B Reactor Museum AssociationB Reactor Museum AssociationAuthor: Michele S. Gerber, Ph.D., Facility Operations Division, Westinghouse Hanford Company, April 22, 1993 (approved for public release May 19, 1993)
Following the completion of Project CG-558 in November 1956, B Reactor scientists believed that power levels between 1,500 - 1,750 MW would be possible, although "not feasible" because they would produce excessive fuel element rupture rates. In other words, they believed that the true technical limits of the reactor would never be reached or known, because the "rupture control limit" (i.e., the level at which ruptures become so numerous and disruptive as actually to decrease plutonium production rates) would be reached first. Consequently, B-Pile was scheduled to begin operations at 850 MW and to operate at that level for four months, until several charges had reached their goal exposures. Thereafter, power levels would be raised in increments of 100 MW, until an "optimal" fuel element rupture rate (i.e., 2-3 ruptures per month) was attained. However, the drive to higher and higher power, production, and test levels was such that the first restart plan was dropped and B Reactor resumed operations in November 1956 at 950 MW. It used solid-core uranium fuel slugs of 1.44-inch O.D., generally 8.9-inches long each, with 32 slugs per fuel column. The coolant annulus around the fuel rods was approximately 90 mils, and the coolant pressure at the front face risers was 542 psig. The coolant flow volume stood at 71,000 gpm, and the approximate annual fuel throughput was 560 tons of uranium. Due to the long outage of the final Project CG-558 modifications, the reactor's TOE rating was only 75 percent.
Following the restart, B Reactor's power levels rose steeply. By mid-1957, the power level stood at 1,250 MW. Two years later, the operating level had climbed to 1,640 MW, while the administrative (maximum allowable) power level imposed by the AEC was set at 1,900 MW. The latter was raised to 2,090 MW in January 1961. Most of the increase in power level was made possible by the use of I&E fuel elements, which were loaded into B Reactor during the latter half of 1958. These fuel elements had a vastly augmented cooling capacity, due to their much higher surface (cooling area) to volume ratio. They also provided less hydraulic resistance to the coolant flow, but required that more process water be pumped around and through them. They removed a recurrent worry among HW operators that higher operating temperatures would transform the interior of the fuel slugs into the gamma phase, with all of the attendant problems of uranium grain size and dimensional instability, fuel rupture, and possible fuel melt. Another innovation that allowed higher operating power levels in the late 1950's and early 1960's, specifically in the hotter summer and autumn months, was artificial cooling of the Hanford Reach of the Columbia River by a controlled spill of cold water from the bottom levels of Lake Roosevelt, the reservoir behind the Grand Coulee Dam. The cooling program began during the summer of 1959, and continued through the summer of 1965. It cooled the Columbia's water in the vicinity of Hanford reactors by an average of 1°C, and by a maximum of 2.7°C during September 1960. The lowered inlet water temperatures made possible higher power levels in reactor operations, while still not exceeding the maximum bulk outlet temperatures that were allowed.
Other operating efficiencies that came quickly in the late 1950's and early 1960's resulted from the gradual replacement (on an as-needed basis) of aluminum process tubes with zircalloy-2 tubes. The more tensile strength of zirconium allowed tubes that were thinner and that consequently had a greater cooling annulus. Also, the melting point of zirconium is approximately 1200°C higher than that of aluminum, making the zircalloy-2 tubes safer in events involving the loss of coolant. Additionally, a "probolog" instrument that checked for process tube leaks and corrosion by means of a radioactive tracer, was developed at HW in about 1956. An important operating efficiency came during the same period when the use of self-supported ("projection" or "bumper") fuel elements began to be discussed, at least as a concept, at HW.
At the same time that I&E fuel elements offered previously unimagined advantages in terms of plutonium production gains, the higher power levels brought a multiplicity of operating challenges to B Reactor. Time operating losses due to slug ruptures increased 40 percent in 1959, as a much greater proportion of the ruptures that did occur were serious, side-split types that required tube removal and could not be handled with the "quickie" procedure. Overall, the reactor's TOE rating fell to approximately 70% (7,061.1 hours of operating time) that year. The vastly increased coolant flow also corroded more and more process tubes, strained the rear face piping systems and fittings, stressed the downcomer, and wore large and destablizing leaks in the effluent disposal system. New studies were undertaken in process tube film deposition and internal tube corrosion. It was learned that I&E fuel rods fitted into the process tubes in an asymmetrical fashion, causing more heat and corrosion damage at the top of the tube. Consequently, a new top-of-annulus (TOA) corrosion pattern was identified at HW. Between 1955 and 1957, most of the original process tubes in B Reactor had been replaced, but a 1960 HW study found that even the "average" second generation process tube had lost about 25 percent of its thickness to internal corrosion. During 1959 and 1960 combined, such corrosion led to the complete failure of 35 process tubes in B-Pile, causing worrisome wetting of the graphite and the loss of over 275 hours of production time. By this time, tritium (with a working limit of 20 curies per day) was being injected into coolant water as a tracer to detect process tube leaks.
Concomitantly, the 100-B last-ditch cooling system, electrical system, and much of the pile instrumentation was rendered obsolete and inadequate by the new operating levels. One 1960 study of the safety and reliability of the reactor electrical power supply system concluded that equipment was approaching its "maximum capability," and that "operation of the present loop under critical power conditions is unsatisfactory." This document recommended immediate increase in transformer capacity at the 100-B/C Area. During the same years of the late 1950's and early 1960's, other studies were launched in graphite distortion, since the three-inch variation in stack elevation measured in 1958 at B Reactor was determined to be one of the most severe of any HW pile. By 1963, the overall deflection at the old reactor had increased to five inches, and flexible VSRs had to be installed.
In the same time period, safety reviews by the ACRS and HW scientists called for a mounting list of improvement projects, including confinement of reactor exhaust gases, and systems and instrumentation upgrades of many types. One 1958 site study listed nearly 20 reactor-area systems that would need attention "in the foreseeable future." This list included the rebuilding of water plant filters, repairing the underground raw water pipes, painting and applying cathodic protection for the high (emergency coolant) tanks, repairing 183 Building valves, replacing 190 Building tank bottoms, dredging the 181 Building forebays, repairing the effluent systems, replacing the pump impellers in the 183 and 190 Buildings, replacing reactor thermocouples, replacing process tubes (an ongoing but growing need), boring the graphite channels to reduce process tube curvature, replacing the gas seals on gunbarrels and VSRs, freeing stuck gunbarrels, replacing inlet and outlet process tube fittings, installing and maintaining operational "C-D" equipment, renovating the Ball-3X system components and electric circuitry, repairing various instruments, installing exhaust containment or confinement systems, and further increasing water plant capacities. Between 1957 and the mid-1960's, many key projects were scoped, designed, and carried out in the 100-B Area.
The initial operating difficulty noted after B Reactor's post-Project CG-558 restart was cavitation in the 183-B and 190-B Annex process pumps and vibration and extreme noise in the process piping and valve pits at the pile's front face. The cavitation caused serious pitting in the first stage impeller vanes of the pumps, so that wearing ring inserts ("choke" rings) had to be welded into the pumps. These rings failed during 1957 and 1958, until finally the impellers were replaced and straightening vanes were installed in the pump suction inserts during 1959 and 1960. Also, several pump motors failed, and coils (and sometimes entire motors) were substituted throughout the 1958-1960 period. Additional pipe supports and concrete anchors were installed in the front face valve pits, acoustical instrumentation was provided, and new orifices were emplaced downstream of the bypass cone valves.
