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DOE OSTI2019

In-can Melter for Fukushima Waste: From Laboratory Scale Tests to the Industrial Design Definition - 19021

The In-Can melter process is being developed and qualified to treat waste arising from decommissioning and dismantling operations. This robust, simple and versatile in situ vitrification process is designed for high or medium level waste and is compact enough to be implemented in a decommissioned cell or close to the waste to be treated. It is developed to treat liquid and solid waste, to produce a small amount of secondary waste and to minimize investment and operating costs. The In-Can vitrification process is also developed with a modular design in order to be adapted regarding to nuclear operators' needs and requirements. The close collaboration between the CEA and Orano constitutes a unique structure able to provide a full and integrated solution, from the scenarios studies and matrix design definitions to the industrial implementation. The evaluation of the In-Can process for Fukushima water treatment secondary waste is a project led by Orano, CEA and ANADEC. This paper presents the first results of this project from the scenarios and glass formulation studies to the feasibility evaluation to Fukushima Daiichi Nuclear Power Station implementation. (authors)

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DOE OSTI2018

The DISTINCTIVE University Consortium: Transforming High Activity Materials Research in the UK - 18212

The DISTINCTIVE consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) is a partnership of 11 research universities, working to decommission and clean up the legacy of civil nuclear energy generation, sponsored by the Engineering and Physical Sciences Research Council (EPSRC). An important component of the research programme is to develop the skills and facilities to address key research questions, requiring the use of highly radioactive materials. Here we highlight the outcomes of four research projects in which access to state of the art research facilities for working with radioactive materials, supported by international partnerships, has advanced research in plutonium storage and immobilisation, and the disposal of spent nuclear fuels. First, we have developed a viable formulation and process for immobilisation of the UK plutonium stockpile, including commissioning of a radiological hot isostatic press and validation studies using plutonium-239. Second, we have further underpinned the interim storage of PuO{sub 2}, through understanding of water and chlorine sorption / desorption, pertinent to the historic practice of containment within, and degradation of PVC bags. Third, we have developed a novel methodology to investigate the in situ radiolysis induced dissolution of actinide oxide thin films, using synchrotron radiation, to underpin the geological disposal of spent nuclear fuel. Fourth we monitored in situ the oxidation and self-ignition mechanism of uranium carbide, a legacy fuel from the Dounreay Prototype Fast Reactor that is waiting for oxidation treatment prior to permanent storage and/or disposal. Through this programme of research, the DISTINCTIVE consortium is rebuilding the UK capability for research with transuranic materials, which, hitherto, was eroded over decades of disinvestment. (authors)

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DOE OSTI2017

Drying Model for Fukushima Fuel Debris - 17489

The FATE{sup TM} code is a well-established software tool for simulation of processing, shipping, and interim storage of damaged spent nuclear fuel (SNF) that was used for process design and safety analysis for the Hanford Spent Nuclear Fuel Program (SNFP). This paper describes new model developments that allow FATE to be used for simulation of drying of damaged LWR fuel and, in particular, for simulation of processing, shipping, and interim storage of Fukushima fuel debris. The model is capable of quantifying important tradeoffs in the debris chunk size, process conditions such as temperature and pressure, container size, etc. to optimize process design. The model is also capable of demonstrating proof-of-dryness and to demonstrate the link between observed quantities such as pressure and the amount of residual water. (authors)

