ML20196J504

Forwards Info to Facilitate Review of ISFSI License Application 239

ML20196J504
Person / Time
Site:	Trojan  File:Portland General Electric icon.png
Issue date:	07/30/1997
From:	Quennoz S PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20196J508	List: ML20196J504 ML20196J513 ML20196J524
References
VPN-054-97, VPN-54-97, NUDOCS 9708040213
Download: ML20196J504 (74)

Text

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Portland General Electric one World Trade Center 121 SW Salmon Street Portland, OR 97204 July 30,1997 VPN-054-97 Trojan Nuclear Plant Docket 50-344,72-017 7 License NPF-1 U. S. Nuclear Regulatory Commission Document Control Desk m

Washington, DC 20555

Dear Sirs:

License Change Application (LCA) 239 - Debris Processing in the Fuel Building TransmittalmfIntegrated Test Report and Assessment of Steam-Fuel Affects This letter transmits the Summary and Data Reports of the Integrated Test for the Steam Reforming Process, the procedure for the segregation of material for the Spent Fuel Pool Debris Project and a separate assessment of any potential steam-fuel and corrosion affects in the process can or the process can capsules during long-term storage in the Independent Spent Fuel Storage Installation (ISFSI). LCA 239 requested NRC authorization for the processing of spent fuel pool debris in the Trojan facility Fuel Building. The LCA was formally issued as a 10 CFR Part 50 License Amendment (Number 198) on June 9,1997. The License Amendment was not intended to, and did not address the long-term storage issues related to the placement of the process can capsules into the 10 CFR Part 72 ISFSI storage casks. This information is being provided to the NRC staff to facilitate the review of the ISFSI license application.

Attachment I to this letter is the Assessment of Steam-Fuel Affects and Enclosure 1 is a copy of the Duratek and Intellergy Corporation literature search on steam-fuel affects that form the basis for Attachment 1. This assessment documents that the materials of construction for the process can capsule are adequate for the long-term storage of the steam reformed spent fuel pool debris. '

The literature search determined that there are no byproducts of the steam reforming process that pose a credible corrosion concern.

Attachment 2 to this letter is a Summary Report of the Integrated Test and Segregation Methodology for the Spent Fuel Pool Debris Project. Enclosure 2 is the Integrated Test Data Report which documents the detailed results of the recently completed testing of the steam reforrning equipment. The Summary Report documents the basis for the administrative and process controls that will be used to provide confidence that each process can capsule will have no more than the established limit of 0.250 moles hydrogen, and that this limit will not be i[

exceeded over the container's anticipated storage life.

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Natural gas. Electricity. Endless possibilities.

July 30,1997 VPN-054-97 Page 2 of 2 Enclosure 3 is the procedure for the segregation of material for the Spent Fuel Pool Debris Project. This procedure establishes the positive administrative controls on the quantity of black filter flange material that may be steam reformed in each process can. The administrative controls

- will ensure that the limitations assumed in the Integrated Test Report are implemented, thereby providing confidence that the hydrogen generating material limit of 0.250 moles per process can capsule will be satisfied.

' If there are any questions related to these documents or the planned long-tenn storage of the process can capsules in the ISFSI, please contact Mr. H. R. Pate of my staff at (503) 556-7480.

Sincerely, fy'%s m Stephen M. Quennoz Trojan Site Executive Attachments Enclosures c: M. T. Masnik, NRC, NRR D. G. Reid, NRC, NMSS R. A. Scarano, NRC Region IV David Stewart-Smith, OOE A. Bless, OOE

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i ATTACHMENT 1 i

i LICENSE CHANGE APPLICATION (LCA) 239  !

! SPENT FUEL POOL DEBRIS PROCESSING IN TIIE FUEL BUILDING  :

(TAC NUMBER 96936)  !

l ASSESSMENT OF STEAM-FUEL AFFECTS l

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! STEAM REFORMING OF TROJAN SPENT FUEL POOL DEBRIS 1

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4 Attachment 1 Assessment of Steam-Fuel Affects Steam Reformine of Troian Spent Fuel Debris Introduction Following shutdown of the Trojan Nuclear Plant (TNP) reactor in November 1992, PGE decided to decommission TNP. PGE submitted a Decommissioning Plan, dated January 26,1995, that detailed prompt decontamination and dismantlement of contaminated structures, systems and components. In addition, PGE's Decommissioning Plan stated that relocating the contents of the spent fuel pool to an Independent Spent Fuel Storage Installation (ISFSI) would be the most economical method for storing the TNP spent fuel until a permanent storage facility offsite would be available for offsite shipment of the spent fuel. Relocating the spent fuel and other high-level radioactive waste from the spent fuel pool to the ISFSI allows for decontamination and 1 dismantlement of structures, systems, and components at TNP to occur sooner than if the spent )

fuel and other high-level radioactive wastes were left in the spent fuel pool until shipment to an offsite facility were possible. By letter dated April 15,1996, the NRC approved the Trojan Nuclear Plant Decommissioning Plan.

The spent fuel pool contains not only the spent fuel assemblies, but also a number of other components and debris that have been placed in the pool for shielding and housekeeping .

purposes. Specifically, the pool contains non-fuel bearing components (such as parts of spent fuel assemblies that do not contain any fuel), fuel debris (whole or part pellets, spent fuel fragments and any portion of a spent fuel assembly that could contain spent fuel material in any quantity),

and debris (material, equipment, segments of components, filter material, clad segments, buckets, etc, other than intact fuel assemblies or fuel assembly skeletons). In order to address this material that is not actual spent fuel assemblies, PGE developed Specification TD-012," Trojan Nuc' ear Plant Spent Fuel Pool Debris Project." After assessing the technical options for the processing of the non-spent fuel assembly materials currently in the spent fuel pool, PGE has selected a thermal treatment technology proposed by Scientific Ecology Group, Inc. (SEG).

The non-spent fuel assembly materials will be segregated (non-fuel bearing components, Low Level Waste, potential gas generating materials), characterized, processed and packaged.

Materials that meet the criteria for low level radioactive waste disposal will be packaged and shipped for disposalin accordance with existing processes and procedures. The potential gas generating materials and fuel debris will be processed through the SEG thermal treatment (steam reformer) to destroy organic materials and remove ' free', ' interstitial' and ' hydrated' water. The removal of the water and destruction of organics prevents radiolysis and assures that the quantity of hydrogen gas generated is less than 0.250 moles in the sealed container (process can capsule) and that this limit is not exceeded over the containers' anticipated storage life.

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The SEG steam reforming process does not require the separatio, of fuel debris from organic material. This permits the batch processing of fuel debris, corrosion products and greater than Class C waste that is intermixed with organic filter material. The final product is free of organic and water bearing materials and eliminates the need to vacuura dry the final storage capsule. The dehydration of organic materials and the removal ofinterstinal hydrated water from fine metallic and clay debris cannot be achieved by traditional fuel canister vacuum drying techniques. The steam reforming process eliminates any concern of flammable gas generation or over pressurization in the final Trojan fuel storage / transport containers.

Initial Materials Assessment Related to Steam-Fuel Affects SEG proposed that the materials of construction for the process can and process can capsule be stainless steel. The basis for this material selection was engineering judgement recognizing the chemical and radiological materials that could be formed and exist (even for comparatively brief periods of the steam reforming process) in the process cans and the suitability of the process can capsule for long-term storage in the ISFSI.

The process can and process can capsule are fabricated from 300 series stainless steels. The process can 5 micron filters are fabricated from sintered inconel. The process cans are suitable for operating temperatures as high as 1400 degrees F. The use of stainless steel for both the process can capsule and the process cans also ensures that no corrosive galvanic interactions of dissimilar metals will be present during long-term storage. The process can capsule is not exposed to the processing extremes of temperature but is placed in the spent fuel pool for storage until loaded into an ISFSI storage cask. The spent fuel pool is filled with a minimum concentration of 2,000 ppm boron (acidic) for criticality control. The 300 series stainless steel process can capsule will not be negatively affected by the boric acid solution in the spent fuel pool.

l The specific reaction kinetics of the steam reforming process are proprietary, but the stainless steel materials of the process can and process can capsule were selected to ensure that corrosive byproducts would not result in compromising the integrity of the process can capsule during long-term storage. The comparatively mild operating temperature (approximately 1,100 degrees F) and the reducing reaction process (as compared to an oxidizing process) will not produce byproducts from the Trojan debris that will compromise the process can capsule. The stainless steel of the process can capsule will also not interact with the internal materials of the ISFSI PWR i basket during long-term storage. The PWR basket, which is also fabricated from 300 series stainless steel, will be filled with helium (an inen gas) during long-term storage.

As part of the evaluation of the proposals for the spent fuel pool debris project, PGE contracted with Battelle, Pacific Nonhwest National Laboratory, to assess the viability of the technical options proposed. The Summary Report (PNWD-2350, dated June 1996) documents the results of the Battelle review and was provided to the NRC stafTby letter VPN-029-97, dated March 31, 2

1997. The report contains a discussion of the potential for volatilization of cesium, ruthenium, technetium, iodine and the potential entrainment of transuranics and other radionuclides. l The report concludes that "available data suggests that a small fraction of cesium remaining in the filter media waste may leave the can feed evaporator by either volatilization or entrainment in the circulating gas." In the detailed discussion ofliquid-fed ceramic melters with simulated high level -

tank waste (for the Hanford Waste Vitrification Plant) the report documents that approximately 76 percent of the cesium aerosol leaving the melter was less than 1 micron in size. The report continues "the conclusion drawn from this information is that the cesium is in vapor form rather than an entrained particulate. This conclusion seems to make sense until reviewing previous literature. At 600 degrees C, there is significant potential for volatilization considering the boiling point of cesium is 671 degrees C. Earlier studies support the idea that volatilization occurs However, it was observed that the cesium actually condenses and forms sub-micron cesium particulate, which traveled through micron filters at ambient temperature using inert gas / steam as a carrier." The report also states "the operating experience of SEG and Synthetica suggests that entrainment should be minimal at the low projected flow rates through the evaporator can and that at 600 degrees C, volatilization of cesium should be minimal."

l A cesium trap, including a 1 micron filter, was designed and included with the steam reformer /can feed evaporator process to capture any volatilized or entrained cesium, as described in the PGE response to NRC Questions 6,7,8,9 and 10, dated December 12,1996 (VPN-074-96). In addition, the process can is constmeted of 300 series stainless steel and has 5 micron metallic filters in both the can bottom and lid, to contain solid residue from the processed media. The potential for entrainment of cesium, ruthenium, transuranics and other radionuclides is, thereby, reduced.

l The Summary Report discussion of the potential for volatilization of ruthenium, technetium and iodine is based on Oak Ridge National Laboratory investigation of the volatility of these elements

l. during the calcination of high-level radioactive nitric acid waste. While calcination is not the same process as steam reforming, the operating temperatures (400, 500, and 725 degrees C) were useful in assessing the volatility of ruthenium, technetium, and iodine. The volatility of RuO.

appeared to react with the heated stainless steel walls of the spray calciner to form RuO, which is relatively nonvolatile. A similar situation would likely exist in the can feed evaporator ifit was at 750 degrees C, since the process cans will usually contain a significant amount of stainless steel dross and the process can is constmeted of 300 series stainless steel. In addition the volatility of RuO. was very low (less than 1 percent). Ruthenium was evaluated and not considered to be volatilized in any significant quantity. Due to the 600 degree C operating temperature of the can feed evaporator, mthenium is not considered to offer a corrosion risk.