Another key difficulty identified during 1957 and 1958 concerned the rise in bulk outlet water temperature, caused by the new power levels, and its effects on rear crossheader pressure and on the B Reactor downcomer. In 1958, HW officially replaced the trip-before-boiling limit for bulk exit water temperature with the trip-after-instability (TAI) limit. This new specification meant that bulk outlet water would be allowed to reach the point where nucleate boiling (pre-bulk boiling) had begun. By 1960, the original B Reactor process flow had increased from 30,000 gpm to 80,000 gpm, with coolant delivery pressure up from 350 psi to nearly 600 psi. The bulk outlet water temperature had risen from 65°C in World War II, to 95°C. Such changes, of course, increased pressure in the rear crossheaders, causing vibrations, cavitation, and actual movement of component parts. The possibility that such torsion and twisting might cause crossheader failure, with attendant flooding of the rear pile area, was seen as serious enough that a mockup study of crossheader stresses was undertaken in the 100-D Area Flow Laboratory. The amounts of both horizontal and vertical expansion in the crossheaders from their concrete support walls, the torsion in the risers, and the effectiveness at various bulk exit water temperatures of sliding supports that had been installed from the crossheaders to the vertical centerline of the reactor, all were tested. Also, the maximum relocation, in three dimensions, of the expansion loops at the interface of the rear crossheaders and risers was measured.
Additionally, the welds and piping at four critical points in the downcomer were examined: the approach section, the top of the "elbow" of the downcomer, the first cascade tray, and the bottom walls. The component most at risk was seen as the elbow, or the point where the effluent made a 90° bend from horizontal flow to vertical flow through the downcomer. It was known that hotter effluent temperatures would greatly increase the water's velocity here, with the same mass flow. Again, the possibility of component failure, with consequent flooding of the rear reactor area, was seen as a distinct possibility. An overall 1960 HW study of the rear face piping on the older reactors found cracks in the risers and stress corrosion, caused by leaks and buildup of corrosive scale, to the extent that "the probable current quality of all rear face...piping is a real cause for concern." Specifically, at B Reactor, the study found the crossover piping cracked near the downcomer, stress corrosion and cracked welds in the outlet connectors and risers, and a crossheader fitting that had failed completely in May of that year. Nearly 600 hours had been spent in repairing such problems at B-Pile from 1957-1959, according to this document. It recommended replacement of all rear face nozzles, connectors, seals, crossheaders and supports, risers, crossunder and crossover lines, and rear face thermocouples. Another study that same year advised the perforation of all downcomer baffles, removal of existing baffle vent pipes, reinforcement of the downcomer's concrete walls, and venting (steam release) from the effluent lines adjacent to the downcomer.
As a result of these studies, several solutions were proposed and modifications were undertaken. It was known that if the rear face piping itself could be "pressurized" with flow restrictions, the water temperature at which boiling would occur would rise. However, this solution would require a steam disposal mechanism. The three such mechanisms proposed by HW operators were the addition of cold "quench" water upstream from the flow restriction (to prevent steam formation), venting the steam to the atmosphere, or harnessing the steam to produce electrical power. The venting idea was not considered desirable due to radiological and meteorological effects, and electrical power recovery was not considered technically feasible until at least 1961-1962. Therefore, a combination of other solutions was adopted. In 1959 and 1960, crossheader pressure differential indicators and an audible alarm system to signal significant changes in crossheader pressure were installed in B Reactor. In 1961, new digital outlet temperature monitoring equipment, with a much higher degree of accuracy than the old temperature logging equipment, was tested at B Reactor. That summer, it was installed on a permanent basis. Additionally, the process tube outlet fittings, pigtails, and nozzles were enlarged, in order to reduce bulk effluent temperatures by altering the mass-to-velocity ratio. Larger fittings would decrease the velocity for the same mass of water, thereby lowering the temperature. In the original test of this idea, the standard tube outlet fittings of 0.469-inch I.D. (15/32-inch) were reamed out to just over one-half inch I.D. A process flow rate increase of between 5.1 - 8.2 percent was achieved, but the reaming process weakened the fittings. In the final 1962 installation, new, specially fabricated, enlarged fittings were used.
Another modification undertaken at B Reactor to help solve the rear face piping difficulties was the installation of crossunder lines in 1960. This modification not only diverted some of the coolant flow through newer, stronger piping, but it was used in lieu of installing operational "C-D" equipment, which was considered too expensive for the five oldest HW reactors (see Section 3.11). In the then-current, manual "C-D" system, the rear face work area needed to be entered repeatedly by personnel, in order to operate individual crossheader valves and allow crossheaders to drain through the risers. Then, after each crossheader was valved to the drain, the corresponding front crossheader had to be valved to provide minimum water flow. Such procedures required close coordination between front and rear face operating crews, and therefore frequently interrupted other maintenance activities and added to the reactor's down time. In the new system, remotely operated crossunder drain lines were installed, along with the necessary piping, valves, instrumentation, and electrical facilities. The new crossunder lines at 105-B were connected between the base of the near-side rear process riser and the downcomer. Using this system, the entire reactor could be valved to a minimum flow in one continuous operation prior to "C-D." In one entry to the rear face area, all crossheader valves on the side opposite the drain line could be closed, thus isolating one riser. The crossunder valves then could drain the rear face piping, so that all nozzle caps could be removed. Finally, normal operating flow could be re-established in one valving procedure prior to re-start.
Beyond the difficulties imposed on B Reactor's rear face piping by higher process water flows and temperatures, lay the challenges inflicted on the old (albeit modified) effluent disposal system. By late 1959, the 107-B retention basin leakage rate, despite earlier repairs, had reached nearly 10,000 gpm! This huge rate was causing an increase in the height and temperature of local groundwater, and, due to the existence of unique shore line springs in this area, actually was transporting effluent leakage upstream, in a westerly direction. Such transport was routing the effluent through an opening between the walls of the B and C sections of the 181-B/C River Pump House, into the 181-B/C forebay. The quirks of circulation were such that the leakage flowed more to the 181-B side than to the 181-C side, but the overall condition was such that engineers warned that voids might develop under the entire 181-B/C structure, thus rendering it unstable. Additionally, the leaking effluent heated the 181-B/C intake water by about .5°F, thus limiting the overall production capacity of both B and C reactors when they were operating on bulk outlet water temperature restrictions. Furthermore, diversion of unusual effluents from the 107-B basin to the 107-B crib was impossible, because the basin leakage rate was faster than the pumping rate.
As a solution to this leakage problem, Project CGI-889 was authorized in March 1960, with the work actually begun in December 1960 and completed in the late summer of 1961. This project laid approximately 550 feet of new, 66-inch steel pipe that routed normal effluent from B Reactor to the 107-C West (CW) retention basin. Thereafter, normal effluent from C-Reactor went only to the 107-C East (CE) basin. Two new steel diversion boxes and associated motor-operated sluice gates also were emplaced, along with power supply and controls (operated from the 105-B control room), and a new inlet box and expansion joint at 107-CW. The controls and piping at the diversion boxes were equipped to divert abnormal effluent from 105-B to the existing 107-B basin, and abnormal effluent from 105-C to the 107-CW basin. In the latter case, normal B Reactor effluent would be temporarily diverted to the 107-B basin. A pipe wye and anchor also was installed for the tie-in of the new 66-inch effluent line to the existing line from the 107-B basin to the 1904-B outfall line. New fission product monitoring instrumentation, and basin liquid level alarms also were added at 107-CE and CW.