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DOE OSTI2018

Test Bed: Treatability of Hanford Tank Waste - 18501

In an effort to advance the Low Activity Waste Pretreatment System technology readiness level at the Hanford Site, the effectiveness of a critical technology element, sRF ion-exchange resin, was demonstrated in the selective removal of Cs-137 from 11.5 L of composited Hanford supernate tank waste. Prior to ion-exchange, solids filtration was performed on the tank waste composite. Between the time of filtrate collection and filtrate processing on the ion-exchange columns, additional solids had formed. These secondary solids resulted from the combination of supernates high in fluoride content with supernates high in phosphate content and the relatively short blend time. Samples were collected before, during and after the ion-exchange process for characterization of a full suite of constituents (inorganic, organic, and radionuclides) to aid in the classification of the cesium-removed waste for shipping, receiving, treatment, and disposal determinations. Following chemical analysis, the remaining 10.8 L (2.9 gal) of cesium-depleted waste was shipped to an off-site vendor for grout-encapsulation and eventual disposal at the Waste Control Specialists Federal Waste Disposal Facility in Andrews, Texas as low-level waste. (authors)

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DOE OSTI2017

Modular Caustic-Side Solvent Extraction Unit (MCU) Improved Solvent Quality and Processing Results - 17496

The Modular Caustic-Side Solvent Extraction (CSSX) Unit (MCU) is a process for Cs-137 removal from legacy nuclear waste. MCU began hot operations in 2008 with an initial planned operating life of three years and a design life of five years. After eight years of operation, MCU has processed more than 23.8 million liters (6.3 million gallons) of waste. The MCU process uses a solvent developed by Oak Ridge National Laboratory (ORNL) to extract the fission product Cs-137 from liquid salt waste, or salt solution (SS), stored in the Tank Farms at Savannah River Site (SRS) after the material has been treated through the Actinide Removal Process (ARP). Batches of SS are prepared and first fed to the ARP, where it is treated to become clarified SS (CSS), then fed to MCU. Once the Cs-137 has been extracted by the solvent, the CSS becomes decontaminated SS (DSS), which is ultimately processed into a cementitious grout at the Saltstone Production Facility (SPF) and subsequently disposed of at the Saltstone Disposal Facility (SDF). The Cs-137 is then stripped from the solvent into an aqueous strip effluent (SE) stream for incorporation into borosilicate glass at the Defense Waste Processing Facility (DWPF). The solvent used in the MCU process is made of four components: the extractant (extracts Cs-137 from CSS), the modifier (increases solubility of extractant and suppressor in the diluent), the suppressor (suppresses the effects of anions and improves stripping), and the diluent (controls density and viscosity). There have been two solvents utilized during the lifetime of MCU. The first solvent was developed by ORNL and was used to design the plant based on its cesium removal capability. Due to the extended life of MCU, in order to continue to minimize the curies disposed at SDF, a new solvent capable of removing more cesium from the CSS was developed by ORNL and implemented at MCU. The second MCU solvent, Next-Generation Solvent (NGS), uses the same modifier and diluent as the original solvent, but a different extractant and suppressor. Subsequent changes to the overall MCU flowsheet were made for the use of NGS. The success of cesium removal from the CSS is often presented as the cesium decontamination factor (DF) achieved for each batch processed. The DF is the ratio of the concentration of cesium in the CSS to the cesium concentration in the DSS. The BOBCalixC6 based solvent was capable of achieving DFs between 100 and 400 for each DSS batch. The NGS solvent is capable of achieving DFs greater than 40,000 for each DSS batch. The NGS blend solvent is currently meeting the goals of the pure NGS solvent. (authors)

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DOE OSTI2018

Savannah River Site Defense Waste Processing Facility Melt Cell Remote Cleanup - 18037

This paper documents experiences at Savannah River Site (SRS) related to both the modification and use of a commercial remote vehicle to remove years of process-related litter (glass and various objects) from a hot cell floor during a production outage. The outage was necessary to replace a Department of Energy Environmental Management (DOE-EM) remotely-operated glass melter that had vitrified waste for 14 years at the Defense Waste Processing Facility. Both chunks of waste glass and numerous items littered the floor of the cell all associated with the long production run. The presence of litter become problematic as it could eventually impede movement or placement of items necessary to continue facility operation. The cleanup also reduces radiation rates within the hot cell which is beneficial. (authors)

