D
HISTORICAL DEVELOPMENT OF CURRENT COMMERCIAL POWER REACTOR FUEL OPERATIONS

There are 103 commercial power reactors operating in the United States at this time. Almost all of them are operating with spent fuel pools that are too small to accommodate cumulative spent fuel discharges. This short appendix was prepared to provide a historical background for power reactor fuel operations and pool and dry-cask storage of spent fuel.

D.1 DESIGN FOR A CLOSED FUEL CYCLE

The first large generation of commercial reactors in the United States were almost all light water reactors (LWRs), that is, nuclear reactors that use ordinary water to cool the core and to moderate the neutrons emitted by fission. The hydrogen atoms in the water coolant moderate, or slow down the fission-emitted neutrons to an energy level that is more likely to cause fission when the neutron strikes a fissile atom. These reactors were designed, developed, and licensed in the 1960s and 1970s, although many were not completed until the 1980s. Their design power output increased rapidly, as it did for non-nuclear power plants, in order to achieve economies of scale. Thus, the earlier plants in this generation were designed to produce 500–900 megawatts of electrical power (MWe) while later units increased to 1000–1200 MWe. The number of LWRs built and ordered by the U.S. industry began to approach 200. All of these plants were being designed for a closed fuel cycle, that is, for the uranium oxide fuel, enriched to 2–5 percent uranium-235, to be loaded and “burned” to a level of 20–30 gigawatt-days per metric ton of uranium (GWd/MTU), then reprocessed in commercial plants to separate the still usable fissionable, or fissile, materials in the spent fuel from the radioactive waste. The reprocessing plants would recover the fissile plutonium-239 formed from uranium-238 during reactor operations and residual fissile uranium-235 for use as fuel in LWRs and later in breeder reactors (USNRC, 1976).

By the mid-1970s commercial reprocessing plants were built, under construction, or planned in New York, Illinois, South Carolina, and Tennessee, with a combined projected capacity to reprocess more than 6000 MTU of spent fuel per year. For comparison, a large LWR discharges about 20 MTU of spent fuel at a refueling. By this time the price of fresh uranium was dropping and the cost of fuel reprocessing made it difficult for recycle fuel to compete with fresh fuel. Also, there was controversy about the risk of fissile material diversion if recycled plutonium was moved in commercial traffic. Both existing fuel reprocessing plants withdrew from licensing for technical reasons and then, on April 7, 1977, President Carter issued a policy statement that “we will defer indefinitely the commercial reprocessing and recycling of the plutonium produced in the U.S. nuclear power programs.” The statement went on to say: “The plant at Barnwell, South Carolina, will receive neither federal encouragement nor funding for its completion as a reprocessing facility.” After consultation with the White House, the U.S. Nuclear Regulatory Commission (USNRC) terminated its Final Generic Environmental Statement on the Use of Recycled Plutonium in Mixed Oxide Fuel in Light-Water Cooled Reactors (GESMO) proceedings.

Thus, the U.S. nuclear industry was immediately changed from a closed fuel cycle, with recycle, to an open or once-through fuel cycle with the fuel loaded into the reactor in



The National Academies | 500 Fifth St. N.W. | Washington, D.C. 20001
Copyright © National Academy of Sciences. All rights reserved.
Terms of Use and Privacy Statement



Below are the first 10 and last 10 pages of uncorrected machine-read text (when available) of this chapter, followed by the top 30 algorithmically extracted key phrases from the chapter as a whole.
Intended to provide our own search engines and external engines with highly rich, chapter-representative searchable text on the opening pages of each chapter. Because it is UNCORRECTED material, please consider the following text as a useful but insufficient proxy for the authoritative book pages.

Do not use for reproduction, copying, pasting, or reading; exclusively for search engines.