The other most major consequence of higher, post-Project CG-558 power levels at B Reactor concerned pile exhaust gases. Such emissions became the subject of intense debate and scrutiny in ACRS deliberations that took place throughout 1958. At that time, the AEC was under heavy pressure from local and state-wide agricultural and Congressional interests to open to farming land on the Wahluke (North) Slope across the Columbia River from the HW reactors. Studies conducted in secret for many years by Hanford scientists had shown consistently that, in the event of any airborne fission product release from one or more piles, "such a release may exercise its effect mainly over the Wahluke Slope...the Slope may be the principal off-site region of radiological hazard involvement in the event of...reactor incidents," due to proximity and "meteorological conditions."
As a compromise measure, the AEC in December 1958, announced the release of both secondary (non-central) zones of the Wahluke to the U.S. Department of the Interior, Bureau of Reclamation, for development as farmland. At the same time, it authorized Project CGI-791, the phased-in installation of rear reactor area fog sprays and an exhaust filtration system that would entrap a small percentage of the pile noble gases (krypton 85, argon 39, 41, and 42, and xenon-135), 70-95 percent of the halogens (iodine 131 and bromine 82), and most of the remainder of the particulates and aerosols in reactor gases (including cesium 137, tellurium 129, selenium 79, ruthenium 103/106, and others). Lastly, in a follow-up project between 1966-1968, front face fog sprays were installed in the HW reactors that still were operating. According to Hanford scientists, the confinement endeavor was a "high priority" for the AEC, and was expected to "greatly reduce...the...probability of significant contamination of the Wahluke Slope." However, they admitted in secret, "effective confinement in the event of a full reactor meltdown cannot be assured." The plan was adopted in lieu of other options that were deemed too expensive, including the emplacement of a full steel containment dome over each HW reactor.
Project CGI-791 got underway at B Reactor in late 1959, and the rear face fog spray system (Phase I) was manually operable by September 1960. By that time, Phase II, construction of the below-grade filter building (117-B) duct work and exhaust plenum was well along. When it was completed, the 117-B site was backfilled, and a berm was constructed around it. The 117-B structure housed a series of horizontal absolute filters for fine particulate removal, and activated charcoal filters for halogen removal. A 119-B sampling building was also provided to house instrumentation that would indicate water flow and high/low pressure through the fog spray system, pressure drop across the filters, air pressure differentials, and that would detect I-131. During Phase III, the tie-ins, exhaust fan modifications, and other work associated with routing pile gases through the new system and back out the existing ventilation stack were completed. During the tie-in phase, two 20-foot high, temporary reactor exhaust stacks were constructed, and for the 8-16 days that these stacks were used to remove pile gases, the operations of 105-B were scaled down for safety reasons. Once the new system was operational, the entire air flow was maintained at a slight negative (internal building) pressure, and steam-driven emergency power systems were installed. A crib was constructed to receive liquid effluents generated within the filter system.
Another key system that was rendered inadequate by the high power levels and flow demands that followed Project CG-558 was the "last ditch" or backup coolant system for B Reactor. This reserve supply was supposed to provide enough coolant for 30 minutes of normal reactor operation, five hours of transition flow to shutdown conditions, and then 24 hours of shutdown cooling (for a total of 29.5 hours of cooling). In an emergency, the high tank would empty first, and then the export water lines from one or more adjacent reactor areas would cool the stricken reactor. Operations under the new power levels that followed Project CG-558 quickly rendered the old last ditch system inadequate to supply the needed flow, especially during the summer months when solar heating pushed water temperatures up to 36°C in the high tank, and up to 27°C in the export lines. By late 1959, one HW study estimated that, at the then-current flow rates, the actual last ditch coolant capacity available to the 100-B/C totalled only 21.8 hours. It described the sufficiency of the last ditch system as "questionable at the present time," and "definitely...inadequate" in view of planned power expansions. The following year, after rust and corrosion build-up inside the high tanks and export lines were removed with a chemical cleansing, it was found that corrosion pits had nearly penetrated through tank and pipe walls. At that time, a report on the system piping stated that it "lacks the degree of reliability felt to be required." New piping, high tanks, spare pumps, surge suppressors to prevent water hammer in the flow, and cone valves to shut off export water to the 200 Areas completely in case of a 100 Areas emergency, were recommended. Five "degrees of improvement" were postulated, with cost estimates ranging from $860,000 to $5,300,000 for full expansions, complete with earthquake protection, to the system.
In 1961, in Maintenance Job Authorization 36 (MJA-36), two 3,000-gpm pumps that supplied the B-Area export water system were exchanged for two large 6,000-gpm pumps. In 1963, under Project CGI-905, the high tanks were chemically cleaned and then painted, new strainers and orifices were installed, surge suppressors were emplaced to prevent water hammer, and new electrical, bleeding, and temperature sensing instrumentation was installed in the 100-B/C Area last ditch and export systems. Additionally, one of the backup steam turbine pumps was placed into continuous service to provide extra, ongoing flow capacity. Two years later, after F and H reactors had been deactivated and their export water systems shut off, supplementary last ditch piping and power systems were required at B, C, and D reactors. In Project CAI-142, a four-engine, diesel-powered emergency backup pumping system, as well as a steam turbine pump excessed from the 182-H facility, was installed in the 182-B Building, and a diesel oil storage tank was located outside. Concrete reinforced piping to 105-B, C, and D and electrical control systems also were emplaced, to augment the diminished emergency water supply, and additional export water was made available to the 200 Areas.
Many other minor projects and modifications were performed at B Reactor in the late 1950's and early 1960's, as I&E fuel rods and the gradual emplacement of zircalloy-2 process tubes made possible the dramatic increases in power levels. The need for more coolant flow through 105-B led to many studies of how to supply the additional needed process water. Although electrical and pumping expansions were considered, the solution implemented in 1960 was the installation of a 30-inch cross-tie between the 183-C filtered water plant and the 183-B clearwells. This modification allowed some of the surplus water pumping and filtration capacities of the 100-C systems to supply coolant to B-Pile. In April 1960, after nearly four years of reports detailing the inadequacies of the current system and after many project proposals and revisions, the Ball-3X electrical system at B Reactor was upgraded. Inadvertent ball drops cost many hours of down time each year (more than 150 hours for a complete, full-reactor drop), and the recovery and sorting of hot balls after these drops added undesirable personnel exposure. As part of an overall "exposure reduction program" undertaken by HW's Irradiation Processing Department, dual and independent Ball-3X power and control systems were installed, along with continuous monitoring equipment and some changes in the layout of system controls and components to provide easier access for maintenance operations. In 1965, a four-inch discharge chute was installed to carry irradiated balls from the top of B Reactor to a storage pad outside of the building. The following year, baffles were installed in this chute and it was connected to a ball holding tank with one-foot-thick concrete block shielding all around it.
Another outgrowth of the "exposure reduction program" at HW was the installation of a dummy slug decontamination facility in Project CGI-871 at B Reactor in 1961. The system, located in the rear face area, consisted of remotely controlled buckets, chutes, and a dummy elevator complete with noise abatement equipment. Its use allowed the recovery and re-use of approximately 95 percent of the non-expendable dummies needed at the reactor. While a pilot demonstration of the system showed that about 25 percent of expendable dummies could be recovered, the recycling process was not time or cost effective, so all of the expendable dummies were routinely buried. The decontaminating agent was nitric acid, more expensive but also much more effective than the oxalic acid that served as the alternate choice in pilot testing. In 1960, noise abatement equipment, consisting of acoustical wall panels and pump motor shields, was installed in the 182-B Reservoir and Pump House. A year later, remotely operated poison spline coiling (removal) equipment, along with a recessed spline tank in the "C" elevator, was installed at B Reactor. At the same time, aluminum railings and floor grates replaced the steel ones on the "C" platform, as new equipment had made the elevator too heavy for designed weight limits. The following year, the exterior of the 190-B storage tanks were painted to prevent rust accumulation and potential rust-through. Between 1963 and 1965 (ongoing), the interior of the four 190-B process water storage tanks also were cleaned, sandblasted, and painted with two coats of a vinyl-resin, corrosion-resistant paint.