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DOE OSTI2019

Hanford Tank Side Cesium Removal System: Process Description and Performance Metrics - 19613

The Tank Side Cesium Removal (TSCR) System is a modular, quickly-deployable and passively safe system for efficiently removing solids and cesium from alkaline liquid waste raffinate created by spent fuel reprocessing on the Hanford site. It further describes TSCR component design features, system performance metrics and operational sequences used for liquid waste treatment, as well as spent ion exchange column (IXC) replacement and storage. This presentation should be of great interest to parties responsible for treatment and disposal of liquid waste containing high concentrations of cesium. (authors)

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DOE OSTI2016

AK Re-evaluation of Rocky Flats Pond Sludge: Planning for Surprises - 16218

The Advanced Mixed Waste Treatment Project (AMWTP) uses the Sludge Repack Process (SRP) to treat prohibited liquids in retrieved containers of sludge originally stored in the early 1970's. Many of these containers were generated in the Rocky Flats Plant 2. Stage Sludge process. In June, 2015, operators at SRP identified containers of waste with historical IDs beginning with the prefix 742, indicating they were from the 2. stage process, but which were marked with the words 'Pond Sludge' and contained unusual waste such as rolled asphalt and plant material. This event required a reevaluation of the AK, which concluded that the pond sludge containers were from the same generating process as the rest of the 2. stage sludge, and did not change the hazardous waste numbers, waste material parameters or waste matrix code. This case study demonstrated the uncertainty inherent in all AK because of the long time period between generation of the waste and current characterization. Records are missing, unreadable, and difficult to interpret; memories are faulty; and containers continue to surprise us with unexpected waste. AK must be continually reevaluated and updated throughout the life of the project. When disposing of legacy waste, conditions such as that described in this case study should be expected and planned for. (authors)

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DOE OSTI2019

NM in HLW Borosilicate Glass: Focus on the High Performances of the IHMM Operated at La Hague Plant - 19019

Ruthenium, palladium and rhodium, which belong to Platinum-Group Metals (PGM), are part of fission products resulting from nuclear fuel reprocessing. All High Level Waste (HLW) vitrification facilities have experienced operational issues that can be related to the presence of Noble Metals (NM) in the melter. These metals have low solubility and modify glass properties such as viscosity or electrical and thermal conductivity. These dense metals also tend to settle to the bottom of the melter. Whatever the vitrification process operated, the specific properties and behavior of NM have a propensity to impact the production capacity by decreasing the glass throughput and increasing downtimes. Orano and the CEA have studied the behavior of NM in borosilicate glass for many years. It has led to several technological improvements and process operation optimizations on the Induction Heated Metallic Melter (IHMM) currently operated at La Hague plant. These enhancements, implemented since active commissioning, have headed to a high efficiency in terms glass production and NM incorporation (more than 21000 canisters produced with up to 3 % of Noble Metals in the glass). This paper sets out the continuous improvement program implemented on the La Hague vitrification lines that has led to outstanding records of operation in terms of production capacity and NM incorporation. The benefits of the close collaboration between R and D, engineering, and operation teams is particularly emphasized. (authors)

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DOE OSTI2019

Criticality Safety Aspects of Higher Fissile Gram Equivalent at the AMWTF - 19296