The report further documented that technetium volatilization varied from 0.2 percent to 1.4 percent over the temperature range of 250 to 600 degrees C in the stainless steel spray calciner.

Volatilization of technetium was not considered to be a factor in material selection. It should be noted that technetium is not a source isotope for the Trojan fuel, rather it is an activation product

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which was included in timinelt.ing analyses for the activation product source term. Similarly, iodine was not considered in the material selection due to the limited quantity that remains afler i approximately 5 years after operation. Though iodine-129 would be volatilized (57 to 97 )

percent), there is not any significant quantity remaining in the spent fuel pool debris after exposure to the pool environment for approximately 5 years. Depending on the velocity of the process gas j through the can feed evaporator, it is probable that transuranics and other radionuclides will be l entrained in the gas stream and captured either on the 5 micron filter of the process can, the cesium trap (including its I micron filter) or the HEPA filter. The gaseous efiluent from the operation of the can feed evaporator is processed through the steam reformer and routed to the Fuel Building ventilation system (see also PGE Response to NRC Questions 6 and 7, in letter VPN-074-96, dated December 12,1996).

Recent Steam-Fuel Affects Assessments In May 1997, SEG was requested by PGE to perform a more detailed assessment of the potential for any steam-fuel effects, for the purpose of documenting the suitability of the process can and process can capsule materials to contain the products of the steam reforming process for the storage period of the ISFSI. Enclosed is the results of the literature search on steam-fuel effects which was performed by Dr. Terny Galloway, ofIntellergy Corporation, for SEG (now called "Duratek"). The largest area of work reported in the literature involves the steam-enhanced outgassing oflight fission products when there is a fuel element melt-down or reactor failure.

This effect is consistent with what would be expected with steam reforming. There does not l appear to be any corrosion enhancement of either the fuel pellets or the container from the process gases directly.

l l There is, however, an effect of steam-reforming-enhanced outgassing of cesium, tellurium and

iodine as final stable species which would then form Csl and CsTe salts. As stated in the previous l section, iodine is not available in any significant quantity due to the extended period since plant l operation and the exposure of the fuel debris to the environment of the spent fuel pool. If all of l the CsTe and Cs! existed in the gas phase in the process can (rather than being transported out of l the process can by steam flow) at maximum temperature, and uniform corrosion of the process  !

l l can were to occur, only 25 microns or 1 mil would be corroded over the process can interior l surface. This mechanism would not resul tin a long-term storage concern for the materials of the process can capsule or the process can in the ISFSI.

Alternatively, if the CsTe salts were assumed to be concentrated at the wall of a process can (with I l 1/5 of the maximum 375 fuel pellet material placed in a single process can [75 pellets], no removal by the process steam flow, and no mixing within the matrix of material being processed), and the theoretically calculated Csl salts were also assumed to be present, the maximum calculated pit depth in the process can would be 3.6 mm or about 1/8 inch. The theoretical conclusion could then be that, in a worst case scenario, a pit could corrode through the process can wall (process can wall thickness is equal to 0.105 inches) but would n_o1 penetrate the outer process can capsule 4

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(wall thickness equal to 0.250 inches). This pit formation process is a chemical process that depletes the Csl and CsTe salts by reacting with the iron in the process can material. This results <

in the termination of pit formation when the Cs salts are fully reacted. The theoretical formation of cesium salts in the process can during steam reforming will not result in a long-term storage concern for the materials of the process can capsule or the process can in the ISFSI.

There is a bcdy ofliterature revealing the various conversions of uranium (such as U 0,) and3 plutonium oxides to the lower oxidation states, such as UO, and PuO , under steam-reforming 2

conditions, but no corrosion enhanced effects were revealed. Also the SEG steam reforming conditions are sufficiently mild that there is no further reductior, of these oxides to the metallic state. There is no evidence of a steam reforming (reducing) environment converting uranium oxides to higher oxidation states (UO, to U3 0,). The expandan of the oxide fuel material, due to conversion to a higher oxidation state, is not a concern for the T:ojan fuel debris project. It should be noted, however, that the volume reduction of the hydrogen generating waterials in the process can during the steam reforming process results in ample space for any pwntial expansion of the fuel material in the process can. The theoretical conversions of uranium and plutonium in the process can during steam reforming will not result in a long-term storage concern for the i materials of the process can capsule or the process can in the ISFSI. '

l Mixtures of cesium-uranium-oxygen were also examined in the literature search to identify any eutectic-enhanced corrosion. Although the cesium-uranium-oxygen converted to Cs UO. which 2 l was the stable state, no corrosion enhancement was observed. The theoretical conversions of cesium-uranium-oxygen in the process can during steam reforming will not result in a long-term storage concern for the materials of the process can capsule or the process can in the ISFSI.

Conclusion l The materials of construction for the process can capsule are adequate to provide for the long-term storage of the steam reformed spent fuel pool debris. The literature scarch performed originally as part of the initial assessment of the process and the recent more detailed review determined that there are no byproducts of the steam reforming process that pose a credible corrosion concern.

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i ATTACHMENT 2 L i l

l LICENSE CHANGE APPLICATION (LCA) 239 l SPENT FUEL POOL DEBRIS PROCESSING IN THE FUEL BUILDING l (TAC NUMBER 96936)  ;

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SUMMARY

REPORT OF INTEGRATED TEST FOR  ;

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STEAM REFORMING OF TROJAN SPENT FUEL POOL DEBRIS i

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Attachment 2 Summary Report of the Integrated Test Report Spent Fuel Pool Debris Proiect introduction i Following shutdown of the Trojan Nuclear Plant (TNP) reactor in November 1992, PGE decided to decommission TNP. PGE submitted a Decommissioning Plan, dated January 26,1995, that detailed prompt decontamination and dismantlement of contaminated structures, systems and components. In addition, PGE's Decommissioning Plan stated that relocating the contents of tne spent fuel pool to an Independent Spent Fuel Storage Installation (ISFSI) would be the most economical method for storing the TNP spent fuel until a permanent storage facility offsite would be available for offsite shipment of the spent fuel. Relocating the spent fuel and other high-level radioactive waste from the spent fuel pool to the ISFSI allows decontamination and dismantlement of structures, systems, and components at TNP to occur sooner than if the spent fuel and other high-level radioactive wastes were left in the spent fuel pool until shipment to an .

offsite facility were possible. By letter dated April 15,1996, the NRC approved the Trojan )

Nuclear Plant Decommissioning Plan.

The spent fuel pool contains not only the spent fuel assemblies, but also a number of other components and debris that have been placed in the pool for shielding and housekeeping purposes. Specifically, the pool contains non-fuel bearing components (such as parts of spent fuel assemblies that do not contain any fuel), fuel debris (whole or part pellets, spent fuel fragments and any portion of a spent fuel assembly that could contain spent fuel material in any quantity),

and debris (material, equipment, segments of components, filter material, clad segments, buckets, etc, other than intact fuel assemblics or fuel assembly skeletons). In order to address this material that is not actual spent fuel assemblies, PGE developed Specification TD-012, " Trojan Nuclear Plant Spent Fuel Pool Debris Project." Atter assessing the technical options for the processing of the non-spent fuel assembly materials currently in the spent fuel pool, PGE has selected a thermal treatment technology proposed by Scientific Ecology Group, Inc. (SliG), which has since been purchased by GTS Duratek.

The non-spent fuel assembly materials will be segregated (non-fuel bearing components, Low Level Waste, potential gas generating materials), characterized, processed and packaged.  ;

' Materials that meet the criteria for low level radioactive waste disposal will be packaged and  !

shipped for disposal in accordance with existing processes and procedures. The potential gas generating materials and fuel debris will be processed through the SEG thermal treatment (steam j reformer) to destroy organic materials and remove hydrogen as well as ' free', ' interstitial' and

' hydrated' water. The removal of the water and destruction of organics prevents radiolysis and assures that the quantity of hydrogen gas generated is less than 0.250 g-moles in the sealed i 1

container (process can capsule) and that this limit is not exceeded over the containers' anticipated storage life.

The SEG steam reforming process does not require the separation of fuel debris from organic material. This permits the batch processing of fuel debris, corrosion products and greater than Class C waste that is intermixed with organic filter material. The final product is free of hydrogen and water bearing materials and eliminates the need to vacuum dry the final storage capsule. The dehydration of organic materials and the removal ofinterstitial hydrated water from fine metallic and clay debris cannot be achieved by traditional fuel canister vacuum drying techniques. The steam reforming process eliminates any concern of flammable gas generation or over pressurization in the final Trojan fuel storage / transport containers.

As part of the spent fuel pool debris project, the steam reforming process was initially evaluated for effectiveness, a safety evaluation perfonned, a License Amendment was requested and ultimately received to perform processing in the Trojan Plant Fuel Building and a series of demonstration tests were scheduled and perfonned. The demonstration testing involved 3 major elements.

First, treatability testing was completed in a small 6-inch tube steam-reformer system at IT, Corporation Technology and Development Laboratory in Knoxville, TN. Each type of filter and flange material that was projected to be steam reformed was tested for treatability and an ,

assessment of potential operating conditions was performed. l Second, a Proof-of-Principle Test was completed at the SEG Bear Creek Facility in Oak Ridge, j TN. This test was performed in the commercial steam reformer to verify the ability of the steam reforming process to achieve the required low hydrogen content in the final residue of surrogate sock filters.

Third, two Integrated Tests were performed at the WNP-1 facility, in Richland, WA. The first test confirmed the physical fit characteristics and sorting capabilities of the Trojan spent fuel pool and transfer canal equipment and tooling and the second test demonstrated the integrated operation of the can feed evaporator and steam reformer, with detailed testing of the surrogate residue to confirm the equipment could achieve the required low hydrogen content and to establish appropriate process controls.

The second integrated test identified that steam reforming oflarge pieces of the black filter flange material (hereafter referred to as PVC) was more difficult than the poly;sapylere filter bag Mth polypropylene flange, as expected from the Proof of Principle Test results. The Test Plan had described the then existing plan of segregating the PVC seal rings of all bag filters with the other fuel pool debris which were capable of being treated as low-level radioactive waste. The PVC top and bottom of the cartridge filters in particular were to be removed and treated as low-level radioactive waste. Since the potential exists that some PVC material (primarily small fragments) may not be able to be segregated, the conservative approach oflimiting the quantity of the PVC 2 )

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material in each process can has been implemented. The empirical testing successfully l demonstrated that steam reforming could reduce the surrogate residue hydrogen generatmg  !

material to a quantity containing less than the required limit of 0.250 g-moles elemental hydrogen per process can capsule, when limited to 53.5 grams of the PVC (approximately 10 percent of a l PVC top and bottom flange) per process can (assuming 5 process cans per process can capsule). j While laboratory testing has indicated that, if the pieces of PVC material are relatively small in l size (less than or equal to 1/4 inch), the steam reforming process should be effective for almost i any expected quantity of PVC. The current Trojan procedures conservatively limit the quantity of PVC in any one process can to less than 52 grams of the PVC material, consisting of pieces which are less than or equal to 1/4 inch in size.