OCR for page 100
About this PDF file: This new digital representation of the original work has been recomposed from XML files created from the original paper book, not from the original typesetting files. Page breaks are true to the original; line lengths, word breaks, heading styles, and other typesetting-specific formatting, however, cannot be retained, and some typographic errors may have been accidentally inserted. Please use the 100 D D HISTORICAL DEVELOPMENT OF CURRENT COMMERCIAL POWER REACTOR FUEL OPERATIONS There are 103 commercial power reactors operating in the United States at this time. Almost all of them are operating with spent fuel pools that are too small to accommodate cumulative spent fuel discharges. This short appendix was prepared to provide a historical background for power reactor fuel operations and pool and dry-cask storage of spent fuel. D.1 DESIGN FOR A CLOSED FUEL CYCLE The first large generation of commercial reactors in the United States were almost all light water reactors (LWRs), that is, nuclear reactors that use ordinary water to cool the core and to moderate the neutrons emitted by fission. The hydrogen atoms in the water coolant moderate, or slow down the fission-emitted neutrons to an energy level that is more likely to cause fission when the neutron strikes a fissile atom. These reactors were designed, developed, and licensed in the 1960s and 1970s, although many were not completed until the 1980s. Their design power output increased rapidly, as it did for non-nuclear power plants, in order to achieve economies of scale. Thus, the earlier plants in this generation were designed to produce 500–900 megawatts of electrical power (MWe) while later units increased to 1000–1200 MWe. The number of LWRs built and ordered by the U.S. industry began to approach 200. All of these plants were being designed for a closed fuel cycle, that is, for the uranium oxide fuel, enriched to 2–5 percent uranium-235, to be loaded and “burned” to a level of 20–30 gigawatt-days per metric ton of uranium (GWd/MTU), then reprocessed in commercial plants to separate the still usable fissionable, or fissile, materials in the spent fuel from the radioactive waste. The reprocessing plants would recover the fissile plutonium-239 formed from uranium-238 during reactor operations and residual fissile uranium-235 for use as fuel in LWRs and later in breeder reactors (USNRC, 1976). By the mid-1970s commercial reprocessing plants were built, under construction, or planned in New York, Illinois, South Carolina, and Tennessee, with a combined projected capacity to reprocess more than 6000 MTU of spent fuel per year. For comparison, a large LWR discharges about 20 MTU of spent fuel at a refueling. By this time the price of fresh uranium was dropping and the cost of fuel reprocessing made it difficult for recycle fuel to compete with fresh fuel. Also, there was controversy about the risk of fissile material diversion if recycled plutonium was moved in commercial traffic. Both existing fuel reprocessing plants withdrew from licensing for technical reasons and then, on April 7, 1977, President Carter issued a policy statement that “we will defer indefinitely the commercial reprocessing and recycling of the plutonium produced in the U.S. nuclear power programs.” The print version of this publication as the authoritative version for attribution. statement went on to say: “The plant at Barnwell, South Carolina, will receive neither federal encouragement nor funding for its completion as a reprocessing facility.” After consultation with the White House, the U.S. Nuclear Regulatory Commission (USNRC) terminated its Final Generic Environmental Statement on the Use of Recycled Plutonium in Mixed Oxide Fuel in Light-Water Cooled Reactors (GESMO) proceedings. Thus, the U.S. nuclear industry was immediately changed from a closed fuel cycle, with recycle, to an open or once-through fuel cycle with the fuel loaded into the reactor in