During 1963 and 1964, rear header expansion joints (bellows) were installed at B Reactor, in order to reduce the high corrosion and failure rate being experienced with the rear stainless steel connectors. Both cyclic stress, caused by rear crossheader vibration, and thermal expansion caused by high bulk exit water temperatures, were undermining the strength and performance of components at the crossheader/riser junction. In 1966, it was found that longitudinal expansion of the rear crossover piping and vertical expansion of the rear riser again had stressed the juncture materials. Two years later, flexible crossheader supports (permitting movement in a vertical plane that corresponded to the downward movement of the risers from thermal expansion), floating riser supports, new seals at the downcomer elbow (approach and crossover points), and additional vents in the approach to the elbow, all were installed at B-Pile.
During the same years that systems modifications were being made so rapidly at B Reactor, many other enhancements were provided to pile control and safety instrumentation. In 1958, attempting to reduce spurious reactor scrams caused by power surges and/or minor variations in the flux meters and controllers, a new flux monitor "dual trip" system was installed at B-Pile. New controllers and circuitry modifications were emplaced, along with bypass switches, relays, and control board upgrades. Under the "dual trip" system, two concurrent "trips" above or below pre-set flux limits, as registered on the flux monitors, were needed to initiate an automatic reactor scram. The following year, in a separate project, sub-critical neutron flux monitors were installed in test holes A and D of B Reactor. The purpose of this new instrumentation was to monitor the rate-of-rise carefully during startups, or during the "high sub-critical" periods, when rapidly changing power levels could cause spurious scrams, the formation of localized hot spots, and other operating abnormalities. It was justified on the basis of increased reactor safety during startups as the result of continuous presentation of data. It could signal, for example, inadvertent approaches to criticality caused by gross loading errors in the fuel configuration. The previous equipment was not able to calculate neutron flux levels (density) accurately during times of low but quickly shifting power conditions. The new instrumentation consisted of neutron sensitive chambers, log rate meters, recorders, amplifiers, and an alarm relay system. Additional rate-of-rise metering equipment was installed in late 1960 and early 1961.
In April 1960, automatic gas make-up equipment was installed at the 100-B Area, with components in both the 105-B and 115-B Buildings. The system consisted of an electronic analyzer for the measurement of both the gas mixture and the pressure/flow rate, as well as valving to preclude fluctuations and maintain constant gas pressure and flow. Soon after the installation however, it became apparent that the control valves were not sized in proportion to the flow characteristics of the gas, and many system malfunctions occurred. Several thermoswitches failed, and the composition of the gas after varied three to five percent (plus or minus) away from the desired levels. Such variations had serious consequences for reactor heat levels (and consequent power level limitations) because of the vastly different heat removal capacities of the two reactor gases, CO2 and He. Finally, in 1962, new prototype gas control instrumentation was installed in the 115-B Building, to serve C-Reactor. It proved very satisfactory, and soon such equipment also was emplaced to serve B Reactor. In 1963, pressure monitoring system improvements were emplaced at B Reactor, in order to reduce false reactor scrams caused by component failure in the Panellit gauges and their related circuitry. A secondary purpose of the system replacements was to standardize the gauges, switches, tubes, relays, and other equipment across all HW reactors, in order to achieve easier maintenance.
During 1962 and 1963, important restorations and improvements were made to the gamma monitoring, rupture detection equipment at B Reactor. According to HW management, caliche deposits along sample lines and valves had caused couplings on the rear crossheaders to split, thus degrading the safety situation to the point where "current reactor operating conditions cannot be long supported, much less improved upon." Rear face sample lines from the crossheaders to the sample rooms were replaced, gamma monitoring equipment in the X, Y, and Z sample rooms at B-Pile were consolidated into the X room (where sample room piping was replaced and pulse height signal generators and oscilloscopes were seated), heat exchangers, automatic flow regulators and shutoff valves were installed on all sample lines, combined isokinetic flow probes and shutoff valves were emplaced on sample tops, and new cooling water supply and drain lines were provided to the affected areas. Additionally, range-change kits were installed in all count rate meters and gamma system recorders, and portable rupture confirmation instrumentation was provided. However, by late 1963, serious problems were evident in the new equipment. Several sample valve probe seals failed, the rupture confirmation instrumentation sometimes gave spurious readings. Most importantly, the flow regulators malfunctioned. HW operators were able to repair and improve the first of the two problems, but new flow regulators had to be procured and installed in late 1964.
Additional safety and instrumentation improvements installed in the 105-B Building in the latter half of the 1960's included resistance thermal detectors (RTDs) that were emplaced on the rear face piping in 1965 and 1966 to measure the rate at which effluent tube temperatures changed. This information was translated by HW operators into an indication of power rate-of-rise, and was considered to be an improvement over "stop-gap" rate of rise instrumentation installed in 1958 and 1960 (see Section 4.10). Also, new and extended automatic sprinkler systems for fire protection were installed in B Reactor's change and storage rooms (for SWP clothing), the lunch and locker rooms, blue coverall storage room, maintenance shop, miscellaneous storage areas, and in the 1713-B Storage Building. The new equipment included audible alarms, annunciation in the control room, and automatic transmission to the Central Fire Station. During 1966 and 1967, high resistance neutral grounding and ground detection equipment was installed in the existing 2,400-volt power systems in the 100-B Area. A three-phase grounding transformer was emplaced to supply ground current, along with a secondary circuit to limit fault current valves. Also, a current pulse generator and ground detection instrument for the location of ground faults and an annunciator alarm were installed. The purpose of the new equipment was to provide sufficient ground current to suppress transient over-voltage, and to supply a means of locating ground faults without shutting down the power system. Two other instrument upgrades planned for B Reactor in the mid-1960's were canceled, due to the impending closures of the HW reactors announced by President Lyndon Johnson in 1964. Intermediate range, rate-of-rise instrumentation, intended to provide automatic shutdown "without dependence on procedural adjustments or interpretations," was proposed in 1962. The proposal was revised and again requested in 1964 and early 1965, and finally refused by the AEC in April 1965. Likewise, work on a front face work area fog spray system at B-Pile, begun in 1966 as the completion of the earlier Project CGI-791 Reactor Confinement Project, was stopped in 1967 due to the imminent closure of the reactor. Only the initial carbon steel piping was emplaced.
After the Presidential announcement that Hanford's reactors would be shut down in a phased sequence beginning in December 1964, it became harder and harder to gain approval for any improvement projects. Yet, maintenance and upgrades continued to be necessary at the old facility. A January 1966 Maintenance Work Forecast for the 100-B/C Area illustrated the growing list of needs. At B Reactor alone, it was estimated that, in the coming year, work would be needed on the Ball-3X system, the diversion box and effluent basin, the HCR and VSR systems, the inlet water coolant lines, process tube caps and orifices, the leaking gas system in the 115-B Building (including helium compressor, relief valves, nearly all of the piping, and the gas dryer system), the "D" platform, the water treatment and filtration systems in the 183-B Building, and the underground water lines from the 183-B facility to the 190-B facility to the reactor itself. This last set of pipes was described as "particularly susceptible to breakage...[and] deteriorated badly." Additional upcoming project needs of B-Pile were listed as replacement or repair of the bottom omega seal in the "D" area, the front face pressure monitoring sensing lines, the rupture monitoring system, an estimated 175 process tubes, and nearly all of the rear nozzles. Also, maintenance repairs were needed to the reactor confinement systems (including the rear face fog spray and the equipment in the 117-B and 119-B Buildings), and to several 1700-B series shops, garage, and storage facilities (including new roofs, floors, and other building parts). All of these repairs, replacements, and upgrades, according to the B-Area maintenance supervisors, would consume 23,606 man-hours. This total figure excluded the major diversion box and effluent basin repairs, because they would be handled by the site construction contractor and not by regular 100-B Area personnel. Furthermore, a fresh metal handling conveyor system had been recommended to improve the efficiency of the charging process. If installed, this project would add 300 man-hours to the maintenance staffing needs of the reactor.