The detailed criteria for the successful completion of any given proce',s can production run was  !

determined during the second integrated test. This was accomplished by the empirical j demonstration of the potential hydrogen generating material remnining in the surrogate residue for l each process run. Enclosure 2 is the Trojan Second Integrated 'I est Data Report, which provides I a detailed description of the production runs that were performed during the test and the operational experience derived therefrom.

The second integrated test was also used to fine-tune the operating procedures and the operating j l

parameters for the can feed evaporator and the steam reformer. Modifications to the support equipment to facilitate effective steam reforming were made as well as enhancements to procedural operating controls.  ;

Description of Administrative Controls for Segregating PVC Material The Second Integrated Test Plan had described the then existing plan of segregating the PVC seal rings of all bag filters with the other fuel pool debris which were capable of being treated as low-level radioactive waste. The PVC tops and bottoms of the cartridge fP.cers in particular were to be removed and washed, if needed, to support disposal as low-level radioactive waste. Since the potential existed that some PVC material (primarily small fragments) may not have been able to be segregated, the conservative approach oflimiting the quantity rf the PVC material in each process can has been implemented. The empirical testing successfully demonstrated that steam reforming could reduce the surrogate residue hydrogen generating material to a quantity containing less than the required limit of 0.250 g-moles elemental hydrogen per process can capsule, when limited to 53.5 grams of the PVC (approximately 10 percent of a PVC top and bottom flange) per process can (assuming 5 process cans per process can capsule). While laboratory testing has indicated that, if the pieces of PVC material are relatively small in size (less than or equal to 1/4 inch), the steam reforming process should be effective for almost any expected quantity of PVC.

The current Trojan procedures conservatively limit the quantity of PVC in any one process can to less than 52 grams of the PVC material, consisting of pieces which are less than or equal to 1/4 inch in size.

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l The difficulty in steam reforming the PVC material required a change to the planned methodology for segregating the material that is to be steam reformed. Screen baskets have been installed in

series with the underwater vacuum system used to collect the debris material. This limits the size l of any potential PVC material deposited in a vacuum filter to less than approximately 1/32 of an inch. After using the vacuum system to collect large amounts of PVC material, the system was inspected. The screen baskets had collected all of the PVC material and none was visible on the vacuum filters. Even though no PVC was visible it is being assumed that approximately 2 grams of PVC material; less than 1/32 of an inch in size, passed through the screen basket and deposited on each filter.

Administrative controls are in place to limit the number of vacuum filters placed in any process can to a maximum of six. Therefore, there is assumed to be 12 grams of PVC material in a .

process can before any is intentionally added. Subtracting 12 grams from the 52 gram )

administrative limit leaves 40 grams of PVC material that may be added to each process can and I still remain below the limit of 0.250 gram-moles of hydrogen per process can capsule. The 40 gram limit will be implemented by the placement of any PVC material trapped in screen baskets, that cannot be released as low level radioactive waste, in a ' calibrated' cup in the Transfer Canal work platform prior to placement in a process can. Enclosure 3 contains the detailed procedural requirements for the segregation of spent fuel pool debris to ensure that the PVC material is administratively limited to be consistent with the qualifying test runs from the second integrated test.

Summary Description ofIntegrated Test Results SEG successfully conducted a full scale Proof of Principle Test to confirm that the steam reformer, a prototype can feed evaporator, and filters (referred collectively as the steam reforming system) was capable of meeting the nuclear spent fuel storage requirements. The first integrated test was conducted on the underwater handling equipment, and was successful. The results of these tests are not reported in this document, since the primary purpose of this report is to address steam reforming process issues. The second integrated test was conducted to establish appropriate process controls. Enclosure 2 is the Trojan Second Integrated Test Data Report, which provides the system descriptions, detailed data results and recommendations from the testing that was performed between May 12 and June 15,1997.

The steam reformer uses steam to heat, vaporize, and reduce organics into CO , CO, H , and 0 .

2 2 2 In addition, the high temperature steam drives out any inorganic hydrates to further reduce the residual hydrogen content. The surrogate waste to be processed was placed in a process can which has 5 micron filters in both the top and bottom. Superheated steam (at low flow rate and pressure) was passed through the process can and the resulting vaporized organics gas stream was reduced as it passes through the steam reformer's reactor. At the end of the process, the process can was purged with nitrogen to remove the steam and dry any remaining moisture, cool the process can, and prevent the reintroduction of moisture from atmospheric humidity. A nitrogen l purged transfer bell was used to transfer the process can to an argon purged glove box at the end l

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of the cool down period without exposure to air. The residue remaining in the process can was then removed, weighed, and analyzed for hydrogen content in the glove box.

The were two objectives for the second integrated test:

Confirm that the Trojan production debris processing system can reduce organic filters and dross material to a quantity containing less than 0.050 g-moles elemental hydrogen per process can. Since five process cans fit into one process can capsule, this limit equates to less than 0.250 g-moles total elemental hydrogen per process can capsule.

• Determine the process control limits for operating the debris processing system to assure that each batch processed meets the hydrogen specification. Specifically, determine the length of time, processing temperature, and flow rate required to successfully reform a process can full of organic filters and dross, determine the minimum drying and cooldown time for the post-processing nitrogen purge, and correlate on-line hydrogen and carbon monoxide instrumentation with measured hydrogen levels to provide process control information.

A total of nine surrogate waste runs were performed during the second integrated test program.

Four of these runs were test runs that did not contain any of the PVC material and five mns did contain at least some PVC material in addition to the polypropylene filter materials, dross and other surrogate wastes. Several mns tested the effectiveness of the use of stainless steel Raschig rings to facilitate the steam reforming process and varied the end-of-run criteria based on results of previous test run hydrogen analysis results. ,

i The test mns demonstrated that the established hydrogen limits can be successfully met for the polypropylene surrogate waste. The test mns also demonstrated that if PVC material is present in the surrogate waste, the PVC material must be limited in both quantity and form to successfully meet the hydrogen limits. The empirical limit for the quantity of PVC material in a given process l can is ten percent of a PVC filter top and bottom flange (53.5 grams) and is based on test run l number 9, and the experience derived from runs 3,4,6 and 7. Similarly, the form (size) of the PVC material is empirically limited to less than approximately 1/4 inch by 1/4 inch pieces.

Conservative administrative limits have been implemented for both of these requirements.

A single, conservative end-of-run criterion was established from the polypropylene and PVC material test runs. The single criterion was selected to ensure that should any PVC material be present (whether expected or not), the processed residue will always meet the hydrogen limit.

The integrated test equipment utilized in-line gas sensors for hydrogen, carbon monoxide, oxygen

and benzene that displayed near real-time concentrations in the steam reformer vent. Various gas

! analysis criteria for defining the end-of-run were used throughout the tests.

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I After nine runs with varying degrees of success, the end-of-run criteria which correlated best with a satisfactory residual hydrogen analysis was:

Hydrogen levels measured in the steam reformer vent line at or below 200 ppm for three continuous hours.

• If available, carbon monoxide and benzene levels remain at or below background levels throughout the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> period described above. (Background level for benzene is defined as less than 0.6 ppm)

The basis for these criteria is that the in-line gas analyzers displayed sharp peaks for hydrogen, carbon monoxide and benzene. All peaked nearly simultaneously. The carbon monoxide analyzer lacked sensitivity at levels below 200 ppm and, therefore, always returned to background level well before the hydrogen or benzene levels during the end portions of each run. Benzene levels also tended to level out at very low levels well before processing was complete.

Rather than establish a processing time limit (e.g. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />), the above criteria were selected as a direct measure of performance. In addition, the processing temperature was determined to be as follows:

• Superheater outlet temperature setpoint of 1175 degrees F, l

Steam inlet to process can temperature setpoint of 1150 degrees F, 1 1 Steam Jacket, or ' Wall' temperature of 1125 degrees F, and l

l

• Steam flow to the process can of 2 to 5 scfm.  !

l The minimum drying and cooldown time for the post-processmg mtrogen purge was empmcally  !

l derived by runs 5 and 9. The initial purge of steam from the can feed evaporator with 1000 '

i degree F nitrogen will be a minimum of 25 minutes with a nitrogen cooldown of the process can l to 500 degrees F in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (plus or minus 20 minutes).

The test results can be best summarized by concluding that surrogate waste residue with no PVC material and with seven filters placed into the process can passed the 0.05 g-mole hydrogen criterion more easily than if PVC material is present. As PVC particle size is reduced, the PVC is more thoroughly processed. For a given particle size, however, the resultant hydrogen content in the residue is proportional to the amount of PVC placed in the process can. This requires that there be both a limit on the size of the PVC material present in the process can and a limit on the quantity of PVC present in a process can. Controls have been implemented to ensure that each process can is loaded consistent with the limitation derived from the test results and that the process cans are processed such that the residual hydrogen is well within the established hydrogen limit.

6

Summary Discussion of Conservatism in Debris Processing A discussion of some of the various sources of conservatism in the entire spent fuel pool debris

, project processes is provided to facilitate NRC Staff review of the long-term suitability of the fmal I

process can capsule for storage in the ISFSI. The conservatisms are discussed within the context of each of the various process elements.

Segregation conservatisms include assigning 2 grams of PVC to each vacuum filter that is loaded into every process can whether or not any PVC material is visible, or whether the material vacuumed is suspected of containing any PVC.

The derivation of the process can capsule hydrogen limit contains the following conservatisms:

. The calculation that established the 0.250 g-mole per process can capsule was originally developed based on the preliminary process can and capsule design configurations. These containers were somewhat different in physical dimensions and their assumed volumes were somewhat different. These differences are conservative when compared to the as-built dimensions. PGE calculation TI-056 concluded that the existing limit of 0.250 g-moles of elemental hydrogen per process can capsule is conservative by approximately 18 percent.

i

• The calculations for the hydrogen limit (by both Battelle and PGE) conservatively assumed that the volume of debris that is steam reformed inside the process cans is i

reduced to less than 50 percent of the usable internal volume of each process can. The integrated test mns and proof of principle test runs have shown significantly larger reductions in volume to less than 20 percent of the usable internal volume. This additional free volume provides greater margin to the published NRC hydrogen limit of no more than 5 percent hydrogen by volume.

. The calculations for the hydrogen limit conservatively assumed that each process can capsule contained 5 process cans that contributed equally to the totallimit for the process can capsule, and that each was loaded with the maximum amount (0.050 g-moles) to reach the calculated limit. There are variations in the loading density of the process cans and the amount of potential hydrogen generating material loaded into each process can. The loading process for the process can capsules is based on the conservatisms demonstrated i in the proof of principle and integrated tests, which were based on maximizing the potential hydrogen generating material in a given process can (See also conservatisms in empirical testing methodology).

1 i The hydrogen generating material estimates derived from the empirical testing of the surrogate residue contain the following conservatisms:

7 l

l l

. The sample handling blank correction of 0.007 g-moles elemental hydrogen was not subtracted from the results. This represents 14 percent of the 0.050 g-mole limit, so the reported results overstate the actual identified hydrogen in the residue by this amount.

This typically introduced 10 to 23 percent overstatement of the actual hydrogen content of the residue.