OCR for page 100
About this PDF file: This new digital representation of the original work has been recomposed from XML files created from the original paper book, not from the original typesetting files. Page breaks are true to the original; line lengths, word breaks, heading styles, and other typesetting-specific formatting, however, cannot be retained, and some typographic errors may have been accidentally inserted. Please use the 101 D several consecutive locations to obtain maximum economic use of the fuel before it was finally removed as waste. The USNRC changed the legal definition of high-level radioactive waste to include the high-level waste from both nuclear fuel reprocessing and spent nuclear fuel. For this study, the significance of this closed fuel cycle design is that this entire generation of more than 100 reactors was designed with small spent fuel pools, relying on prompt shipment away from the reactor to the reprocessing plant to make room for later discharges of spent fuel. Early spent fuel shipping casks were being designed with active cooling systems to support shipment of fuel less than a year out of the reactor to a reprocessing plant. BOX D.1 discusses the spent nuclear fuel at reprocessing plants. Supplementary wet and dry storage systems had to be developed to receive the older spent fuel to make room for fresh spent fuel from the reactor. Many plants had to remove and modify the storage racks in their spent fuel pools to accommodate more spent fuel in the pool itself until licensed supplementary systems were available. D.2 RETRENCHMENT OF U.S. REACTOR PLANS As noted in Section D.1, in the 1970s the United States was building reactors at a high rate. Then, in the late 1970s, three factors produced a retrenchment in power reactor plans: rising interest rates, reversal of the U.S. fuel reprocessing policy, and the Three Mile Island-2 accident. D.2.1 Effect of Interest Rates Commercial power reactors have characteristically high initial capital costs. The regulated public utilities have had to raise the capital with various debt instruments; to build, license, and operate the finished plant for a time before it can be declared commercial; and to change the electricity rates charged consumers to retire the debt on the capital cost. The soaring interest rates in the United States during the late 1970s drove the costs of new nuclear plants that were under construction to extreme heights. This, combined with slackening demand for electricity, led to the cancellation of many plants, some even in advanced stages of construction. D.2.2 Effect of Reversal of U.S. Fuel Reprocessing Policy President Carter enunciated a change in U.S. policy for reprocessing of spent nuclear fuel in early 1977. Those reactors then operating and those under construction had to begin modifying their reactor fuel cycle design to go from the closed (reprocessing) cycle to a “once-through” fuel cycle. This induced the designers to go to higher levels of uranium-235 enrichment in the new fuel, but still within the 5 percent licensing limit. It also induced the designers to revise the core loading and operating plans in order to burn or use the fissile content of the fuel to the greatest extent economically possible since the fissile residue could not be retrieved by reprocessing. As a result, spent fuel burnup levels rose to levels that are now almost double the 20–30 GWd/MTU characteristic of the original closed fuel cycle. This results in an increase in the decay-heat power of the spent fuel assembly by the time it is put into print version of this publication as the authoritative version for attribution. the spent fuel pool.

OCR for page 100
About this PDF file: This new digital representation of the original work has been recomposed from XML files created from the original paper book, not from the original typesetting files. Page breaks are true to the original; line lengths, word breaks, heading styles, and other typesetting-specific formatting, however, cannot be retained, and some typographic errors may have been accidentally inserted. Please use the 102 D BOX D.1 SPENT FUEL AT NUCLEAR FUEL REPROCESSING PLANTS Up until the mid-1970s the commercial nuclear industry was expected to operate several nuclear fuel reprocessing plants to recover fissile plutonium from virtually all of the commercial spent fuel from U.S. reactors. These plants would use aqueous reprocessing methods developed by the Atomic Energy Commission (AEC). The recovered plutonium was to be used as mixed oxide fuel (PuO2 and UO2) in water reactors and, later, as fuel in breeder reactors. Each reprocessing plant had one or two storage pools to receive and store the fuel temporarily until it was reprocessed. No long-term storage of the spent fuel from commercial reactors was planned. Only two commercial reprocessing sites have received spent fuel, West Valley, New York, and G.E.-Morris, Illinois, The first commercial reprocessing plant began operations by the Nuclear Fuel Services Company on a site in West Valley, New York, owned by the State of New York. The State of New York licensed a low-level radioactive waste disposal site adjacent to the reprocessing plant. The West Valley plant had a reprocessing capacity of about 1 metric ton of uranium (MTU) per day. It operated at reduced capacity because there was not yet much commercial spent fuel to reprocess. In fact, about half of the spent fuel reprocessed there was from the last in the series of plutonium production reactors, the N-Reactor, at the AEC site in Hartford, Washington. This spent fael was provided to the West Valley plant to keep it working in the early days when little commercial spent fuel was available. The West Valley plant suspended operations in 1972 in order to expand its capacity to about 3 MTU per day. The work and the re-licensing effort went on until 1976 when the company withdrew its application for the new license and terminated reprocessing operations. The U.S. Department of Energy (DOE) took over the task of high-level radioactive waste retrieval and decommissioning under the West Valley Demonstration Project Act of 1980. About 137 MTU of commercial spent fuel remaining in the cooling pool was returned to its owners (USNRC, 1987). In 2003 the last of this spent fuel, about 25 MTU in two shipping casks, was shipped to the DOE-ldaho National Lab where it remains in dry storage in those casks. The General Electric Company built a nuclear fuel reprocessing plant at Morris. Illinois, near the Dresden Nuclear Power Station. The plant was expected to reprocess 3 MTU per day. When the G.E.-Morris plant was in its final testing in 1975, the company determined that its performance would not be acceptable without extensive modifications. The request for a reprocessing plant operating license was withdrawn and the plant was licensed only to possess the spent nuclear fuel that it was under contract to reprocess. After modifying the storage system in its below-grade pool to hold more spent fuel, G.E.-Morris has received and stores 700 MTU of spent fuel for various owners. Power reactors are refueled, and spent fuel is discharged to the storage pool, every one to two years. The decay- heat power of recently discharged spent fuel dominates the heat load of all the spent fuel in the pool, both freshly discharged and old, since the decay heat from a spent fuel assembly decreases by one to two orders of magnitude in the first year after it is removed from the reactor increasing the capacity of the spent fuel pool by reracking, that is, modifying the storage racks to provide for closer spacing of the fuel assemblies,1 allows older fuel to be print version of this publication as the authoritative version for attribution. accumulated in the pool rather than being removed for