The AEC's reluctance to support extensive upgrades in the aging HW reactors, combined with the wide-ranging needs of the 100-B Area, often produced operating conditions that were far from ideal during the last years of B-Pile. Many planned projects were canceled or never completed, including one to add new air brakes, compressors, shafts, lighting, bumpers, railings, and controls to the "C" and "D" elevators. In 1964, new emergency brake systems had been installed on the platform hoisting shafts. However, additional, anticipated upgrades not completed were to have addressed safety issues that had arisen as tooling modifications, added to the elevators over the previous few years, had increased the weights of the platforms beyond design capacity. Among the few projects that were funded in the late 1960's was the addition of two 600-HP pumps and motors, with associated piping and valving, in the 181-B Building. These pumps, excessed from the shut down H-Reactor, provided extra raw water pumping capacity to the 100-B/C Area beginning in late 1966.
The following year, repairs were undertaken to the concrete walls and floor of the 107-B Retention Basin, which was still being used as the diversion vessel for unusual effluent from B-Reactor. According to an HW report, the basin had "deteriorated to the extent that leakage...creates a structural as well as contamination problem...Several cracks...have widened to about 1/2 inch, with displacement of the floor slab adjacent to the cracks. In addition...there are two holes about 8 inches in diameter...[and] additional holes may exist...The leakage...[has] created channels and void spaces under the floor slab...During reactor operation, a lake of hot effluent forms along the north and west side." Although complete repair of the basin was termed "not feasible," the major cracks and holes were grouted with a mixture of cement and clay. Other portions of the basin were cleaned of debris, sandblasted, then caulked with adhesive EC 1293 B/A,* and sprayed with an additional sealant. Also in 1967, new security fencing and fence lighting was emplaced around the 100-B/C Area, thus consolidating and reducing the sizes of the 100-B and 100-C exclusion areas to about three-fifths of the previous area. Additionally, a new patrol building (1720-B Structure) was constructed between the 105-B and 1704-B buildings, and new asphalt concrete roads, parking areas, and walkways, along with one new gravel-surfaced road and turnaround area, were built. That same year, another improvement was made to the last ditch cooling system for the 100-B/C Area. The turbine drive of one of the two 7,500-gpm, 150-foot head steam pumps in the 181-B/C Building was replaced with a diesel engine drive excessed from the 100-D Area. The new engine had a horsepower rating of at least 20 percent above that required by the pump, and its fuel was supplied from the existing, aboveground oil storage tank located outside of 181-B.
Almost as soon as the dramatic power level augmentations began at B Reactor after Project CG-558, the Health Instruments Division and the production staff at HW became increasingly concerned with the effects of pile effluent in the Columbia River. Throughout 1957, 1958, and 1959, virtually every aspect of the bioaquatic and potential downstream health consequences of reactor effluent were examined, including the effects of temperature, operating purges, various purge agents and filtration aids, fuel element ruptures, sodium dichromate, and the radionuclides themselves. In 1957, the aquatic biology staff reported that, at an overall, in-river concentration level of 0.04 ppm, the key corrosion inhibitor sodium dichromate could cause "significant retardation in growth and a measurable increase in mortality...[in] important species of fish" such as salmon and trout. The chemical also could be "toxic to irrigated crops."
At the same time, various purging agents and frequencies were scrutinized, as the greater volumes of coolant flow caused increased film buildup on the process tubes. By 1960, each Hanford reactor was being purged an average of two to three times a month, and the standard diatomaceous earth slurry Super-Cel, especially when used during reactor operations, brought forth copious amounts of P-32, As-76, and Zn-65 in particulate form. In 1957, Site scientists had concluded that these particulates did not undergo ion exchange with the riverbed. In fact, they settled in river silts in only the most transitory way, and were available for ready recirculation in water between Hanford and nearby, downstream cities during periods of high river turbulence. Another purge agent, the chemical Turco 4306-B,* began to be used in 1957 to decontaminate reactor rear-face piping prior to maintenance operations, in order to reduce personnel exposure. However, almost immediately, tests demonstrated that the use of this compound resulted in the release of 20 times the strontium isotopes as were released in normal effluent water, and in the augmented discharge of other radionuclides of biomedical concern, including Fe-56, Zn-65, and Np-239. Fish, these tests revealed, "appeared to detect even low concentrations and [to] make an effort to avoid the reagent...[Moreover] the potential of polluting downstream reactor drinking water with radioactive materials is greater than any of the other disposal consequences." Still, the exposure reduction value of the chemical decontamination of rear-face piping was considered to be so great that facilities for such decontamination were installed at each HW reactor for the major tube replacement programs that took place in 1962 and 1963.
During the same period, temperature effects of the increasing volumes of reactor effluent in the Columbia River were evaluated. "Valuable species of Columbia River fish, and especially the fall run of Chinook salmon, are definitely vulnerable to further temperature increases," wrote the site's chief aquatic biologist in early 1958. At the same time, unpublished HW laboratory data demonstrated that temperature increases of only 2 to 3°C above normal "significantly increased the mortality of both the eggs and young of whitefish. By November 1960, river temperatures at the old Hanford townsite (just downstream from F-Reactor) were, in fact, measured at 2°C higher than the water temperatures upstream of B Reactor.
The years of the late 1950's and early 1960's also witnessed dramatic augmentations in the amounts of radionuclides released to the Columbia as a result of pile operations. Chief among the isotopes of concern, according to site scientists, were P-32, due to its "extreme concentration in aquatic organisms and white fish," As-76, which was thought to contribute "approximately 50% of the exposure to the G.I. tract at Pasco;" Np-239, another nuclide that delivered its dose directly from drinking river water; Sr-89/90, due to its bone and gastrointestinal tract effects; and Zn-65, due to accumulations in bones and in shellfish at the mouth of the Columbia River. Chromium-51 was the nuclide "released in greatest quantity" downstream of the reactors, but its contribution to dose in living creatures was thought to be small. Sodium-24 (Na-24), nitrogen-16 (N-16), and manganese-56 (Mn-56) were released in even greater quantities, but decayed in such short half-lives that they almost did not factor into dose calculations. The P-32 was an activation product of sulfur that formed from the sulfuric acid used in process water treatments, Zn-65 formed from neutron activation of the corrosion products of the aluminum process tubes and fuel element jackets, Cr-51 was formed from the sodium dichromate added to process water, and Na-24 was an activation product of aluminum. Most of the other radionuclides in reactor effluent formed directly from parent elements in Columbia River water. By 1960, the total volume flow from the HW reactors had increased approximately ten-fold over that of the World War II period, shortening the practical retention time to only about 30 minutes and making diversion of unusual effluents to cribs or other holding areas virtually impossible. Furthermore, the total amount of radioactivity reaching the Columbia River stood at nearly 14,000 Ci per day!