. Two methods of determining the residual hydrogen in the residue were used. The more conservative of the results were reported. This typically increased the reported residual hydrogen by approximately 5 to 20 percent.

. Several sampling selection methods were used during the residue hydrogen content analysis process. The selection processes resulted in variations in the calculated hydrogen content in the residue. The " Biased" sample methodology generally resulted in the highest calculated hydrogen content.

. The early run hydrogen content calculations did not reflect a correction for the non-hydrogen generating material (e.g., stainless steel dross) that is a significant ponion of the surrogate material residue. The lack of dross correction had the result of overstating the hydrogen content oy as much as 70 percent.

Use of the NRC five percent hydrogen by volume limitation, which forms the basis for the calculated hydrogen limit, contains the following conservatisms:

. The NRC Information Notice 84-72, " Clarification of Conditions for Waste Shipments Subject to Hydrogen Gas Generation," states that it is " pertinent to shipments of resins, binder, waste sludge, and wet filters. It is not pertinent to dry compacted or uncompacted waste and irradiated hardware." This NRC Notice is focused on the potential for the generation of hydrogers from wet waste while being transported. The process can capsules used for the Trojan spent fuel pool debris residue will be dried at elevated temperatures and inerted during storage prior to being placed in the PWR baskets that will be used in the ISFSI. The concerns of this Notice are not applicable to the steam reformed debris package which will not contain free, interstitial and hydrated water.

. The NRC Notice permits limiting hydrogen generation to no more than 5 percent by volume of the secondary container gas void or inerting the container to ensure that oxygen is limited to 5 percent by volume. PGE has implemented both of these limitations, thereby providing greater confidence that a combustible concentration of hydrogen is not ponible.

. The Notice addresses the potential for hydrogen generation during shipment. The criterion is based on the hydrogen generation rate for twice the expected period of shipment. This means that the criterion reserves a margin of safety to the hydrogen flammability limit of at least two. This criterion has been conservatively implemented by assuming that all 8

potential hydrogen generating materials present in a given process can capsule eventually are converted to hydrogen, regardless of ti.me and radiation dose rates. l l

The process can capsule is not the formal containment boundary for the storage or transportation package. The process can capsule will be loaded into PWR baskets (that will be sealed) that are then loaded into an ISFSI stcrage cask. The percent volume limitation actually should be applied to the entire storage and transportation package, which is the PWR basket as a whole. The PWR j basket has a much larger free volume than that of the process can capsule. This represents an additional conservatism of approximately a factor of 10.

The design of the process can capsules provides confidence that no ignition sources exist in the loaded process can capsule, and that any residual hydrogen generated (i.e., from water) will not result in a hydrogen / oxygen mixture that could be combustible. In addition, the process can  :

capsule is purged and inerted with helium to ensure that oxygen is not present prior to being sealed for long-term storage and loaded into a PWR basket.

Corrosion bounding estimates described in Attachment I contain the following conservatisms:

• Assumes the presence of radioactive iodine, which no longer exists in the debris due to decay over the greater than 5 years since fuel use in the reactor and also due to exposure to the spent fuel pool environment.

• Assumes no mixing of the cesium and iodine salts within the residue, but rather assumes the salts are concentrated at the process can wall. ,

• Assumes no transport or carry-over of the outgassed cesium and iodine from the process can, rather assumes that the entire inventory of cesium and iodine remain captured in the process can during and following the steam reforming process.

• Assumes the quantity of cesium, telluriuni and iodine produced by fuel assemblies that

have 30,000 MWD /MTU burn-up, when the failed fuel assemblies that were subject to the bafile jetting problem at Trojan (the source of the fuel debris) varied in burn-up from approximately 8630 to 5878 MWD /MTU. This overstates the amount of cesium, tellurium and iodine available to form salts by approximately a factor of four.

The effect of these conservatisms is that PGE is confident that the planned spent fuel pool debris project debris segregation and steam reformation processes will result in a package that satisfies the pertinent regulatory requirements with margin and will ensure safe, long-term storage of the residue.

9 l

l

Conclusion The test program has demonstrated the suitability of the steam reforming process to reduce the potential hydrogen generating materials in the spent fuel pool debris to levels that will conform to regulatory criteria. The testing program has identified, and PGE has implemented, enhancements to the design of the steam reforming process equipment, segregation of materials processes and operation of the steam reformer that provide a high level of confidence that the process can capsules will contain no more than 0.250 g-moles elemental hydrogen. With the implementation of the established hydrogen limit, the conservatisms present in the methodologies used to implement the hydrogen limit, the inerted nature of the process can capsules and the lack of any ignition sources in the ISFSI PWR basket or the process can capsules, the steam reformed spent fuel pool debris is suitable for long-term storage in the Trojan ISFSI.

l i

i 10 i

t ENCLOSUREI t

DURATEK AND ITELLERGY CORPORATION LITERATURE SEARCH ON STEAM-FUEL AFFECTS I

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8 = URATEK 10100 Old Columbia Road Columbia, Maryland 2l046 410/312 5100 Fax 410/290-9070 June 11,1997 Steve Schneider Portland General Electric Trojan Nuclear Power Plant 71760 Columbia River Hwy.

Rainier, OR 97048

SUBJECT:

Affects of Steam on Nuclear Fuel Components Dear Mr. Schneider Attached, per your request, is the literature search and evaluation concerning the potential formation of corrosion products as a result of steam reforming.

Ifyou have any questions or require further information, please contact me at (509) 736-0626.

Sincerely, f S p/ A&_ - }- f Randal J. Roberts Manager, Western Operations l

AS97-150 l

INTELLERGY CORP.

6801 Sherwick Drive Berkeley, CA 94705-1744 Voice & FAX: (510) 841-3841 Home: (510) 841-9774 May 27,1997 TO: Randy Roberts / Gene Atchison FROM: Terry Galloway k Re: LITERATURE SEARCH ON STEAM-FUEL EFFECTS I have reviewed sixty pages of U.S. and foreign literature " hits" from 7 data bases using our literature search strategy for extracting about 120 relevant articles from 1948 to present on the effect of steam-reforming conditions on fuel cycle components that j might generate corrosion effects on final storage cask containers.

Below are the results of the literature search completed to date using article abstracts. We have requested the actual articles for review that document the effects in more detail and may reveal some subtle smaller effects. We have summarized our chemistry findings as listed below which provide negative corrosion effect results:

STEAM-ENHANCED OUTGASSING of Csl and CsTe:

The largest area of work reported in the literature involves the enhanced ,

outgassing of light fission products when there is a fuel element melt-down or reactor I failure. This effect appears to be completely consistent with what we would expect with l steam-reforming. There does not appear to be any corrosion enhancement of either  :

the fuel pellets or the container from the syngas components directly.

However, there is the effect of steam-reforming-enhanced outgassing of cesium (137Cs), tellurium (125Te), and iodine (i.e.1291) as final stable species which would then form Cs1 and CsTe salts, assuming the corrosion occurs only on the process can wall.

This mixture could be corrosive to stainless steel and other super-stainless and nickel-l based alloys. To test this corrosion mechanism, we had John Gonsky and Larry l Brackenbush of the GTS Duratek Northwest Office use the " LWR Radiological DATABASE" (copy attached as Appendix 1) to calculate the cumulative maximum fission yields of iodine, tellurium and bromine in their table " Trojan Spent Fuel Waste Suspected Radionuclides" that was originally used for shielding calculations (copy attached as Appendix 2) using the G.E. Chart of the Nuclides,13th Edition. The results were: I

! 129I = 0.06%

125Te = 1.02%

79Br = 0.00001%

and the cesium is in excess such that CsI, CsTe, and CsBr would be formed.

l l i

l

[

If we assume that we have 30,000 mwd /MTU x Ikg fissioned /925 mwd or 32.4 kg fissioned /MTU, then the above percentages can be used to obtain the worst' case upper bound iodine and tellurium mass fractions as:

1291 = 1.90 x 10-5 125Te = 3.34 x 104 If we assume that the worst case number of loose pellets is 375 and that the UO2 pellets are 14 mm in diameter (p=11 gm/cc) or 15 grams each, the total weight of pellets would be 5600 grams which would produce 1.9 gms CsTe and 0.11 gms Csl in the steam-reforming operation as worst case, upper bound. CsBr would have a negligible effect by comparison. Clearly this is conservative since the burnup on the Trojan Nuclear Plant failed fuel assembies varied only from 8629.7 to 5878.0 mwd /MTU (copy attached as Appendix 3).

The worst case scenario is that these cesium salts could eat a pit into the metal process can. To estimate the maximum pit depth, it _is assumed that 1 mole of cesium salt (MW=260) reacts with 1 mole of iron (MW=56). It is also assumed that 1/5 of the total number of pellets or 75 pellets find themselves in a single process can. (Although our recent sorting experience in the fuel pool with the high radiation level of the ~

individual pellets suggests that no more than a few pellets could find their way un-noticed into the process can; thus, this assumption is very conservative.) Thus,0.4 gms of cesium salt corrosives would react at most with 0.086 gms iron (p=7.86 gm/cc) or 0.011 cm3 of iron. If the pit is a hemielipsoid with minor axis of 3 mm in diameter at the top, the depth would be about 3.6 mm or 140 mils or around 1/8" depth. The conclusion would be that in a worst-case scenario, a pit would corrode through the process can wall (process can wall thickness = 0.105 inches), but would not penetrate the outer boundary provided by the process can capsule (process can capsule wall thickness

= 0.250 inches).

As a comparison,if all of the Csl and CsTe existed in the gas phase in the process can at maximum temperature and uniform corrosion of the can were to occur, only 25 pm or 1 mil would be corroded over the surface. This mechanism is of no consequence.

CESIUM CORROSION EFFECTS ON FUEL ELEMENT CLADDING:

The literature shows there to be an effect of cesium compounds on fuel element cladding corrosion as observed in sodium-loop breeders. For the purpose of this study, we have assumed that this corrosion mechanism related to corrosion of the process can by the scenario iliustrated above.

URANIUM OXIDE CONVERSIONS WITH STEAM-REFORMING:

There is a body of literature revealing the various conversions of the uranium and plutonium oxides to the lower oxidation states, such as UO2 and PuO2, under i

i

l e , s l

l steam-reforming conditions, but no corrosion enhancement effects wee revealed.

Also our steam-reforming conditions are sufficiently mild that there is no further reduction of these oxides to the metallic state. Appendix 4 includes a copy of the internal Westinghouse Science and Technology Center study completed for this  :

purpose).

MIXTURES OF Cs-U-O SYSTEMS RESPONSE TO STEAM-REFORMING:

This area was also examined in the search to examine any cutectic-enhanced corrosion. Although the Cs-U-O converted to Cs2UO 4which was the stable state, no corrosion enhancement was observed.

Tropn-Lit. Search 3 I

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.. . . - . . . .- - - - . . - . - - . _ - . . - - . - - - . ~ . _ , . . _ - _ _ . -

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,07/ggty: ue A#t wi Wts:,uraat-Rmee SEG-raJ.e.

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509 735 6920 TO 15108413841 P.03/03 App:ndix 1 PAGE: 1 lwr Radiological DDramacE l CURIES BY ISOTOPE REACTOR TYPE & RURNUP: PWR 10000 l ENRICMMENT: 2.391 l DECAY TIME: 15 YEARS DATA SENSITIVITY: 0.01%

............=.....................=.=.. ..................=.........===.==......