OCR for page 100
About this PDF file: This new digital representation of the original work has been recomposed from XML files created from the original paper book, not from the original typesetting files. Page breaks are true to the original; line lengths, word breaks, heading styles, and other typesetting-specific formatting, however, cannot be retained, and some typographic errors may have been accidentally inserted. Please use the 103 D shipment or dry storage. Re-racking can make it more difficult to cool the freshly discharged fuel if there is catastrophic loss of the fuel pool water. D.2.3 Effect of the Three Mile Island Accident The final factor driving the retrenchment of the nuclear power industry was the Three Mile lsland-2 (TMI-2) accident that occurred on March 28, 1979, in Pennsylvania (Walker, 2004). In that accident a small failure in the reactor coolant system was compounded by operator errors to result in catastrophic damage; a partial core melt occurred. The inability of the operators to understand and control the events, and the confusion among the state, the USNRC, and other responsible agencies about public protection had a devastating effect on public trust in the safety of nuclear power. The USNRC escalated safety requirements after the TMI–2 accident. These new requirements substantially modified the operation of licensed plants, delayed completion of new plants, and further increased their construction costs. The accident also resulted in the retrenchment of nuclear power in the 1980s and led to the cancellation of many plants, decommissioning of some plants, and the sale of some plants to other owners. The fleet of operating U.S. reactors was reduced to the presently operating 103 described here. D.3 COMMERCIAL POWER REACTORS CURRENTLY OPERATING IN THE UNITED STATES All of the commercial power reactors operating in the United States are light water reactors. BOX D.2 describes the LWRs that are currently operating in the United States. D.3.1 Pressurized Water Reactors About two-thirds of the U.S. reactors are pressurized-water reactors (PWRs), dual-cycle plants in which the primary cooling water is kept under a pressure of about 2000 pounds per square inch absolute (psia) as it circulates to remove fission and decay heat from the reactor fuel in the core and carry that energy to the steam generators, to generate steam in the lower-pressure secondary loop. The reactor, primary loop piping, and steam generators are all located in the containment structure; the steam lines penetrate the containment carrying the steam to the turbine to generate electrical power. About one-third of the U.S. reactors are boiling-water reactors (BWRs), single-cycle plants in which the primary coolant of the reactor core is operated at about 1000 psia as it recirculates within the reactor core. The fission and decay heat generated in the core cause a substantial amount of the reactor coolant water to boil into steam that passes out directly from the reactor pressure vessel to the turbine-generator system. Plant differences stem initially from the different designs of the nuclear steam system supplier, the different designs of the architect-engineers that built the plants, and the owners that often specified additional modifications. print version of this publication as the authoritative version for attribution. 1The capacity of spent fuel pools has typically been increased by replacing the original storage racks with racks that hold the spent fuel assemblies closer together. The fuel assembly channels in these replacement racks typically have solid metal walls with neutron-absorbing material for nuclear safety reasons. This configuration inhibits water or air circulation more than the earlier configuration.

