phases. 1st generation BWR: BWR/I with Mark 1 containment. 2nd generation BWRs: BWR/2, BWR/3 and some BWR/4 with Mark 1 containment. Other BWR/4, and BWR/5 with Mark-Il containment. 3rd generation BWRs: BWR/6 with Mark-Ill containment. Browns Ferry Unit I drywell and wetwell under construction, a BWR/4 using the Mark I containment Containment variants were constructed using either concrete or steel for the Primary Containment, Drywell and Wetwell in various combinations. ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 2
  3. BOILING WATER REACTOR The advanced boiling water reactor (ABWR) A newer design of BWR is known as the Advanced Boiling Water Reactor (ABWR). The ABWR was developed in the late 1980s and early 1990s, and has been further improved to the present day. The ABWR incorporates advanced technologies in the design, including computer control, plant automation, control rod removal, motion, and insertion, in-core pumping, and nuclear safety to deliver improvements over the original series of production BWRs, with a high power output (1350 M We per reactor), and a significantly lowered probability of core damage. Most significantly, the ABWR was a completely standardized design, that could be made for series production. The ABWR was approved by the U.S. Nuclear Regulatory Commission for production as a standardized design in the early 1990s. Subsequently, numerous ABWRs were built in Japan. One development spurred by the success of the ABWR in Japan is that GE's nuclear energy division merged with Hitachi Corporation's nuclear energy division, forming GE Hitachi, who is now the major worldwide developer of the BWR design. The simplified boiling water reactor (SBWR) General Electric (GE) also developed a different concept for a new boiling water reactor (BWR) at the same time as the ABWR, known as the simplified boiling water reactor (SBWR). This smaller (600 megawatt electrical (MWe) per reactor) was notable for its incorporation—for the first time ever in a light water reactor of "passive safety" design principles. The concept of passive safety means that the reactor, rather than requiring the intervention of active systems, such as emergency injection pumps, to keep the reactor within safety margins, was instead designed to return to a safe state solely through operation of natural forces if a safety-related contingency developed. For example, if the reactor got too hot, it would trigger a system that would release soluble neutron absorbers (generally a solution of borated materials, or a solution of borax), or ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 3
  4. BOILING WATER REACTOR materials that greatly hamper a chain reaction by absorbing neutrons, into the reactor core. The tank containing the soluble neutron absorbers would be located above the reactor, and the absorption solution, once the system was triggered, would flow into the core through force of gravity, and bring the reaction to a near-complete stop. Another example was the Isolation Condenser system, which relied on the principle of hot water/steam rising to bring hot coolant into large heat exchangers located above the reactor in very deep tanks of water, thus accomplishing residual heat removal. Yet another example was the omission of recirculation pumps within the core; these pumps were used in other BWR designs to keep cooling water moving; they were expensive, hard to reach to repair, and could occasionally fail; so as to improve reliability, the ABWR incorporated no less than 10 of these recirculation pumps, so that even if several failed, a sufficient number would remain serviceable so that an unscheduled shutdown would not be necessary, and the pumps could be repaired during the next refueling outage. Instead, the designers of the Simplified Boiling Water Reactor used thermal analysis to design the reactor core such that natural circulation (cold water falls, hot water rises) would bring water to the center of the core to be boiled. The ultimate result of the passive safety features of the SBWR would be a reactor that would not require human intervention in the event of a major safety contingency for at least 48 hours following the safety contingency; thence, it would only require periodic refilling of cooling water tanks located completely outside of the reactor, isolated from the cooling system, and designed to remove reactor waste heat through evaporation. The Simplified Boiling Water Reactor was submitted to the United States Nuclear Regulatory Commission, however, it was withdrawn prior to approval; still, the concept remained intriguing to General Electric's designers, and served as the basis of future developments. The economic simplified boiling water reactor (ESBWR) During a period beginning in the late 1990s, GE engineers proposed to combine the features of the advanced boiling water reactor design with the distinctive safety features of the simplified boiling water reactor design, along with scaling up the resulting design to a ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 4