! ISOTOPE CURIES /MTIHM tTOTE L ........... ....... .............................=...=.......................... ~

! E 3 1.337E+02 0.13%

PE 55 2.731B+01 0.03%

CO E0 4.954E+02 0.48% #MA[ gY

• M 2.25

! NI 63 1.383E+02 0.13% / f KR 85 1.265E+03 1.21% W_ 4"af SR 90 1.784B+04 17.12%

Y 90 1.784E+04 17.12% g/g ,

S3125 1.328E+02 0.13%

TS125M 3.242E+01 0.03%

! CE134 1.2985+02 0.12%

CS137 2.270E+04 21.79%

RA137M 2.148E+04 20.61%

PM147 1.703E+03 1.63%

SM151 2.053E+02 0.20% .

EU154 2.674E+02 0.26%

EU155 1.436E+02 0.14%  ; g k sys,, k/

i PU238 1.440E+02 0.14% J

PU239 2.270E+02 0.22%

l PU240 PU241 1.632E+02 1.845E+04 0.15%

17.71%

Q g ,.%L

• i AM241 6.548E+02 0.63%

1 i TOTE 1.042E+05 99.97% 4 I b d5 M I NE-E

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M 'M 15:M 509 946 0199 PAE.83

• • TOTAL FVGE.03 **

App ndix 2 TROJAN SPENT FUEL WASTE SUSPECTED RADIONUCLIDES (activity per pellet used in shielding calculations)

. n ;:. :

Nuclide curies becquerels uCi/cm^3 Bqc/cm^3 Am-241 3.9290E-03 1.4537E+08 5.5389E+03 2.0494E+08 Ba-137m 1.2890E-01 4.7693E+09 1.8172E+05 6.7235E+09 Co-60 2.9720E-03 1.0996E+08 4.1898E+03 1.5502E+08 Cs-134 7.788-E-04 2.8816E+07 1.0979E+03 4.0623E+07 .

Cs-137 13620E-01 5.0394E+09 1.9201E+05 7.1043E+09 Eu-154 1.6040E-03 5.9348E+07 2.2612E+03 8.3666E+07 Eu-155 8.6160E-04 3.1879E+07 1.2146E+03 4.4942E+07 Fe-55 1.6390E-04 6.0643E+06 23106E+02 8.5491E+06 H-3 8.0220E-04 2.9681E+07 1.1309E+03 4.1843 E+07 Kr-85 7.5900E-03 2.8083E+08 1.0700E+04 3.9590E+08 Ni-63 8.2980E-04 3.0703E+07 1.1698E+03 4.3283E+07 Pm-147 1.0220E-02 3.7814E+08 1.4408E+04 53308E+08 Pu-238 8.6400E-04 3.1968E+07 1.2180E+03 4.5067E+07 Pu-239 13620E-03 5.0394E+07 1.9201E+03 7.1043E+07 Pu-240 9.7920E-04 3.6230E+07 1.3804E+03 . 5.1076E+07 Pu-241 1.1070E-01 4.0959E+09 1.5606E+05 5.7742E+09 Sb-125 7.9680E-04 2.9482E+07 1.1233E+03 4.1562E+07 Sm-151 1.2320E-03 4.5584E+07 1.7368E+03 6.4262E+07 Sr-90 1.0700E-01 3.9590E+09 1.5084E+05 5.5812E+09 -

Te-125m 1.9450E-04 7.1965E+06 2.7420E+02 1.0145E+07 Y-90 1.0700E-01 3.9590E+09 1.5084E+05 5.5812E+09 ISOTOPE Ci/MTIHM 375 Pellets 1 Pellet 20 Pellets 25 Pellets 30 Pellets (Ci) (Ci) (Ci) (Ci) (Ci)

Am-241 6.548E+02 1.473E+00 3.929E-03 7.858E-02 9.822E-02 1.179E-01 Ba-137m 2.148E+04 4.833E+0! 1.289E-01 2.578E+00 3.222E+00 3.866E+00 Co-60 4.954E+02 1.115E+00 2.972E-03 5.945E-02 7.431E-02 8.917E-02 Cs-134 1.298E+02 2.921E-01 7.788E-04 1.558E-02 1.947E-02 2336E-02 Cs 137 2.270E+04 5.108E+01 1362E-01 2.724E+00 3.405E+00 4.086E+00 i Eu-154 2.674E+02 6.017E-01 1.604E-03 3.209E-02 4.OllE-02 4.813E-02 Eu-155 1.436E+02 3.231E-01 8.616E-04 1.723E-02 2.154E-02 2.58$E-02 Fe-55 2.731E+01 6.145E-02 1.639E-04 3.277E-03 4.097E-03 4.916E-03 H-3 1.337E+02 3.008E-01 8.022E-04 1.604E-02 2.006E-02 2.407E-02 l Kr-85 1.265E+03 2.846E+00 7.590E-03 1.518E-01 1.898E-01 2.277E-01 Ni-63 1.383 E+02 3.112E-01 8.298E-04 1.660E-02 2.075E-02 2.489E-02 Pm 147 1.703 E+03 3.832E+00 1.022E-02 2.044E-01 2.555E-01 3.065E-01 j Pu-238 1.440E+02 3.240E-01 8.640E-04 1.728E-02 2.160E-02 2.592E-02

{

Pu-239 2.270E+02 5.108E-01 1.362E-03 2.724E-02 3.40$E-02 4.086E-02 l Pu-240 1.632E+02 3.672E-01 9.792E-04 1.958E-02 2.448E-02 2.938E-02 I Pu-24 l 1.845E+04 4.151E+01 1.107E-01 2.214E+00 2.768E+00 3.321E+00 Sb-125 1.328E+02 2.988E-01 7.968E-04 1.594E-02 1.992E-02 2390E-02 Sm-151 2.053E+02 4.619E-01 1.232E-03 2.464E-02 3.080E-02 3.695E-02 Sr-90 1.784E+04 4.014E+01 1.070E-01 2.141E+00 2.676E+00 3.211E+00 Te-125m 3.242E+0! 7.295E-02 1.945E-04 3.890E-03 4.863E-03 5.836E-03 Y-90 1.784E+04 4.014E+0 l 1.070E-01 2.I41E+00 2.676E+00 3.211E+00

)

. -- __-.- .- . _ _ . - - - - - - . - . . - . - _ . . . . - - . . . - . - . . . ~ - . . _ . . - . - . - .

. ,  ; l 509 ~735 6920 To 15108413841 P.02/03

.......u.....w .y a yz g.t t... W J.S.,., *-wir ==a a ss i .

yo 2 aisiid"" *p.et et l

Appendix 3 1.0 Specincetion TD-012 j I

Spent FeelPeel Debris levantory I

! The ameumofdebrisin spentfbalpool ceDs A13. A14, A27 A29, A43. A4 EE12. FF12, and 00121s speauledve as indicated byihe videos prus l ,

prior to this Invitation to Bid. The foBowing pages idenbry the materials thens seus.

detain regart's The to the fbel eens*debris. quanehies cannot be daarnimods for orical servain dueto The Asel peusts and fbel fragments are from fuel ====mhhes F02. F0 F19 and F56. These fbel assemb6es were tumoved from the more in , ,

i jettbig incident. The precise number ofpeusts or thes aruount of dee ets in the debris containers or in the vacuum 6hers is oncertain. 7eBed fad d estimates the total number of peness lost was 375. Of these 375 pelles  ;

accounted for and ers in ths 81 tars or debris containers.

! g l

snrichment, and bomup ofthe assembEcs desmus ,

, FaDedFuel AssarnblyData '

Fuel Initial Uranium

_ Assembly Initia1 Enrichment ]pumup t , (ka) (*t % U*") mwd /MTU F02 458.04 _

) 3.2 6088.7 F03 459.63

• 3.2 6o15.s F04 461.39 .

3.2 5878.0 7 05 456.63 3.2 6143.4 F07 459.25 3.2 6070.1 F18 459.72 3.2

• 6226.8 F19 460.51 3.2 5629.7 F56 459.54 " .

3.2 5998.] .

,e &

w a r. , sv.+g o ..**s"rv

% har 19g2, cs u3 , #

L 01

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Appendix 4__

From- Science and Technology Center ws: 236-2238 Date: March 12,1996 subiect Reactions of Uranium Dioxide in Svnthetica Process To: D. Schmoker (SEG) l t i l i ec: D. L. Keairns Per your request, I have looked at the potential for forming uranium metal from uranium dioxide in a variety of reducing gases. The two gas compositions and reaction conditions which were used in the. analysis were:

1. 25% each of CO, H,, H,0 and CO,.

2. 50% each of CO and H,, i

3. weight ratios of UO, to gas of.01, .1,1,and 10;

4. temperatures from 373*K (212*F) to 1373*K (2011*F) in 100*K increments.

The results indicate that UO,is:

1. not oxidized to higher oxide levels,

2. nor reduced to metal.

These calculations were performed using the NASA Lewis CET89 thermodynamic code. The results are attached. Note that all the potential compounds which were being looked for (U(S), U(L), UC(S), U,C,(S), UH,(S), UO,(S), U,0,(S) and U,0,(S)}

were considered and rejected on a thermodynamic basis as being unstable under the given conditions.

Edward Lahoda em.m aww.sw I

85/86/97 16:42:24 -> 518 841 9774 Pcue 882 tr ' l Toltech' May 6.1997 IJterature Search l i

mie.  :

Effect ofSteam on Nuclear s

Fuel Components For:

Mr. Terry Galloway 1234 Columbia Driw SE Richland, WA 99352 Fax: 509-7354920 Comment: I searched in sewral of the nuclear science, energy, engineering and chemical databases for literature on the effects of a reducing environment on nuclear fuel components. I tried several different approaches to the strategy. In general, I tried

• uranium or plutonium or strontium or cobalt or cesium " and " nuclear" and " fuel" and

" reducing or reduction or hydrogen" and " steam". This still gaw quite a wide range of topics. I tried to pick through them to get rid of

-all of the waste recowry related articles.

I hope you find some helpful articles in here. If you want to pursue this a bit further, let me know which articles are the best and I can look at how they are indexed to try to find similar articles. For example, I saw quite a bit indexed under" fission products" but I am not sure if that is what you are after.

Please giw rne a callifWu how any questions regarding this search or require further assistance.

Thank you for using Teltech!

-Tina Analyst: Tina Mattia Page 1 of 25 Phone: 1-800-367-8358 Ext: 7156

_ - _ _ _ h

85/86/97 16:43:00 -> 518 841 9774 Pes 2 883

)

O File 109:NuclearSci. Abs. 1948-1976 (c)1997 Contains copyrighted material File 103 Energy SciTec 1974-1997/Apr B1 (c) 1997 Contains copyrighted material File 8:Ei Compendex(R) 1970-1997/Jun W1 (c) 1997 Engineering Info. Inc.