OCR for page 100
About this PDF file: This new digital representation of the original work has been recomposed from XML files created from the original paper book, not from the original typesetting files. Page breaks are true to the original; line lengths, word breaks, heading styles, and other typesetting-specific formatting, however, cannot be retained, and some typographic errors may have been accidentally inserted. Please use the 104 D BOX D.2 U.S. NUCLEAR POWER PLANTS In the United States, 32 utility companies are licensed to manage the 103 operating reactors. There are also 27 shutdown reactors in storage or decommissioning. These reactors are situated at 65 nuclear power plant sites across the United States; a plant site may have 1, 2, or 3 reactors. The fleet of 103 operating reactors in the United States is composed of the following: • 69 pressurized water reactors (PWRs) and • 34 boiling water reactors (BWRs). The containment design for PWRs is divided into dry (56 reactors), ice condenser (9 reactors), and sub- atmospheric (4 reactors) containments. Among the BWR containment designs, 22 reactors are of design type Mark I, 8 of Mark II, and 4 of Mark III, The PWRS operating in the United States were designed by three different nuclear steam system suppliers; Westinghouse Electric, Combustion Engineering, and Babcock & Wilcox. Most PWRs have what are called farge dry containments, that is, containment structures of about 2 million cubic feet volume that can absorb the rapid release of steam and hot water from a postulated rupture of the primary coolant system without exceeding an internal pressure of about 4 atmospheres. FIGURE D.1 illustrates a PWR in a large dry containment. Some PWR containments are essentially as large but use ventilation fans to maintain the initial containment pressure mildly sub-atmospheric to provide an additional pressure margin. Finally, one set of nine Westinghouse PWRs uses ice- condenser containment structures, in which the containment has about the same pressure capability but is smaller, relying on massive baskets of ice maintained in the containment to condense steam releases and mitigate the pressure surge. D.3.2 Boiling Water Reactors The BWRs in operation today were designed by the General Electric Company. They all use pressure suppression containments, two-chamber systems with the reactor located in a dry well that is connected to a wet well containing a large pool of water. In the event of a rupture of the reactor system in the dry well, the steam and hot water released are channeled into the water in the wet well, condensing and cooling the steam to mitigate the pressure surge. BOX D.2 lists the three successive generations of BWR containment design, and the number of each still operating. FIGURE D.2 illustrates three types of BWR containments: Mark I. Mark II, and Mark III, The Mark I containment is the most common type with 22 in operation. The reactor pressure vessel, containing the reactor core is located in a dry well of the containment in the shape of an inverted incandescent light bulb. print version of this publication as the authoritative version for attribution.

OCR for page 100
About this PDF file: This new digital representation of the original work has been recomposed from XML files created from the original paper book, not from the original typesetting files. Page breaks are true to the original; line lengths, word breaks, heading styles, and other typesetting-specific formatting, however, cannot be retained, and some typographic errors may have been accidentally inserted. Please use the 105 D FIGURE D.1 A PWR in a large dry containment. SOURCE: Modified from Duderstadt and Hamilton (1976, Figure 3–4). The dry well is connected by large ducts to the wet well, a large toroidal (i.e., doughnut-shaped) part of the containment that is partially filled with water. Gas and steam releases from an accident in the dry well would be passed through the connecting ducts into the water in the wet well, cooling the gas and condensing the steam to mitigate the accident pressure rise in the containment. The containment building Mark II BWR is similar to the Mark I except that in the Mark II containment the conical dry well is directly above the cylindrical wet well. Nine Mark II reactors are still operating in the United States. In the Mark III, the dry well around the reactor vessel is print version of this publication as the authoritative version for attribution. vented to the top of a cylindrical wet well that surrounds it. Four Mark III BWRs are currently operating. The entire dry well-wet well system is contained within a large steel containment shell and a concrete shield building. D.3.3 Reactor Fuel and Reactor Control TABLE D.1 presents the range of dimensions and weights for a wide variety of the LWR fuel assemblies used in the operating reactors. The spent fuel pools and the dry storage systems used at a reactor must be tailored to the specific fuel design for that reactor.