  5. BOILING WATER REACTOR larger size of 1,600 MWe (4,500 MWth). This Economic Simplified Boiling Water Reactor design has been submitted to the U.S. Nuclear Regulatory Commission for approval, and the subsequent Final Design Review is near completion. Reportedly, this design has been advertised as having a core damage probability of only core damage events per reactor-year. (That is, there would need to be 3 million ESBWRs operating before one would expect a single core-damaging event during their 100-year lifetimes. Earlier designs of the BWR (the BWR/4) had core damage probabilities as high as Ix 10- core-damage events per reactor-year.) This extraordinarily low CDP for the ESBWR far exceeds the other large LWRs on the market. Advantages and disadvantages Advantages The reactor vessel and associated components operate at a substantially lower pressure (about 75 times atmospheric pressure) compared to a P WR (about 158 times atmospheric pressure). Pressure vessel is subject to significantly less irradiation compared to a PW R, and so does not become as brittle with age. Operates at a lower nuclear fuel temperature. Fewer components due to no steam generators and no pressurizer vessel. (Older BWRs have external recirculation loops, but even this piping is eliminated in modern BWRs, such as the ABWR.) Lower risk (probability) of a rupture causing loss of coolant compared to a PW R, and lower risk of core damage should such a rupture occur. This is due to fewer pipes, fewer large diameter pipes, fewer welds and no steam generator tubes. NRC assessments of limiting fault potentials indicate if such a fault occurred, the average BWR would be less likely to sustain core damage than the average P WR due to the robustness and redundancy of the Emergency Core Cooling System (ECCS). ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 5
  6. BOILING WATER REACTOR Unlike PWRs, BWRs have at least a few steam-turbine driven ECCS systems that can be directly operated by steam produced after a reactor shutdown, and require no electrical power. This results in less dependence on emergency diesel generators of which there are four in any event. Measuring the water level in the pressure vessel is the same for both normal and emergency operations, which results in easy and intuitive assessment of emergency conditions. Can operate at lower core power density levels using natural circulation without forced flow. A B WR may be designed to operate using only natural circulation so that recirculation pumps are eliminated entirely. (The new ESBWR design uses natural circulation.) BWRs do not use boric acid to control fission burn-up, leading to less possibility of corrosion within the reactor vessel and piping. (Corrosion from boric acid must be carefully monitored in PWRs; it has been demonstrated that reactor vessel head corrosion can occur if the reactor vessel head is not properly maintained. See Davis-Besse. Since BWRs do not utilize boric acid, these contingencies are eliminated.) BWRs generally have N-2 redundancy on their major safety-related systems, which normally consist of four "trains" of components. This generally means that up to two of the four components of a safety system can fail and the system will still perform if called upon. Due to their single major vendor (GE/Hitachi), the current fleet of BWRs have predictable, uniform designs that, while not completely standardized, generally are very similar to one another. The ABWR/ESBWR designs are completely standardized. Lack of standardization remains a problem with PWRs, as, at least in the United States, there are three design families represented among the current P WR fleet (Combustion Engineering, Westinghouse, and Babcock & Wilcox), within these families, there are quite divergent designs. ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 6
  7. BOILING WATER REACTOR Additional families of PWRs are being introduced. For example, Mitsubishi's APWR, Areva's US-EPR, and Westinghouse's APIOOO/AP600 will add diversity and complexity to an already diverse crowd, and possibly cause customers seeking stability and predictability to seek other designs, such as the BWR. BWRs are overrepresented in imports, if the importing nation doesn't have a nuclear navy (PWRs are favored by nuclear naval states due to their compact, high- power design used on nuclear-powered vessels; since naval reactors are generally not exported, they cause national skill to be developed in P WR design, construction, and operation), or special national aspirations (special national aspirations lead to a marked preference for the CANDU reactor type due to special features