File 2:lNSPEC 1969-1997/Apr W5 (c) 1997 Institution of Electrical Engineers File 6ETIS 6+1997/Jun W3 Comp & distr 1997 NTIS, Int! Copyright All Rights Re File 144: Pascal 1973-1997/Apr (c) 1997 INIST/CNRS File 434 Scisearch(R) Cited Ref Sci 1974-1997/Apr W4 (c) 1997 Inst for Sciinfo 0 14/9/14 (Item 9 from file: 103)

DIALOG (R) File 103: Energy ScsTec (c) 1997 Contains copyrighted material. All rts, reserv.

01561535 AIX-16020392;EDBA5068309

Title:

Materials for *nuclea r* energy Author (s): Rand, M.H. (UKAEA Atomic Energy Research Establishment, Harwell.

Materials Dewlopment Div.)

Source: Pure Appl. Chem. (United Kingdom) v56:11. Coden: PACEA Publication Date: Nov 1964 p 1545-1554 Document Type: Journal Article Ianguage: English Journal Announcement: EDB8502 Country of Origin: United Kingdom Abstract: The paper reviews the state of some current areas of interest in reactor materials, with particular reference to reactor

• fuels

• and to the release of fission products from a degrading core and their subsequent containment. The following topics are discussed.1. The properties of urania at high temperatures. An analysis of the wrious theoreticalcontributions to the heat capacityof urania accounts satisfactonly for the enthalpy increment up to 2500 to 2600 K. Abow this temperature an additional contnbution to the enthalpy is required the data suggest that the heat capacity of UOlsub 2/ is essentially constant from 2600 K to 3120 K, the melting point. This phenomenon is presumably related to the increasing Frenkel disorder found in UO/sub 2/ (and other compounds with the fluorite s'ructure) at temperatures abow c. 0.8 times their melting points. The methods used to extrapolate the oxygen potential of UOsub(2-x) through the '

solidus-liquidus are discussed, with emphasis on the difference in slope of the Gsub(0/sub 20 vs. T curw in the two regions. 2. The interaction of fission products released from a degrading LWR core with

• steam *-hWrogen mixtures, with special emphasis on barium,

• strontium

• and tellurium. 3. The importance of gaseous iodine compounds other than 1/sub 2/(g) owr iodine solutions (for containment studies).;

2 Teltechs Uterature Search Serwce Effect of Steam on Beuclear Fuel Componente May 6,1997

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00203629 ERA 42421905;EDB-7744130

Title:

• Strontium

• heat source dewlopment program Author (s): Jarrstt, J.H.; Fullam, H.T.; Fullam, H.T.; Harmon, K.M.

Title:

Quarterly report on the

• strontium

• heat source dewlopment program and the beneficialisotopes utilization program,

• Nuclear

• Research and Applications Division, July- September 1976 Corporate Source: Battelle Pacific Northwest Ia bs., Richtend, Wash. (USA)

Publication Date: Oct 1976 p 1-13 Report Number (s): BNWL-1845-28 Co ntract Numbe r (DOE): EY-76&O6-1830 .

Document Type: Analytic of a Report Ienguage: English Joumai Announcement: EDB7703 Subfile: ERA (Energy Research Abstracts); TIC (Technicallnformation Center).

Country of Origin: United States Countryof Publication: United States Abstract: The objectims of the program for FY-1976 are to identify and dewlop beneficial uses of nuclear reactor by-products through: (1) estimation of long-term awilability and cost of useful isotopes from commercial suppliers;(2) identification and dewlopment of beneficial .

applications of isotopes, including their use in remote regions of the world;(3) identification and ewluation of the actions required to optimize the /sup 90/*Sr*/sub 2/ and /sup 137/CsCl products from the Hanford Waste Encapsulation and Storage Facility (WESF) for beneficial use; and (4) review and owluate the use of radioisotopes for ' fueling

• thermal systems which might be used by the *U*.S. Army. Progress is i

reported on the four main tasks: (1) isotopes awilability; (2) cold i regions applications; (3) WESF product utilization; and (4) general application of radioisotopes in army thermal" systems.

Major Descriptors:

• CESIUM

• 137 - RADIOISOTOPE HEAT SOURCES ' ISOTOPE ,

APPLICATIONS - RESEARCH PROGRAMS;

• RADIOISOTOPE HEAT SOURCES -

RESEARCH PROGRAMS; ** STRONTIUM

• 90 - RADIOISOTOPE HEAT SOURCES l 3

Teltech' Uterature Search Serece Effect of Steam on Nuclear Fuel Componente

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l 0 14/9/21 (Item 1 from file: 6)

DIALOG (R) File 6NTIS l Comp & distr 1997 NTIS, Int! Copyright All Rights Re. All rts. reserv.

l 0971167 NTIS Accession Number NUREG/CR-2928 Data Summa ry Report for Fission Product Release Test Hi-1 i Osbome, M. F. ; lorenz, R. A. : Travis, J. R. ; Webster, C. S.

! Oak Ridge National Lab., TN.  !

l Corp. Source Codes: 021310000 Sponsor Nuclear Regulatory Commission, Washington, DC. Office of Nuclear Regulatory Research.; Department of Energy, Washington, DC.

Report No.: ORNI/rM4500 Dec 82 50p languages: English Joumal Announcement: GRA18313 NTIS Prices: PC A03/MF A01 Country of Publication: United States Contract No.: W-74054ng-26 The first in a series of high-temperature fission product release t ssts was conducted' for 30 min at 1400C, with the release taking place into flowing steam. The fuel specimen was a 20<m-long section of H.B. Robinson 1 fuel rod, irradiated to 28,000 mwd per metric ton (t). After the test, the Zircaloy cladding of the specimen was almost completely oxidized and was quite fragile. The fission product collection system included a thermal gradent tube (700-1500), fdters, heated charcoal, and cooled charcoal.

Gamma rayanalysis of apparatus components and collectors showed that about 2.83% of the 85Kr and 1.75% of the 137Cs were released from the fuel.

Activation analysis of leach solutions from these components indicated that {

2.04% o{ the 1291 was released. Other analyses rewaled small but >

significant releases of the radionuclides 125Sb and 106Ru,and of the elements Br, Rb, Sr, Zr, Ag, Sn, Te, Ba, and Ia. In addition, sewral other elements which originate as furnace materials or as impuritieswere collected.

Descriptors:

• Nuclear reactor accidents; Nucrear fuels; Steam; Tests identifiers ' Fission product release; Light water reactors; NTISNUREG; O 22/9/14 (Item 3 from file: 103)  !

DIAIDG(R) File 103:EnergySciTec (c) 1997 Contains copyrighted material. All rts. reserv.

?

03109106 AIX-22 030937; EDB-91446539; ERA-16014993

Title:

Kinetes of UO sub 2 (s) dissolution under

• reducing

• conditions:

Numerical modelling t Author (s)/ Editor (s): Puigdomenech, l. (Studsvik Nuclear, Nykoeping (Sweden)); Casas, l. (Universidad politecnica de Cataluna, Barcelora (Spain). Dept. of Chemcal Engineering); Bruno. J. (RoyalInst. of ,

Tech., Stockholm (Sweden). Dept. of Inorganic Chemistry) '

Corporate Source: Swedish Nuclear Fuel and Waste Management Co.,

Stockholm (Sweden)(Code: 9060197)

Publicaton Date: May 1990 (27 p)

Report Number (s): SKB-TR-90-25

Order Number DE91624276 -

l DocumentType Report l 12nguage In English 4

l Toltech' Literatum Search Service Effect of Steam on hcieer Fuel Componente May 6,1997 1

85/86/97 16:45:23 -> 518 841 9774 P:s2 BB6

. t F l l Joumal Announcement: EDB9108 Awilability: OSTl; NnS (US Sales Only); INIS '

Distribution: (Report):9 (MF):3 MN-000 Subfile: ERA (Energy Research Abstracts);ETD(EnergyTechnologyData Exchange); NTS (NTIS). SWDN(Sweden (sent to DOE from))

US DOE Project /NonDOE Project:NP Country of Origin: Sweden Countryof Publication: Sweden Abstract: A numerical model is presented that describes the dissolution and precipitation of UO{sub 2)(s) under reducing conditions. For aqueous solutions with pH>4, main reaction is: UO{sub 2)(s)+2H(sub 2)O{r rewrsible}U(OH){sub 4)(aq). The rate constant for the precipitation reaction is found to be log (k{sub p))=-1.2{plus minus)0.2 h{sup-1) m{sup -2), while the wlue for the rate constant of the dissolution reaction is log (k{sub d))=-9.0{plus minus)0.2 mol/(1 h m{sup~2)). Most '

of the experiments reported in the literature show a fast initial dissolution of a surface film of hexavalent

• uranium *

• oxide *. Making the assumption that the chemical composition of the surface coating is U{sub 3)O{sub 7)(s), we how deriwd a mechanism for this process, and its rate constants haw been obtained. The influence of HCO{sub 3}{sup

-) and CO{sub 3){sup 2-) on the mechanism of dissolution and precipitation of UO{sub 2)(s) is still unclear. From the solubility rneasurements reported, one may conclude that the identity of the ,

aqueous complexes in solution is not well known. Therefore it is not possible to make a mechanistic interpretation of the kinetic data in carbonate medium. (oria.). .

O 22 S/18 (item 1 from file: 8)  !

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(c) 1997 Engineerire Info. Inc. All rts. reserv.  !

i 03462968 E.I. Monthly No: E!9208103083 I

Title:

Microstructures and phase relationships of crystalline oxidation products formed on unused CANDU

• fuel

• exposed to aerated ' steam

• and aerated water near 200 degree C.

Author Taylor, Peter; Wood, Donald D.; Owen, DerrekG.; Hutchings, William O.; Duclos, A. Michael Corporate Source: Whitshell Lab, Pinawa, Ma nit, Can Source: Atomic Energy of Canada Limited, AECL (Raport) n 10476 Nov 1991 p 1-71 Publication Year: 1991 CODEN: AECRAN ISSN:0067-O'M7 Ianguage: English Document Type: RR; (Report Review) Treatment: X; (Experimental); L; (Literature Review / Bibliography)

Journal Announcement: 9208 Abstract: This report reviews the findings from dry, moist and wetair oxidation experiments on unused UO!/2 fuel specimens at 200-225 degree C, l performed in support of the Dry Storage Program for used CANDU (Canada  !

Deuterium Uranium) fuel. The presence of liquid water, or unsaturated steam, adds to the complexity of air oxidatior: of UO//2. The following l

processes haw been identified by using a combination of optical and scanning electron microscopy and X-ray diffraction to detect oxidation products, and are discussed in this report: oxidatite dissolution of U(VI) 5 Toltech' Uterature Search Service Effect of Steem on Nucleer Fuel Componente May 6,1997 l

i l

. _ _ _ _ ~__ _ - _ _ . __. ____. _ _ _ - . _

85/86/97 16:46:16 -> 518 841 9774 Pese 887

\ .. <

l l

l l

l and precipitation of hydrated UO//3; back-reduction of dissolved U(VI) and precipitation of U//30//8 on the UO//2/U//30//7 surface; solid-state surface and grain-boundary oxidation of UO//2 to beta -U//30//7; preferential desolution of UO//2 grain boundaries, sometimes followed by the filling of the resulting gap with higher

• uranium *

• oxide *(s). (Edited author abstract) 50 Refs.