OCR for page 100
About this PDF file: This new digital representation of the original work has been recomposed from XML files created from the original paper book, not from the original typesetting files. Page breaks are true to the original; line lengths, word breaks, heading styles, and other typesetting-specific formatting, however, cannot be retained, and some typographic errors may have been accidentally inserted. Please use the 106 D FIGURE D.2 Three types of BWR containment system: Mark I, Mark II, and Mark III. SOURCE: Modified from Lahey and Moody (1993, Figure 1–9). The fission process is controlled by the reactor operators through the use of neutron-absorbing materials. The primary control is an array of control rods or blades that can be withdrawn from the core to the degree needed. In the PWRs, the control rods are moved within selected empty tubes within the assembly. In the BWRs, cruciform (cross-shaped) control blades are moved across the faces of the fuel assembly, typically narrower than those in a PWR fuel assembly. Reactor fuel designers also use burnable poisons within the fuel assembly to control the fission process. These poisons are placed in appropriate amounts within the fuel assembly so that they burn away, making the fuel assembly more reactive, as the continued fission process is making it less reactive. PWRs also use neutron control by dissolving neutron-absorbing sodium borate in the reactor coolant, gradually lowering the concentration from the peak after refueling to the minimum before the next refueling. REFERENCES American Nuclear Society. 1988. Design Criteria for an (independent Spent Fuel Storage Installation (Water Pool Type): An American National Standard. ANSI/ANS-57.7– 1988. American Nuclear Society. LaGrange Park, Illinois. Duderstadt, J.J. and L.J.Hamilton. 1976. Nuclear Reactor Analysis. John Wiley& Sons. New York. print version of this publication as the authoritative version for attribution. Lahey, R.T. and F.J.Moody. 1997. The Thermal Hydraulics of a Boiling Water Nuclear Reactor. Second Edition. American Nuclear Society. La Grange Park, Illinois.