During 1957, the downstream HW reactors began to detect higher and higher concentrations of radioactivity in their raw water intakes. The activity accumulated in particulate solids in the 183 Building filters, the settling basins, at the riverbank around the 181 Buildings, and in corrosion product in the raw water lines. Slug ruptures increased the total radioactivity levels both at these points, and in river water and on sanitary supply intakes at Pasco, Washington, a point considered to be 24 hours of water travel time downstream from the reactors. A 1959 HW study estimated that fuel element ruptures contributed 20 percent of the total strontium 89/90 (Sr-89/90) content of the Columbia's water at Pasco, and 4 percent of the gross fission product activity there. Furthermore, this study calculated, the annual amount of curies released to the river by slug ruptures had increased from about 16,500 in 1954 to 45,000 in 1958. A 1959, H.I. Division study of the waste disposal criteria for further increases in reactor power levels stated that the exposure limits for P-32 were being exceeded even then by a small percentage of "successful hunters and fishermen," and that "provision to reduce the output of...[P-32] must be included in any [pile] expansion program." Furthermore, the study continued, "provision to reduce the output of other radioisotopes will be required for most cases...[and] it may be necessary to reduce sodium dichromate concentrations." The deleterious effects of temperature increases on Columbia River fish also were emphasized.
As the worrisome findings concerning the effects of reactor effluent in the Columbia River mounted in the late 1950's, various solutions were proposed and tested. Salient among these was the concept of passing reactor effluent through beds of various metals, metal oxides, and ion exchange resins, in order to entrap various radionuclides. A small scale, 1958 feasibility study found that steel wool and several anion exchange resins worked best, and that aluminum, steel, and iron minerals removed up to 90 percent of the P-32 and As-76. Other mineral oxides such as magnetite, limonite, sodalite, geothite, fluorite, and bauxite removed from 50-90 percent of these isotopes, and copper, red brass, zinc, and calcium carbonate retained only about 40 percent. This small study, however, suggested that large scale reactor effluent decontamination of at least P-32 and As-76 was "feasible...[and] possible." In 1959, further tests were undertaken. As a result of these latter tests, iron minerals were rejected due to their decreasing removal efficiency over time, and ion exchange resins were dismissed due to their high costs. Steel wool, even though it worked well, was rejected because it was decided that a study of the equally successful aluminum would include the side benefit of gaining a better understanding of the properties of aluminum components within the reactors. A laboratory-sized bed of aluminum cuttings and turnings was found to have desirable removal rates for several key radionuclides, until corrosion levels in the metal became overwhelming. Ironically however, the removal mechanism was believed to be a reaction between the ions in the effluent and fresh aluminum oxide corrosion product. In other words, the corrosion process itself served to bind up the ions, but the mechanism broke down when too much corrosion became emplaced.
Still, the laboratory size tests seemed so promising that a pilot-scale test bed, 20 feet long, 3 feet wide, and 6 feet deep, was installed in 1960 in a steel-plated tank supported over four concrete slabs near the 107-D retention basin. The bed was filled with 0.020-inch shavings of aluminum, and part of D-Reactor's effluent was diverted through it. After eight months of operations, so much algae had collected in the front end of the bed that the flow rate had to be decreased from 6 feet per minute (fpm) to 2.7 fpm. Such a low flow rate called into question the practicality of the system for the HW reactors, each of which produced 80,000-115,000 gallons of effluent per minute. Additionally, radiation built up on the first few feet of the tank to the extent that it was realized that a full-scale facility would need complete shielding equipment. Most importantly however, corrosion was measured at 0.3 mils per month at the front end of the bed, a rate that would reduce the "effective bed length" progressively to the point that "significant" decrease in efficiency, requiring "virtually total replacement" of the apparatus would be needed after just two and one-half years. As a result of these shortcomings, the idea of decontamination of reactor effluent via aluminum test beds was effectively abandoned in 1961.
The other concept that was explored most thoroughly at HW in the early 1960's for reducing the amount of reactor radionuclides discharged to the Columbia River was that of varying the influent water treatments. In 1963, coolant pH was reduced to 6.6, and facilities were installed at the 183-B Building to add the sulfuric acid for pH adjustment downstream of the filters at the bauxite conversion box. The latter step was taken to prevent corrosion of the carbon steel water plant facilities. However, the most crucial goal of new water treatment experiments was to reduce or remove parent elements from the process water so that radioisotopes could not form from them. Some of the early tests were relatively simple, involving just variations in the chemicals used as flocculation and coagulation aids. In one 1960 test at B Reactor, a 90 percent ionic reduction in the elemental arsenic and phosphorus content of filtered water was achieved when aluminum nitrate at 25 ppm was used as the flocculent in the filtration process. Because this treatment was very expensive, it was considered for possible use only during the August to October "critical season" for P-32, the time when low river flows caused nearly a quadrupling of the average P-32 levels in whitefish taken from the Columbia River between the reactors and the Tri-Cities. However, by 1961, a full-pile, "high alum feed" experiment was underway at the 100-F Area. In this test, flocculation of raw water in the 183-F Building was achieved using 18 ppm of aluminum sulfate, a substantial increase from the normal 3 to 9 ppm aluminum sulfate feed used at most of the HW reactors. Very quickly, a reduction in the P-32 content of reactor effluent to 40 percent of pre-test levels was attained, but ledge corrosion attack on process tubes showed a "marked increase." Despite the worrisome corrosion, the reduction in P-32 levels was so satisfying that the high alum feed program was adopted at all of the HW reactors in 1962. It continued through 1966, but was reduced to approximately 12 ppm at B and C Reactors in 1967, when it was learned that the latter alum feed level achieved a better (i.e., more neutral) balance between the negatively and positively charged ions in raw river water.
The increased build-up of process tube films as a result of the high alum feed program became the subject of intense study at HW in the mid-1960's. Augmented amounts of film were considered highly undesirable because they entrapped parent elements that then became activated by neutron bombardments. Also, the heavier films increased the need for purging, with all of the attendant problems of elevated chemical and particulate releases to the river. Additionally, film buildup quickened the corrosion rate in process tubes, thus elevating the tube failure rate and increasing the need for sodium dichromate. Retubing in the graphite channels of the aging HW reactors was becoming an increasingly risky operation, because it usually involved some degree of graphite loss or damage, and thus further weakened the mechanical strength of the graphite cores and shortened reactor life.
It was learned in early 1963 that the primary corrosion product constituents in the films that accumulated during the high alum feed program were pseudo boehmite (a hydrous aluminum oxide with two to three waters of hydration), boehmite (an aluminum oxide with one water of hydration), and bayerite (an aluminum oxide with three waters of hydration). The primary radionuclides in the films were P-32, Fe-59, Zn-65, cobalt 60 (Co-60), chromium 51 (Cr-51), and scandium 46 (Sc-46). Many new protective coatings and process water additives were tested, in an attempt to reduce the adsorption of parent elements onto the tube surface. The organic coatings that were tried all decomposed when they were subjected to neutron flux within the reactors. Anodic coatings and electropolishing of the tubes actually increased film buildup, because they augmented the absorbency of tube surfaces. Some silicon resins and some ordinary, commercial inks were found to be effective as tube coatings, but cost and various in-reactor difficulties prevented their large scale use. It also was learned that 20 to 100 ppm of sodium silicate addition to the process water reduced P-32 and As-76 concentrations in effluent by factors of two to three, a more favorable decrease than that achieved by any commercial corrosion inhibitor. However, such levels of sodium silicate added an undesirable chemical load to the Columbia River.