Operating since 2003, the Advanced Mixed Waste Treatment Facility (AMWTF) has used high-tech characterization equipment, robotics, and automated treatment processes to accomplish its mission. The facility was designed to treat contact handled waste up to 325 Fissile Gram Equivalent (FGE). As the lower fissile mass drums and boxes were treated and shipped for disposal, other high fissile mass containers needed to be treated and shipped. The facility's mission evolved to include treating containers with up to 3,000 FGE. The nuclear criticality safety group has worked diligently with operations to increase AMWTF's fissile mass limits. As operations change across the Department of Energy facility, the nuclear criticality safety groups must work with operations to meet the mission. How Fluor Idaho LLC is accomplishing this task with respect to criticality safety is the subject of this paper. The signature piece of equipment at AMWTF is the Supercompactor, a giant hydraulic ram capable of exerting 4 million pounds of force that can reduce a 55-gal drum to a 5-inch puck. By compacting waste, the Department of Energy can put more waste into the limited storage space of the Waste Isolation Pilot Plant. Compaction also reduces the number of trips required to transport waste to the Waste Isolation Pilot Plant. There are two box-lines with hydraulic manipulators, nondestructive assay systems for drums and boxes, along with real-time-radiography. The plant's operations are highly automated. As time passed, AMWTF was tasked to treat drums that were not treatable with current operations. Based on analysis and experience of the nuclear criticality safety personnel, we have changed the fissile limits while maintaining a critically safe environment for the facility. Credit was taken for form, geometry, and lack of moderation in the material being treated. As the need arose to process higher FGE containers, more controls were identified for the treatment process. Criticality safety evaluations were expanded to 1,000 FGE for operational flexibility. Those evaluations included the box-line, the hot maintenance cell, and conveyors. Later, criticality safety evaluations were written to allow tanks to be processed in a box-line, as well as the use of a port in the box-line. Criticality safety evaluations for the box-line, hot maintenance cell, and conveyor were expanded to 3,000 FGE. The criticality safety officer is onsite and actively involved in operations. By expanding the nuclear criticality safety operating parameters, the AMWTF will meet the 2008 Record of Decision in which the Department of Energy directed 8,700 cubic meters of waste to be treated at the facility. (author)

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DOE OSTI2020

Panel Session 23: Start-up of Three Major DOE Treatment Facilities - SWPF, IWTU and DFLAW

This panel focused on the DOE-EM planning to complete construction and start up three major treatment facilities in the next 18 months. This included the Salt Waste Processing Facility (SWPF) at Savannah River, the Integrated Waste treatment Unit (IWTU) at Idaho and the Direct Feed Low-Activity Waste (DFLAW) at Hanford. DOE Site Managers from SC, ID and WA introduced each facility then the DOE Contractor provided the details on their biggest challenges in regulatory issues, construction, start-up and operations of each facility. Panelists with presentations: Lessons in Managing Risks in First of a Kind Nuclear Construction and Start-Up (Pamela Marks); Integrated Waste Treatment Unit (Connie Flohr); IWTI-Fluor (Frederick Hughes); Startup and Commissioning (Tom Fletcher)

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DOE OSTI2018

D and D of a Large Waste Treatment Plant in Germany - 18458

With its 40 years of nuclear experience, GNS, the world leading supplier of casks for spent fuel, ILW and HLW, also offers services for conditioning and management of all types of radioactive waste, especially for the German NPP. Since 1985, GNS treats and packages ILW and LLW from the operations and phase-out of German NPP in its Duisburg plant. To enable its permanent safe final disposal, solid wastes are compacted and dried before being shipped to an interim storage facility or - later - the Konrad repository. The total premises cover 20.000 m{sup 2} (215.000 sq ft), of which 9000 m{sup 2} (97.000 sq ft) is controlled area. The Duisburg Plant has three large industrial warehouses (each 25 m x 136 m or 82 ft x 446 ft) with a total capacity of 10.200 m{sup 2} or 110.000 sqft. For the treatment of open contaminated waste, six enclosures were built. These enclosures have a special ventilation system including a specific filtration plant. In one enclosure a high compaction plant was used, in another enclosures a drying facility for drums was in operation. During peak periods, more than 900 tonne of waste were treated annually. One enclosure was mainly used for maintenance, restoration and extension of mobile facilities, which were normally used for campaigns in NPP. Due to the decreasing amount of operational waste during the nuclear phase out in Germany, GNS has decided to close the Duisburg Plant. In the first half-year of 2017, the last waste packages were treated but around 80 20'-containers are still located on site. The D and D started in January 2017. The presentation will show the first experience of decommission processes of the first three enclosures. The following steps including the techniques used will be explained in detail: - Dismantling and removal of the high compaction facility; - Dismantling and removal of the drying facility; - Cleaning and wipe tests of the empty enclosures; - Removal of the stainless steel plates of the floor; - Investigations of the lower floor parts for contamination; - Removal of contaminated underground drainpipes; - Decontamination of the concrete floor; - Dismantling of the ventilation system in three enclosures; - Dismantling of the steel constructions. An important aim during the dismantling is to reduce the amount of nuclear waste and maximize the fraction material for free release. Next to the first three enclosures, in one of the enclosure IV - VI, a 150 tons scrap shear had been installed before the construction of the enclosure building. The scrap shear has to be dismantled and handled by heavy duty lifting equipment. Only afterwards is it possible to start the decommissioning of the last three enclosures. Our goal is to transfer this very large controlled area and the whole plant to the site owner, the Duisburg Port company, after 2 years and 5 months and doing so in a way so that the industrial warehouses can be used again for all possible types of industries. (authors)