OCR for page 100
About this PDF file: This new digital representation of the original work has been recomposed from XML files created from the original paper book, not from the original typesetting files. Page breaks are true to the original; line lengths, word breaks, heading styles, and other typesetting-specific formatting, however, cannot be retained, and some typographic errors may have been accidentally inserted. Please use the 107 D USNRC (U.S. Nuclear Regulatory Commission). 1976. Final Generic Environmental Statement on the Use of Recycled Plutonium in Mixed Oxide Fuel in Light-Water Cooled Reactors (GESMO). NUREG-0002. Washington, DC. USNRC, 1987. Case Histories of West Valley Spent Fuel Shipments. NUREG/CR-4847. January. Washington, D.C, Walker, J.S. 2004. Three Mile Island: A Nuclear Crisis in Historical Perspective. University of California Press. Berkeley, California. TABLE D.1D.1 RangeDimensions andand Weights for Light Water Reactor Fuel Assemblies UsedOperating Reactors TABLE Range of of Dimensions Weights for Light Water Reactor Fuel Assemblies Used in in Operating Reac in the the United States. in United States. PhysicalClharateristicsof of TypicalWRR uel Assemblies s Physica Characteristics TypicalL LWF Fuel Assemblie Reactor Type BWR BWR PWRPWR PWR PWR PWR PWR PWR PWR PWR PWR PWR W R PWRPWR PWR W R PWRWR PWR R BWR Reactor Type BWR P P P PW B&W R&W B&W GE GE GE GE W W wW Fuel Designer esigner GEGE GE E R&W W WW WW W WW Fuel D G Fuel Rod Rod Array 7x77×7 8x8 8×8 15x15 Array 15,15 17x17 17×17 14×14 116.16 14×14 4 14×144x1415×15 15x1515×15 15x1517×17 17x1717×17 Fuel 14x14 6x16 14x1 1 17x17 Active Fuel uel Length (in.) 144 Active F Length (in.) 144 144 44 144144 143143 137137 160150 120120 144 144 121 121 144 141 144 144 18868 1 1 Nominal Envelope (in.)2 5,438 8 Nominal Envelope (in.)2 5.43 S.45 .47 8.536 8.536 8.25 8.25 8.25 7.763 3 7.763 7.763 8.449 9 8.426 8.426 8.426426 8.426426 8.536 8.536 8.25 8. 8. 5 7.76 8.44 Fuel Asembly Length h (in.) 176 176 166 166 157 177166 7 1157 137177 (in.) 176 166 1ST 161 137 160 160 – 13 61 160 160 - Fuel Assembly Lengt 176 Weigth Weight (Iba.) 600 600 1.516 600 (lbe.) 600 1.502 581 kg - 501 kg 573 kg 594 kg 654 kg 665 kg 501 1.516 1.502 581 kg - kg 573 kg 594 kg 654 kg 665 kg _ - Fuel Rod Number r Numbe 494 9 63 63 204 26 264264 164 224-236 6 160180 179179 20420 4 204204 254254 2S4264 224-23 164 Length (in.) 16363 1 153 153 147147 161 _— -— Length (in.) 161 l27 152 127 152 152152 152 127 - 152 127 Pitch. Square (in.) (in.) 0.7388 0.73 0.640 0.580 0.506 0.501 0.580 6 0.55 0.506 0.556 Pitch.Square 0.640 0.588 0.556 0.556 0.563 0.563 0.563 0.496 0.496 0.563 0.496 0.568 0.501 0.496 O.D. . O.D (in.) in.) 0.570 0.570 0.4900.493 0.4300.430 0.3790.379 0.4400.440 0.382. ( 0.382 0.422 2 0.422 0.422 0.422 2 0.422 0.374 0.260 0.374 0.42 0.42 0.422 0.360 Clad Thickness (mils.) 35.5 Clad 22.5 Thickness (mils.) 35.5 34 26.523.5 23.5 26 26 2525 16.5 24.33 16.5 24,3 3 22.5 22.5 34 26.5 24. 16.5 16.5 24. 22.5 Z Clad MaterialMaterial 2 r2 Zr 4 r 4 Zr Zr 4 4 Clad Zr Zr 2 Zr 2 4r 4 Zr 4 r 4 Zr 2r 4 Zr 2r 4 4 mt Zr4 4 ast sst ZR Zr Zr 4 4 Z Z 4 Z sst Zr 0.488 Pellet O.D. .D. ( (in.) Pellet O in.) 0.488 0.416 6 0 370 0 0.3232 0.3232 0.3795 5 0.325 5 0.3835 0.3835 0.3659 9 0.3836 0.3835 0.3659 9 0.3225 5 0.3088 0.37 0.379 0.32 0.365 0.41 0.365 0.322 0.3088 PelletLength (in.)(in.) - Length Pellet _ 0.376 0.680 0.390 0.600 0.390 0.600 0.600 0.600 0.600.530 0.530 0.530 0 0.600 — — 0.600 0 0.375 0.650 0.53 0.600 Cap. Radial (mils.) 5.5 4.5 4.5 3.5 3 13.1 4.34.3 3.S 3. 5 2.8 2.8 3.8 3.8 2.8 2.8 3.8 3.8 3.33.3 3.33.3 Gap, Radial (mils.) 5.5 3.5 Density (STD) - Density (STD) - _— 92.5-95.0 93.5-95.0 0 93.0-96.0 94.75 94.75 93.0-94.0 93.0-94.0 92.0 92.0 95.0 92.0 93.0-94.0 95.0 95.0 92.5-95.0 93.0-94.0 92.0 93.0-95. 0 93.5-95. 95.0 Poison Gd2O21O1 Gd2O2 Nose 2 Noon B,C/A1202 4C/Al2O3B4B4C/Al2O3 - C/Al2O2 _ -_ _ _ _ _ Poison GD Gd1O B None None - - - - 25 Nonfueled Rods 16 11 17 17 16 17 Number O 25 25 66 16 21 21 21 25 25 25 Number 0 17 21 Material - Zr 2 Zr ZR Zr 4 Zr4 Zr 4Zr 4 4 4 Zr 4 4 304 ssi Zr 4 4 304 304 sst Zr 4 r 4 Zr sst 4 4 Zr 4 r 4 Material - 2 Zr Zr Z 2r 304 sst Zr Z Spacer Grids Spaoerr Gride Number 7 7 8 8 8812---8-- .12 _ _ • _ _ _ _ NUMBER - 7 7 8 Material Inancel InconelInancel X Xnconel X Inconel 718Inconel7Zr 4 4ZrZr -4 --- Material I Inancel 718 Incood 718 18Zr 4- - X SOURCE: American Nuclear Society (1988). SOURCE: American Nuclear Society (1988). print version of this publication as the authoritative version for attribution.