! Descriptors:

• NUCLEAR * *FUEM* *0xidation;

• URANIUM

• DOXIDE-Oxidation;

! CRYSTAM-Microstructure; RADIOACTIVE WASTES-Storace

! O 22/9/1 Otem 1 from file: 109)

DIALOG (R) File 109:NuclearSci. Abs.

(c)1997 Contains copyrighted material. All rts. reserv.

I 606574 NSA-24407125

' IRRADIA110N EXPERIMENTS ON FUEL ELEMENTS FOR A STEAM-COOLED FAST l REACTOR.

l Antoni, R.; Hoechel, J.; Plitz, H.

AEG-Telefunken, Grosswelzheim, Ger.

Tra ns. Amer. Nucl. Soc., 12: 606-7(Nov.1969).

Publication Date: 1969 Note: From 17th Conference on Remote Systems Technology, San Francisco, Calif. See CONF 491102.

Joumal Announcement: NSA24 Document Type: Journal Article Language: English Subfile:NSA (NuclearScience Abstracts) l_ Work Location: DE Descriptors: FAST REACTORS; FUEL CANS; FUEL ELEMEhTS; NEUTRONS; PELLETS

PLUTONIUM OXIDES; POWDERS; FOWER REACTORS; RADIATON EFFECTS;  ;

SUPERREATING; URANIUM DIOXfDE l

O 22G2 Otem 2 from file: 109)

DIALOG (R) File 109:NuclearSci. Abs.

(c)1997 Contains copyrighted r6atenal. Al! rts. reserv.

554351 NSA-23407391 l FURTHER REACTOR PHYSICS STUDIES FOR STEAM GENEF_ATING HEAVY WATER i l REACTORS. PART 3. COOLANT TEMPERATURE EFFECTS IN LOssub 2S AND PuOSsub 2S/UOSsub 2S FUEM.

Briggs, AJ.; Johnstone, l.; Newmarch, D.A; Kemshell, P.B.

United Kingdom Atomic Energy Establishment, Winfnth Eng.

J. Brit. Nucl. EnergySoc.,7: 35345(Oct.1968).  !

Publication Dato: 1968 Joumal Announcement: NSA23 Document Type: Journal Article l Language: English l Subfile: NSA (Nuclear Science Abstracts) l Work Locaten: United Kingdom l Descriptors: CONFIGURATION COOLANT LOOPS; FUEL ELEMENTS: FUEM: HGH TEMPERATURE; PERFORMANCE; PLUTONIUM OXIDES; REACTi\ TTY; REACTORS; l SGHWR; l

i URANIUM DIOXIDE i

6

, Teltech' Literature Search Service Effect of Steam on leucdeer Fuel Componente May 6.1997 l

I j

- . . ,. -- - . . -. ~. . - - . . .,

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e* J .

O 2585 (Item 5 from file: 109) '

DIAIOG(R)Fde 109:NuclearSci. Abs.

(c)1997 Contains copyrighted material All rts. reserv.

854202 NSA-29-018780 Chemkal Engineering Division fuels and materials chemistry semianntal report, January-June 1973 (U-Pu-O-Na and U-Nd-O systems; i

Chasanov, M.G.; Johnson, C.E.; Dudey, N.D.; Blackbum, P.E.; Fluss, M.J.;

Heinrich, R.R.; Johnson,1,; Martin, A.E.; Tetenbaum, M.; Crouthamel, C.E.;

1mnson, M.; Vogel, R.C.; Webster, D.S.; Burris, L.

Argonne NationalIab., fit (USA)

Corp. Source Code:0448000 Pubi; cation Dato: Dec 1973 72 p.

Primary Report No.: ANL-8022 Joumal Announcement: NSA29 Awilability AT Document Type: Report Ianguage: English Subfile:NSA (NuclearScience Abstracts)

Work 4 cation: United States Contiact No.: W 31 109-ENG-38 To aid in evaluating the performance of fast-breeder reactor fuels, high-temperature studies are being made of the U - Pu O-Na and U-Nd-O systems and modeling studies of oxygen potentials owr oxide fuels are being performed. Additional tests of the reaction of sodium with l

mixed-oxide pellets and powder are reported. Oxwen potentials owr the U-Nd -O system are reported. Applications of the previously presented theory for oxygen petentets owr urania-plutonia systems are made to the evaluatio,n of analytical methods for oxwen determinations in U-Pu-O fuel.

Laboratory experiments are being performed to assess the effect of selected 1

fission-product elements and other potential fuel-element impurities on the attack of stainless steel cladding. The effect of solution-treating of Type ^

316 stainless steel on the rate of t,ttack of the steel by cesium compounds

& Cg gg  ;

i is reported. Results of ion microprobe analysis of an irradiated fuel for fission-product cesium are presented. Conditions leading to the reaction of M WM .

fission-product cesium with urania blanket pellets are discussed.

Fast-neutron dosimetry techniques are being dewloped to improve the accuracy of fission-rate measurements by foil actiwtion and by solid-state ,

track recorders (SSTRs). Resuits of fission-rate measurements by these two k methods are compared with results obtained from fission chambers in the same irradiation. A study to assess the feasibility of extending the use of l fast-reactordosimetry techniques to higher neutron energies is described; these dosimetry techniques are of interest for studying irradiation damage effects in neutron environments applicable to controlled thermonuclear research (auth)

Descriptor Groups (Splits): BURNUP-CARBON DIOXIDE-CARBON MONOXfDE-l  ;

CHEMICAL ANALYSIS-EQUILIBRIUM-HIGH TEMPERATURE-HYDROGEN-LATTICE '

PARAMETERS-LMFBR TYPE REACTORS-NEODYMlUM OXfDES-NUCLEAR FUETIS-OXYGEN- PHASE DIAGRAMS-PHASE STUDIES-PLUTONIUM OXIDES-REACTrON KINETICS-REDOX POTENTIAL-REDUCTION-SINTERING-SODIUM OXIDES-THERMODYNAMIC PROPERTIES- URANIUM OXfDES-VAPORIZATION HEAT-VARIATIONS-VERY HIGH TEMPERATURE -

! 7 Tottech' Uterature Search Service Effect of Steam on Nuclear Fuel Componente May 6,1997 l

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DIAI4G(R) File 434 Scisearch(R) Cited Ref Sci (c) 1997 inst for Sci info. All rts. reserv.

14423203 Genuine Article #:TL830 Numberof

References:

24

Title:

EXPERIMENTAI-STUDY ON TEMPERATURE-COEFFICIENT OF REACTIVITY IN LIGHT-WATER-MODERATED AND HEAVY-WATER-REFLECTED CYLINDRICAL CORE IDADED WITH HIGHLY-ENRICHED-URANIUM OR MEDIUM-ENRICHED-URANIUM .

FUEL Author (s): SHIROYA S; MORI M; MISAWA T; HAYASHI M; KOBAYASHI K; KANDA K Corporata Source: KYOTO UNIV,1NST RES REACTOR /KUMATORl/ OSAKA 59004/ JAPAN /;

KYOTO UNIV, DEPT NUCL ENGN,SAKYO KU/ KYOTO 60601// JAPAN /

Joumal: JOURNAL OF NUCLEAR SCIENCE AND TECHNOIDGY,1995, V32, N116NOV), P 1081-1089 ISSN: 0022-3131 Ianguage:ENGLISH DocumentType: ARTICLE Geographic Imcation: JAPAN Subfile: SciSearch; CC ENGI-Current Contents, Engineering, Technology &

Applied Sciences Journal Subject Category NUCLEAR SCIENCE & TECHNOIDGY Abstract: By using the Kyoto Uniwrsity Critical Assembly (hTCA), a series of critical experiments was performed to maasure the temperature coefficient of reactivity in a light-water-moderated and heavy-water-reflected cylindrical core loaded with highly 4nriched-uranium (HEU) or medium 4nriched-uranium (MEU) fuel.

The measurement was performed for tho approximately 20 to 70 degrees C range to examine the effects of the size of light-water region in a leterogeneous multiregion type core, the reduced U-235 enrichment, and the existence of boron burnable poison (BP) on this quantity by using six types of core configurations. In all the six types of cores, there were large light-water regions at the center of core and between the outer fuel region and the heavy-water reflector region, and it was found that these light-water regions caused a remarkably positiw effect on the temperature coefficient of reactivity. In the present study, the temperature coefficients of the MEU core and the core without BP were more positiw than those of the HEU core and the core with BP, respectiwly The size of light-water region had a larger effect on the temperature coefficient rather than the reduced U-235 enrichment and the existence of BP, The negatiw temperature coefficient would be realized by reducing the thickness of light-water layer existed in the core.

8 Toltech* Literature Search Serwce Effect of Steam on feuclear Fuel Componente i

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i O 33/9/8 (Item 8 from file: 103)

I DIALOG (R) File 103:EnergySciTec

(c) 1997 Contains copyrighted material. All rts. reserv.

00862879 EDB-82-037719 l Author (s): Ikeda, S.; Kobayashi, H.; Nemoto, S.; Tsunoda, N.

l

Title:

Process for

• reducing *

• plutonium * (Patent)

Patent No.: US 4197274 i Patent Assignee (s): Doryokuro Kankunenryo Kaihatsu Jigwdan, (Japan) l Patent Date Filed: Prioritydate 5 Apr 1977 Japan Publication Date: 8 Apr 1980 pv Document Type: Patent Ianguage: English Journal Announcement:EDB8202 Subfile: IFl (Claims /U.S. Patent Abstracts). l Countryof Origin Japan '

Countryof Publication: United States Abstract: A process is disclosed for reducing plutonium is provided in

( which hydrazine is added as a reducing agent to a nitric acid solution of tetrawlent plutonium in the presence of actim carbon catalyst to thereby reduce tetrawlent plutonium into triwlent l 0 33"3/17 (ltem 17 from file: 103) l DIALOG (R) File 103: Energy SciTec (c) 1997 Contains copyrighted material. All rts. reserv.

00169786 INS-77-000241; ERA-02-010496; EDB-77-007174 i

Author (s): Woodley, R.E.

Title:

Einetics of *hWrogen*

• reduction

• of U/sub 0/.!sub 75/*Pu*/sub 0//sub 25/0/sub 2/

Corporate Source: Hanford Engineering Dewlopment Lab., Richland, Wash.  !

(USA) i Conference

Title:

78. annual meeting of the American Ceramic Society I Conference Location: Cincinnati, OH, USA Conference Date: 3 May 1976 Publication Date: 1 Mar 1976 p 26 Report Number (s): HEDI SA-1073; CONF-760532-7 Contract Number (DOE): E(45-1)-2170 l

Document Type: Report; Conference literature '

language: English Joumal Announcemsnt: EDB7612 Availability Dep. NTIS S4.00.

Subfile: ERA (Energy Research Abstracts); INS (US Atomindex input); TIC (TechnicalInformation Cenier).