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DOE OSTI2020

Assessment of Cold Spraying Mild-Steel Powder onto Mild-Steel Substrate for Development of Waste Tank Repair System - 20515

Infrastructure for nuclear waste storage and processing (e.g., waste tanks, transfer pipelines, evaporators) at the U.S. Department of Energy (DOE) Hanford Site has been in service well beyond its original design life. Past corrosion control methods at Hanford for waste tanks have not been fully successful, as evidenced by leaks forming in tank AY-102, the first double-shell tank (DST) constructed, and significant degradation detected in the primary tanks of AP-102 and AY-101. Failure of infrastructure components can pose a significant risk to the environment, worker safety, and project schedules. Therefore, remotely deployable methods for repairing and extending the service life of the existing tank waste infrastructure are needed. Cold spray has been identified as a potential technique for performing in situ repairs and material buildup for life extension of waste storage and processing infrastructure. Cold-spray repair methods and technology are well-established and are used extensively by industry and the military for making repairs using dissimilar materials. Pacific Northwest National Laboratory is evaluating the process of making repairs using mild-steel powder applied to a mild-steel substrate to minimize the galvanic corrosion potential. This paper presents the initial results for a feasibility ('go/no-go') assessment as to whether low-density deposits with sound metallurgic bonds, to build-back steel plate material, can be achieved using existing commercial cold-spray equipment. An assessment of steel-on-steel deposit is necessary to determine if cold spray is a viable candidate for DST carbon steel plate restoration/repair and thus DST service life extension. Powder properties and preparation requirements have been investigated and operating conditions and parameters for depositing a cold-spray buildup of mild steel on a mild-steel substrate have been determined. Soundness of metallurgic bonding as a function of the operating parameters exercised and results from applying cold spray to lay a deposit over simulated small-diameter flaws in 1/2-inch mild-steel plate are presented. Application of in situ cold-spray repair to the DOE complex has the potential to reduce cost and schedule impacts associated with component failures and the need to procure and construct replacement infrastructure. (authors)

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DOE OSTI2019

Safety Basis Development for the Tank Closure Cesium Removal (TCCR) Demonstration - 19596

The Tank Closure Cesium Removal (TCCR) system has been installed at Savannah River Site (SRS) to supplement the existing salt waste treatment facilities. The TCCR system is intended to demonstrate the capability of ion exchange (IX) technology to remove Cs-137 from high level liquid waste. Separation is performed by Crystalline Silicotitanate (CST) IX media. The media is stored in shielded columns within a small enclosure. Following processing in the TCCR system, decontaminated waste can receive further treatment and be dispositioned at other SRS facilities. The Cesium-laden columns will instead be removed from the enclosure and transported to another location on site for interim storage. Developing the safety basis for the TCCR system and integrating it into the documented safety analysis for existing facilities presented a unique challenge. Most notably, utilizing new methodologies to analyze new hazards resulted in unexpectedly high consequences. The TCCR system therefore required more rigorous controls than had initially been anticipated. However, the integrated control set developed for the TCCR system should have little impact on its operating capabilities and will allow it to begin operating in 2019. This may lay the foundations for future utilization of similar systems at SRS and across the Department of Energy complex. (author)