of that type). This may be due to the fact that BWRs are ideally suited for peaceful uses like power generation, process/industrial/district heating, and desalinization, due to low cost, simplicity, and safety focus, which come at the expense of larger size and slightly lower thermal efficiency. Sweden is standardized mainly on BWRs. Mexico's only two reactors are BWRs. Japan experimented with both PWRs and BWRs, but most builds as of late have been of BWRs, specifically ABWRs. In the CEGB open competition in the early 1960s for a standard design for UK 2nd-generation power reactors, the PW R didn't even make it to the final round, which was a showdown between the BWR (preferred for its easily understood design as well as for being predictable and "boring") and the AGCR, a uniquely British design; the indigenous design won, possibly on technical merits, possibly due to the proximity of a general election. Disadvantages • Much larger pressure vessel than for a P WR of similar power, with correspondingly higher cost. (However, the overall cost is reduced because a modern Complex calculations for managing consumption of nuclear fuel during operation due to "two phase (water and steam) fluid flow" in the upper part of the core. This requires ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 7
  8. BOILING WATER REACTOR more instrumentation in the reactor core. The innovation of computers, however, makes this less of an issue. BWR has no main steam generators and associated piping.) Contamination of the turbine by short-lived activation products. This means that shielding and access control around the steam turbine are required during normal operations due to the radiation levels arising from the steam entering directly from the reactor core. This is a moderately minor concern, as most of the radiation flux is due to Nitrogen-16, which has a half-life measured in seconds, allowing the turbine chamber to be entered into within minutes of shutdown. Though the present fleet of BWRs are said to be less likely to suffer core damage from the "l in 100,000 reactor-year" limiting fault than the present fleet of PWRs are (due to increased ECCS robustness and redundancy) there have been concerns raised about the pressure containment ability of the as-built, unmodified Mark I containment — that such may be insufficient to contain pressures generated by a limiting fault combined with complete ECCS failure that results in extremely severe core damage. In this double failure scenario, assumed to be extremely unlikely prior to the Fukushima I nuclear accidents, an unmodified Mark I containment can allow some degree of radioactive release to occur. This is supposed to be mitigated by the modification of the Mark I containment; namely, the addition of an outgas stack system that, if containment pressure exceeds critical setpoints, is supposed to allow the orderly discharge of pressurizing gases after the gases pass through activated carbon filters designed to trap radionuclides. A BWR requires active cooling for a period of several hours to several days following shutdown, depending on its power history. Full insertion of BWRs control rods safely shuts down the primary nuclear reaction. However, radioactive decay of the fission products in the fuel will continue to actively generate decay heat at a gradually decreasing rate, requiring pumping of cooling water for an initial period to prevent overheating of the fuel. If active cooling fails during this post- shutdown period, the reactor can still overheat to a temperature high enough that zirconium in the fuel cladding will react with water and steam, producing hydrogen ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 8
  9. BOILING WATER REACTOR gas. In this event there is a high danger of hydrogen explosions, threatening structural damage to the reactor and/or associated safety systems and/or the exposure of highly radioactive spent fuel rods that may be stored in the reactor building (approx 15 tons of fuel is replenished each year to maintain normal BWR operation) as happened with the Fukushima I nuclear accidents. Control rods are inserted from below for current BWR designs. There are two available hydraulic power sources that can drive the control rods into the core for a BWR under emergency conditions. There is a dedicated high pressure hydraulic accumulator and also the pressure inside of the reactor pressure vessel available to each control rod. Either the dedicated accumulator (one per rod) or reactor pressure is capable of fully inserting each rod. Most other reactor types use top entry control rods that are held up in the withdrawn position by electromagnets, causing them to fall into the reactor by gravity if power is lost. ER. SANDEEP KUMAR (DEPARTIv1ENT OF NECHANICAL ENGG.) Page 9

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