As the study of reactor films progressed into the late 1960's, a key concept emerged in understanding the attachment and removal of parent elements in the films. Zeta potential was a measurement of free ions in water, or the elemental ions that existed in raw river water to become corroded and activated by neutron flux. The Columbia's water had a naturally existing negative zeta potential of about 20 millivolts (mv). This meant that a slight excess of negatively charged ions existed in the water, ready to attach to metallic surfaces within the reactors and engage in electrolytic corrosion. Aluminum sulfate or alum had a positive zeta potential, or a slight excess of positively charged ions. Its function, at normal alum feed rates of from three to nine ppm, was to bring the natural colloidal (i.e., suspended particle) state of the river's water to neutrality, thus avoiding a state where free ions of either positive or negative charge would be available to engage in a corrosion reaction. However, the high alum feed program, it was learned in 1967, actually had overcompensated for the natural, negative zeta potential of the Columbia's water, and had freed an excess of positive ions to facilitate corrosion. The corrosion situation became even worse as the reactors aged. The growing realizations about the effects of high alum feed, and the desire to return the water to a state of electroneutrality, brought about the end of the program.
During the same years the HW scientists were discovering the fundamental causes of corrosion and film buildup in the old piles, they made a major discovery that would change future reactor operations forever. In 1962, they built the Water Treatment Pilot Plant (WPP - sometimes also called the Micro Pilot Plant - MPP) in the 1706-KE Building. The major, initial purpose of this plant was to experiment with cooling mocked up and actual in-reactor process tubes with deionized water. Ion exchange columns were installed to remove the minerals and mineral salts from the Columbia's raw water. These minerals served as parent elements that became lodged in reactor films and activated by neutron flux. By 1964, the WPP tests had found that the use of deionized water, in combination with Zr process tubes, reduced the concentrations of P-32 and other radionuclides of concern in reactor effluent to minuscule levels. Installation of large scale deionization plants at the HW reactors, however, was not considered economically feasible at a time when the shutdowns and production phase-outs were underway.
As HW operators searched intensively during the early 1960's for ways to reduce radionuclide releases to the Columbia River, they revived the mid-1950's idea of disposing reactor effluent through inland lakes, or directly to the river through trenches (see Section 3.8). A 1959 HW study of the consequences of direct river disposal of effluent after a "flow through" or settling time in a trench concluded that there were no "overriding considerations" that would make this concept "unfeasible." The longer-lived radioisotopes of concern in effluent, the report noted, would not be affected by eliminating the one-hour (or less) hold-up time in the 107 Retention Basins. However, the study emphasized that the public relations value of the 107 basins, as a "symbol of the care used in disposal of wastes...in the public mind," was very important. Mapping of the soil profile beneath 100-B/C Area began in late 1962, to determine the suitability for effluent disposal. Unfortunately, gravelly sediments, intermixed with coarse sand and loamy sand having poor adsorption capacities, were found. In late 1964, a 600-foot trench to test the direct disposal idea was placed into service at 100-F Area. However, a pronounced mound soon developed in the shallow groundwater table beneath the trench, and springs developed along the north-south perimeter road, fed by seepage from the 107-F Basin. This seepage was being diverted by the groundwater mound under the trench.
After F-Reactor shut down in 1965, an ambitious concept was developed for a canal diversion system that would route 100-B/C effluent through a trench to the 100-KE/KW Area, and then would send the combined effluent of these four reactors through a larger ditch to a river discharge point just downstream of 100-F Area. The total length of the canal would allow many hours of radionuclide decay time, but would contaminate groundwater all along its route and would form "polluted...swamp areas" in the sluices surrounding the discharge point. Apparently this idea was not pursued beyond the design stage, because a trench with a completely different orientation was placed into service in the 100-B Area on October 30, 1967. This trench was 500-foot long, 40 feet wide at the bottom and 200 feet wide at the top, and was tied into the 1904-B outfall line that led to the Columbia River. Within two weeks, the effluent flow rate from B Reactor to this trench was increased from 5,000 gpm to 50,000 gpm. By early 1968, according to HW scientists, an "increase in the level of the water table in the vicinity of B Area is [was] apparent...[and] extensive new seepage areas...formed along the riverbank." The shutdown of B Reactor that February ended the trenching test.
As the reactor shutdowns began at Hanford in the mid-1960's, operators and scientists struggled to extend the viability of the remaining piles by developing environmentally acceptable means of effluent disposal. The need for disposal alternatives in which the Columbia River was not the sole or immediate recipient of reactor wastes became especially obvious after a strike idled the five operating, single pass piles at HW in the summer of 1966. Site monitoring scientists, regarding the strike as a valuable and "unusual opportunity" to study the effects of shutdown on radioactivity levels in the river and in aquatic life, took an abundance of special samples. They found that, "within a few days after the shutdown, the concentrations of reactor-produced radionuclides in the Columbia River dropped to very low levels." Exceptions included Zn-65, Mn-54, Sc-46, and Co-60, isotopes that the scientists surmised were "retained in the bed of the river...[and] recycled to the water through continued scouring and leaching of the sediments." In aquatic invertebrates, concentrations of radionuclides fell by one-to-two orders of magnitude, and in whitefish the concentration levels decreased by 75 percent.
The following spring, with five single-pass reactors again operating, a Hanford summary report on alternate methods of reactor effluent treatment and disposal listed many options in addition to those already discussed. Conversion to recirculating cooling systems was listed as economically prohibitive, since it would involve providing 400,000 gpm of additional cooling (pumping) capacity per reactor, with all attendant piping modifications. A water deionizer and four large heat exchangers also would be needed for each reactor, for a total conversion cost of $32 million per reactor. Other expensive potential solutions, all of which also posed awkward siting problems between the reactors and the Columbia River, included the construction of evaporative cooling towers, spray ponds, and/or air heat exchangers. Other, less expensive proposals each came with physical or acceptability barriers. These included a plan to use the export water systems from deactivated reactor areas to transport effluent from the operating reactors to various disposal and dispersal points, thus effecting delay and decay time. However, concrete cylinder (i.e., steel cylinder with concrete on the inside and the outside) export pipes with rubber gaskets and mortared joints were deemed "entirely inadequate" to hold hot reactor effluent. Another idea was dispose effluent to inland Hanford soils and then to pump it out of groundwater mounds to irrigate non-food crops such as animal feed, mint, trees, cotton, and flax. Another idea involved piping and discharge of effluent to the Yakima River at Kiona, and another advocated the construction of 4,000-foot log booms into the Columbia at the 100-B/C and 100-KE/KW Areas in order to channel the effluent and effect some short delay times. The last concept had almost no purpose except to bring in-river concentrations of the very short-lived Mn-56 into compliance with lowered permissible federal levels.
During the post-Project CG-558 period of B Reactor operations, only a few serious radiation incidents occurred. On February 26, 1959, the "C" elevator was moved while the charging machine still was attached to a process tube. The tube's front nozzle broke, causing irradiated dummies and poison elements reading 10 R/hr to be flushed from the tube onto the "C" elevator. Since personnel previously had evacuated the area, individual exposure was low, but a radiation field of 600 mr/hr existed in the front face area. As a result of this incident, a photocell target was installed on the charging machine, making it impossible to adjust the machine outside of a beam from the PCCF. Also, the key-lock control system on the "C" machine, designed to prevent elevator movement, was upgraded. On September 10 and 12, 1959, B Reactor flow monitor trips scrammed the pile when small pieces of neoprene lodged in the orifice screen in the tube coolant supply line. Subsequent investigation showed that the same type of neoprene was found in all near-side crossheader and basket screens in the valve pit. The origin of the material was found to be a 35-square foot seal gasket in one of the 1,900,000-gallon tanks in the 190-B Building. The gasket had been scheduled for replacement, but the repair had been delayed to the point where deterioration set in. In February 1961, two employee errors caused off-normal events at B Reactor. On the 2nd day of the month, an operator in training mistakenly withdrew an HCR, causing an immediate reactivity surge of 70 MW in the area near the HCR channel. The event was noticed by a regular pile operator, who took corrective action and avoided a total scram. Later in the month, an employee entered an air duct while the reactor was operating, thus encountering a high radiation field. The employee received a dose of 170 mr, and work control procedures were revised.