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DOE OSTI2016

WTP Waste Feed Qualification: Hydrogen Generation Rate Measurement Apparatus Testing Report

The generation rate of hydrogen gas in the Hanford tank waste will be measured during the qualification of the staged tank waste for processing in the Hanford Tank Waste Treatment and Immobilization Plant. Based on a review of past practices in measurement of the hydrogen generation, an apparatus to perform this measurement has been designed and tested for use during waste feed qualification. The hydrogen generation rate measurement apparatus (HGRMA) described in this document utilized a 100 milliliter sample in a continuously-purged, continuously-stirred vessel, with measurement of hydrogen concentration in the vent gas. The vessel and lid had a combined 220 milliliters of headspace. The vent gas system included a small condenser to prevent excessive evaporative losses from the sample during the test, as well as a demister and filter to prevent particle migration from the sample to the gas chromatography system. The gas chromatograph was an on line automated instrument with a large-volume sample-injection system to allow measurement of very low hydrogen concentrations. This instrument automatically sampled the vent gas from the hydrogen generation rate measurement apparatus every five minutes and performed data regression in real time. The fabrication of the hydrogen generation rate measurement apparatus was in accordance with twenty three (23) design requirements documented in the conceptual design package, as well as seven (7) required developmental activities documented in the task plan associated with this work scope. The HGRMA was initially tested for proof of concept with physical simulants, and a remote demonstration of the system was performed in the Savannah River National Laboratory Shielded Cells Mockup Facility. Final verification testing was performed using non-radioactive simulants of the Hanford tank waste. Three different simulants were tested to bound the expected rheological properties expected during waste feed qualification testing. These simulants were tested at different temperatures using purge gas spiked with varying amounts of hydrogen to provide verification that the system could accurately measure the hydrogen in the vent gas at steady state.

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DOE OSTI2020

Infrastructure and DFLAW Support at Hanford - 20443

Manhattan Project era infrastructure systems are degrading at an accelerated pace across the Department of Energy's (DOE) Environmental Management complex. Revitalizing, rejuvenating and right-sizing these systems to ensure reliability for ongoing cleanup missions is a major focus at the 580-square- mile Hanford Site in southeast Washington State. The push to complete construction, commissioning, startup, and operation of facilities and systems in the Direct-Feed Low-Activity Waste (DFLAW) program at Hanford by 2023 requires considerable coordination among site contractors and a significant investment in infrastructure. The passage of time also poses a challenge. It has been about 75 years since facilities were first operated at the site, and approximately 40 years of plutonium production created a legacy of unidentified active and abandoned underground obstacles and waste sites that must be avoided when building and upgrading the infrastructure. Converting infrastructure systems originally built to support plutonium production from the 1940's to the 1980's and upgrading those systems to optimize technology is an ever-changing balance of funding profiles, available resources and execution strategies. The DOE Richland Operations Office (RL) has long recognized the need for increased investment in Hanford Site infrastructure to support the future processing of approximately 56 million gallons of waste currently stored in large underground tanks. When DFLAW is fully operational, safe and reliable infrastructure systems will be needed to ensure continuity of operations around the clock, 365 days a year. These include roadways, water, power, sewer, information technology systems and facilities. To ensure the readiness of Hanford's infrastructure to support treating tank waste in the next three years, 15 projects were identified with a combined value of $133.8 million. Through calendar year 2019, 7 infrastructure projects have been substantially completed and the remaining 8 projects, with a remaining value of $103.8 million, are scheduled to be completed by December 2023. RL and its site services/infrastructure contractor, Mission Support Alliance (MSA), regularly evaluate the needs of the DFLAW program and other Hanford Site missions to ensure the highest priority systems are addressed. In addition to a fast-approaching deadline for round-the-clock treatment operations, RL and MSA face another significant infrastructure challenge. As cleanup is being completed in a 220-square-mile area called the River Corridor, most of the cleanup operations going forward will occur in a 20-square-mile area in the center of the Hanford Site, known as the Central Plateau. Infrastructure systems are becoming more congested in an already overcrowded area. This paper/presentation will outline some of the challenges Hanford faces while executing infrastructure reliability projects. (authors)