On April 6, 1962, power to all of Hanford's 100 Areas was interrupted when instrumentation in the Midway Substation detected a double phase-to-ground fault and tripped the breakers. The secondary cooling systems worked well during the 2.5-minute outage, and no reactor damage resulted. On April 17, 1963, B Reactor was shut down to remove a ruptured fuel element that was stuck badly in a process tube. After several unsuccessful attempts at removal, HW operators decided to displace a portion of the gunbarrel, split the tube, and remove it with some of the irradiated fuel charges remaining upstream of the ruptured charge. When the dose rate in the work area reached 3 R/hr, the crew was signaled to retreat, but one worker did not notice the signal and continued working. As he attempted to remove a tool, the full length of the gunbarrel slid out of the shield and the high radiation alarm sounded. The gunbarrel dropped into the fuel storage basin, and the worker retreated. However, he received a dose of four rems (radiation equivalent man units - a measurement of radiation dose that gives the same biological effect as one roentgen of X rays).
Other contamination events in the 100-B Area during the last 11 years of reactor operation usually resulted from the spread of dried residues from retention basins that were undergoing repair, or from wastes that were dropped enroute to, or near, burial grounds and wash pads. Such incidents occurred in April 1957, June 1957, December 1957, January 1958, January 1959, September 1959 (two incidents), April 1961, March 1962, and October 1966.
On January 29, 1968, the U.S. Atomic Energy Commission issued a shutdown order for B Reactor, to take effect on February 12 provided that there were no serious water leaks prior to that time. If such a leak into the graphite occurred before that time, shutdown was to occur immediately after the stack underwent sufficient drying. Since no water leak occurred, final shutdown took place on February 12. As of that date, the reactor was reclassified from Plant and Equipment In-Service to Plant and Equipment for Future Use. Immediately, several key determinations were made. It was decided to keep the irradiated metal storage basin in service, with all ancillary electrical, water, monitoring, and other support services, in order to store existing lags from B-Pile, as well as future lags from C, KE, and KW operations. Likewise, the 107-B effluent basin, the 105-B to 107-B lines, and the 1904 outfall pipe, were to remain in active status, in order to dispose of the constant current of water that sustained the fuel storage basin in a safe condition, and to provide an emergency backup disposal facility for C-Reactor. Additionally, those portions of the 115-B, 181-B, 182-B, and 184-B Buildings that served C-Reactor and/or the 200 Area export water system needed to be kept in operation, along with their various support services and the 115-B Building stack sampling equipment. In order to assure an adequate export water supply to the 200 Areas, as well as emergency backup coolant supply for C-Reactor, an additional diesel drive was installed on one of the raw water pumps in the 181-B Building, and one larger, diesel-driven export pump excessed from the 182-D Building was installed in the 182-B Building. None of the four boilers in the 184-B Building was shut down, but the output of those boilers that had served B Reactor was transferred to supply steam for secondary coolant supply systems for 105-C. Additionally, steam service through the 190-B facility was preserved, in order to provide a second steam loop to the 190-C Building.
The initial deactivation operations at B Reactor, performed within the first few weeks of the shutdown order, consisted of the following procedures
At the close of the most intensive period of B Reactor deactivation work, in the spring of 1968, the AEC inspected the production unit and its ancillary facilities. The agency reported that the shutdown procedures and their execution had been satisfactory and "smooth," despite higher than usual personnel exposure rates during the final discharge operations. At that time, the operation of the irradiated fuel storage basin was planned for an "indefinite" period. No further deactivation activities took place in the 100-B Area until the shutdown of C-Reactor in April 1969 (see Section 6.6).
Some additional deactivation work in the 100-B Area was completed during 1969-1970, in connection with the shutdown of C-Reactor (see Section 6.6). After that time, no substantial decontamination and decommissioning (D&D) activities were carried out until 1974. That year, several Quonset huts from the 100-B Area, as well as the 119-B Sample Building, were moved to 100-N Area. Additionally, the 100-B Area burning pit screen was demolished, and the pit itself was filled and leveled. In 1975, the 1736-B Building was moved to the site of the Fast Flux Test Facility (FFTF), and the 1704-B Building was moved to the 200 Area. Also, oil was removed from pump reservoirs in the 190-B and 105-B Buildings. In 1977, the 107-B Retention Basin was graded along the outside walls and the contaminated soil was covered with four feet of clean earth. Clean fill to a depth of 18-20 inches also was added to the inside of the basin, in order to stabilize contamination. Two years later, the vertical walls of the 1904-B Outfall Overflow Flume were broken down, and the walls and bottom of the flume were covered with earth. At the same time, all electrical and underground water services were removed from the 184-B Power House. Also in 1979, several surplus buildings and associated equipment were sold as excess, and removed by the salvage operator. Among these facilities were the 187-B High Tanks, the 190-B Annex, the 190-B Tank Room and four large tanks, the 1902-B Power House Water Tank, the 1707-B Building and Annex, the 1715-B Building, the 1716-B Building, and the 1719-B Building.
In 1983, the two smoke stacks serving the 184-B Power House were demolished and buried in place, and the ventilation stack for the 108-B Building was demolished and buried in a trench that had been excavated at the base. The 184-B Power House itself then was dismantled and removed by a salvage contractor, but the pad and foundation were not demolished and buried until 1988. In 1984, the contaminated equipment in the 111-B Test Building was removed, packaged, and buried in 200 Area burial grounds. The building itself then was decontaminated, dismantled, and disposed as clean waste. The floor, foundation, and concrete waste tanks were left in place at the site. Additionally that year, the two-year site preparation process for the removal of the 108-B Building was begun. Asbestos, mercury, radioactive, and hazardous waste were retrieved and disposed. Next, equipment was taken out and clean waste was buried in the 184-B coal pit. In 1985, the structure itself was decontaminated and demolished. Also in 1984-1985, a two-year D&D process was carried out to place the 105-B fuel storage basin in a stable mode. The 8-10-foot heel of contaminated water, fuel buckets, and other miscellaneous radioactive materials were removed. The solid materials were buried and the residual water was processed and released according to criteria of that period. Next, the contaminated sludge was removed and stored under protective conditions in the transfer area. Finally, a fixative was applied to the contamination remaining on the basin surfaces.
In 1985, the D&D process was begun on the 117-B Filter Building. All filters and fixtures were removed and buried, and the inside surfaces of the structure were washed with decontaminating rinses. In 1989, the building itself was demolished and the debris was buried in place. In 1986, equipment and fixtures from the blower-dryer rooms of the 115-B Building were packaged and buried in the 200-W Area burial grounds. In 1989, the structure itself was demolished. The aboveground debris was used as fill for the 184-B coal pit, but the floor, basement, and pipe tunnel walls were buried at the 115-B Building site. During 1987-1988, the long process of demolishing the 183-B Chemical Treatment and Filter Building took place. First, the asbestos was removed. Next, the headhouse and sedimentation basin walls were dismantled and used as fill at the bottom of the sedimentation basins. Complete demolition of the building was completed in 1988, with the debris buried in place. The clearwells were left in place to hold clean decommissioning waste.
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