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DOE OSTI2018

Direct Feed Low Activity Waste Test Platform - Pretreatment Unit Operations - 18613

The Low-Activity Waste Pretreatment System (LAWPS) facility is planned to pretreat Hanford tank waste supernate by filtering solids and processing through ion exchange columns to remove cesium. Note that cesium ion exchange is not discussed in this paper. The filtration process is achieved through cross flow filtration using a stainless steel Mott filter with a 0.1-micron grade. The goal of this effort was to assess whether a batch, or dead end, filtration (DEF) device replicating the LAWPS cross flow filter could be employed to qualify each feed campaign prior to delivery to LAWPS. The objective for filtration feed qualification is to determine whether the cross flow filters deployed in LAWPS will remove solids while maintaining a de minimis flow rate of permeate. To that end, the DEF was tested with simulants and the results compared with those obtained from cross flow filtration tests. In addition, actual waste samples were retrieved from Hanford Site Tank AP-105 and were composited for testing. These composited samples were first tested with cross flow filtration to determine the filter flux under typical LAWPS conditions. The same material was subsequently tested with DEF to determine a filter flux. While the DEF correctly predicted the cross flow filter's ability to filter solids from the feed, there was limited correlation between DEF and cross flow filtration flux results. This lack of correlation suggests a fundamental difference between cross flow filtration and DEF as used in these tests. Taken in total, these results indicate that DEF is not a good candidate to provide qualification of feed for LAWPS cross flow filtration. (authors)

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DOE OSTI2018

Improved Heat Exchanger Performance Achieved Through Improved Condensate Drainage - 18478

Hanford's 242-A Evaporator has had the mission of reducing waste volume in the Hanford Tank Farms since the late 1970's. At the very heart of the 242-A evaporation process is the E-A-1 reboiler, a heat exchanger which adds heat to the waste, and the vacuum condenser system, a three-stage condenser/ejector system which condenses the waste vapors and provides the vacuum at which the evaporation process operates. Over a long period of time, significant degraded performance has been observed in the heat exchangers in both the heating system and in the vacuum condensing system. This degraded performance was indicative of degraded conditions in equipment and/or piping associated with both the heating and vacuum condenser systems. However, in searching for degraded condition in these systems, 242-A engineering also discovered that both systems contained design errors in the condensate drainage associated with all heat exchangers in both systems. By addressing the condensate design errors in addition to correcting the degraded conditions, system performance was restored and even possibly improved. All heat exchangers in the heating and vacuum condensing systems are shell and tube type heat exchangers. Both have liquid on the tube side and vapor on the shell side (in the case of the reboiler the liquid is tank waste and the vapor is 10 lb heating steam, and in the case of the vacuum condensers the liquid is cooling water and the vapor is waste vapor). In both systems, engineering proposed that the vapor condensate was backing up into the heat exchanger, resulting in added poor performance. By focusing on basic heat exchanger system design, projects were initiated to redesign the condensate drainage of these systems and re-plumb the condensate drain lines. Data collected before and after the replumbing provides evidence that the corrected drain lines from the heat exchangers improved the performance of the system as a whole, and in the case of the reboiler, removed a potentially damaging steam condensate water hammer condition. (author)

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