Country of Origin: United States Countryof Publication: United States i tstract: The ' hydrogen

• reduction kinetics of U/sub 0/1sub 75/Pu/sub 0/1sub 25.0/sub 2/ have been inwstigated at temperatures from 1000 to M00/sup 0/C and *h)drogen* flow rates not limiting to the reduction ste. Under these experimental conditions, reduction is controlled by l "he surface reaction rate which obeys a kinetic expression analogous to

( the Zeldovitch equation of chemisorption.

l 9

Teltecte* Literature Search Service Effect of Steam on Nucteer Fuel Componente i May 6,1997

. i

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FILE ' CHEMICAL ABSTRACTS' ENTERED AT 11:26:47 ON 06 MAY 1997 USE IS SUBJECT TO THE TERMS OF YOUR CUSTOMER AGREEMENT COPYRIGHT (C) 1997 AMERICAN CHEMICAL SOCIETY (ACS)

CHARGED TO COST =0377083 FILE COVERS 1967 -29 Apr 1997 (970429/ED) VOL 126 ISS 18 0 W ANSWER 1 OF 2 CA COPYRIGHT 1997 ACS AN 99:78794 CA Tl High energy proton interactions with strontium and cesium. A contribution to the waste transmutation study AU De Felice, P; Ocone, R.; Rindi, A.; Tuyn, J.; Dettenre, R.;

Roubaud, G.

CS Lab. Naz. Frascati, INFN, Frascati, Italy SO Nucl. Instrum. Methods Phys. Res. (1983),212(14),35945 CODEN:NIMRD9;ISSN:0167-5087 DT Journal LA English CC 71-11 (NuclearTechnology) ,

AB The prodn. was studied of stable and radioactive isotopes by irradn.

of 90Sr and 137Cs by high energy particle beams. A fast computer program was written which follows the yield of the various nuclides '

through successiw transmutation and decay processes. The cross sections for the high energy reactions were caled. using semiempirical equations. Samples of Sr and Cs were exposed to a 600 MeV p beam. The comparison of exptl. and theor. resuits confirms the w!idity of the semiempirical formulas as well as of tte transmutation code. This work is contribution to the studyof the scientific feasibility of using charged particle beams for transmutation, to reduce the radioactivity, of high-level

• • • nuclear *** waste materials.

l l

l l

10 Teltech' Literature Search Service Effect of Steam on hcieer Fuel Componente May 6,1997 i

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85/06/97 16:58:13 '-> 518 841 9774 P:s2 812  ;

O L35 ANSWER 1 OF 14 CA COPYRIGHT 1997 ACS '

. AN 124:129392 CA Tl Modeling the release behaviorof cesium dunng sewre fuel degradetica AU 14wis, B. J.; Andre, B.; Morel, B.; Debaudt, P.; Maro, D.; Purdy P.  ;

L; Cox, D. S.; Iglesias, F. C.; Osberae, M. F.; lorenz, R. A.

CS ~ Centre d' Etudes Nucleaires de Grenoble, Commissariat a l'Energie Atomique, Grenoble,38054, Fr.

SO J. NucL Mater. (1996),227(1&2),83109 CODEN:JNUMAM;ISSN:0022-3115 DT Joumal LA English AD An anal. model was developed to describe the diffusional release of fission-product Cs from Zirealoy< lad fuel under high temp. reactor accident conditions. The model is based on the anal. of recent I annealing expts performed at the Chalk River Labs. in *** steam ***

. The present treatment accounts for the influence of the atm.

(i.e., changirs O potential)on the state of fuel oxidn. and the ,

elease kinetics. The effects of fuel dissoin on the wlatile J release behavipr(under *** reducing *** conditons)is considered in terms of earliercrucible expts. and a simple model based on bubble coalescence and transport in metal pools. This model was also used to interpret the Cs release kinetics obsd. in

• • • steam *** and H expts. at the Vertical irredn. (VI) Facility in the Oa k Ridge National Lab. and at the HEVNVERCORS Facility.

IT ***10045-97-3*** , Cesium-137, processes RL FMU (Formation, unclassified); PEP (Physical, engineering or 1

chemical process); POL (Pbilutant); FORM (Formation, nonpyperatiw); OCCU (Occurrence): PROC (Process)

(modeling of the release behaviorof fission-product cesium )

during sawre fuel degrdn.)

i l

l l

l l

4 l

1 11 Teltesh' l.iterature Search Serse Effect of Steam on Bludeer Fuel C:n , : z ^e May 6.1997 1

~

85/86/97 16:58:53 -> 518 841 97/4 P na 813

.- e O L35 ANSWER 6 OF 14 CA COPYRIGHT 1997 ACS AN 113:179871 CA Tl Fuel behavior and fission product release under realistic

'"hWrogen"* conditions with comparisons between HEVA 06 test results and VULCAINcomputations j AU Dumas, J. M.; Lhiaubet, G.; Lemarois, G.; Ducros, G.

CS Commis. Energ. At., Fontenay aux Roses, F-92265, Fr.

SO Proc. Int. Cent. Heat Mass Transfer (1990),30(Fission Prod. Transp. ,

Processes React. Accid.),15342 )

CODEN: PCHTD4;ISSN: 0272-880X DT Joumal LA English AB in the anal. of core melt accidents occurring at low primary pressure, the VULCAIN eode predicts major "* steam"* starvation in the upper part of the core at a time when large amts, of fission products (F.P.) are released. The results of cornputations i.ndicate  ;

that the F.P. may thus escape into a strongly *** reducing" environment. This kind of phenomenon may play an important role in detg. the chem. behavior of F.P. and therefore influence the characteristics of the F.P. source. Because exptl. data on F.P.

release and oehaviorin *** reducing *** conditions were particularly deficient, a significant effort was devoted to the definition of a HEVA test involving a retrase of F.P. in H with realistic reactor accident conditions. The results of HEVA06 test indicated that under high "* reducing" conditions and at a temp. of 2100. degree., the measured releases of wlatiles are really lower than those predicted with current models, and Ba and Sr escapes were underpredicted by using the CORSOR'models. The results of F.P release kinetics, reevaluated with the EMIS module by introducing in the data base the new release rates measured at 2100. degree, during the HEVA 06 test, are inconsistent with the assumptions currently made that the rate of release of each element is proportional to the quantity remaining in the fuel and that the const. of proportionality depends on fuel tetnp.

IT *"10045-974"* , Cesium 137, properties ,

RL PRP(Properties)  :

(release of, from ** *n uc le a r *** reactor fuelelements under ]

realistic *"hWrogen*" conditions) i l

12 Toltech' Uterature Search Service Effect of Steam on Nucieer Fuel Coniponents May 6,1997 l

l 85/06/97 16:51:38 -> 510 841 9774 P:se 814

. e O L41 ANSWER 14 OF 14 CA COPYRIGHT 1997 ACS AN 88:29323 CA Tl Efficiency factor of a chemical *** nuclear *** reactor with gamma sources AU Anguis, T. C.

CS Esc. Super. Ing. Quim. Ind. Extr., Inst. politec. Nac., Mexico City.

Mex.

SO Report (1975), INIS-mf-3683,110 pp. Avail.: INIS From: INIS Atomindex 1977,8(19), Abstr. No. 332023 DT Report LA Spanish

'CC 71-8(NuclearTechnology)

Section cross-reference (s): 74 AB A chemonuclear reactor was simulated to calc. the efficiency factor of moi. species in chem. reactions induced by . gamma.-radiation, to obtain information for its design and to consider the electromagnetic energy as a possible soln. to the present problem of energy. The research is based on a math. model of successiw Compton processes applied to spherical and cylindrical geometry, for the radioisotopic sources 60Co [ "*10198-404)*** ] and 137Cs

[10045-97-3] relating the quantity of energy deposited in various cylinders with the G-value; the radius / height ratio of the reactor was optimized according to the mol. prodn. This phenomenon was illustrated with the radielysis of a soln. of MeOH/H2O which forms

"*H2"* and with the prodn. of EtBr which represents an

' industrial process inwlving radiation. The results show a greater deposition with 137Cs, but a larger prodn. of ***H2*" per h with 60Co. The cylinder with most advantage is that whose ratio R/H is 0.5, The final selection of the reactor should be made after a more intense study of the isotope used and the source activity. The j efficiency factor of *"H2*" can be increased by selecting the appropriate type and conen. of solute of the irradiated aq. solns.

O L4 ANSWER 3OF43 CA COPYRIGHT 1997 ACS AN 122:18598 CA Ti Oxidation of uranium in steam AU Haywa rd, P. J.; Evans, D. G.; Taylor, P.; George, I. M.; Duclos, A.

M.

CS Whiteshell Laboratories, Pinawa, Manitoba, Can.

SO J. Nucl. Mater. (1994),217(1 & 2),82-92 CODEN:JNUMAM;ISSN:0022 3115 DT Joumal LA English

[

CC 71-5(NuclearTechnology)

Section cross. reference (s): 78 AB The steam oxidn. rates of U-metal foil and ber stock were measured thormogravimetrically between 473 and 823 K Both materials exhibited Arrhenius-type behavior in their steam oxidn. rates at

<573-623 K, with activation energies of 46 9.+ 3.3 kJ moi-1 for U foil at <623 K and 58.0.+ . 3.8 kJ mol-1 for U bar stock at <573 E Howewr, a significant rate "*redn"* . was obsd. at higher temps. In all cases, the oxidn. products were uraninite-structured UO2-x compds. Non-adherent oxide layers were formed on bar samples below .apprx.573 K, becoming increasingly adherent at higher temps.

13 Toltech' Uterature Search Service Effect of Stoem on Nucteer Fuel Cornponente May 6,1997

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.- e 518 841 9774 P:ge 015

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Similarly, the oxide layers produced on foilsamples below j

.apprx623 K had highly porous cellular structures, whereas the '

layers formed at higher temps. were significantly more dense. For both materials, the decrease in oxidn. rate at >573-623 K was attnbuted to *** reduced *** steam ingress to the underlying l metalcaused by the increasingly protective surface oxide.

O L4 ANSWER 5OF43 CA COPYRIGHT 1997 ACS l AN 120:146690 CA Tl High temperature oxidation of UO2 in steam *" hydrogen ***

mixtures ,

i AU Abrefah, J.; de Aguiar Braid, A.; Wang, W.; Khalil, Y.; Olander, D.

R. i CS Dep. Nuct Eng., Univ. Califomnia, Berkeley, CA, USA )

SO J. Nuct Mater. (1994),20S(1-2),98-110 l CODEN:JNUMAM;ISSN: 00221115 DT Journal l t

LA English I

AB An exptt study of UO2 oxidn. in pure steam and in H20/Ar/ *"H2***

mixts. was conducted in a continuously recording thermogravimetric app. in the temp. range 1273-1623 K at one atm. svstem pressure.

Two surface reaction models were utilized in the anal. of the data:

the phenomenol model assumes the reaction rate to be proportional to the deviation of the oxygen conen. in the solid from the equil.

value established by the ambient gas. Anal. of the results show that the phenomenol model does not fit the exptl. data for pure steam, but the fit is good forH20/Ar/ "*H2*** mixts. The mechanistic modelis based on the rate-controlling step of water  ;

dissocn. at the solid surface. It provides a wry good fit to all of thqdata. As long as the pressure is 1 atm, either of the models l can be used to predict fuel oxidn. in severe accidents because the ambient gas invariably contains ***hy!rogen*** produced by Zircaloy corrosion. However, the kinetics of oxidn. in high-pressure steam "* hydrogen *** mixts. is still uncertain.

Evidence of uranium wlatilization and prefdrential etching at the grain boundaries at high temps. were obsd.

O L4 ANSWER 11 OF 43 CA COPYRIGHT 1997 ACS AN 116:138230 CA Tl Microstructures and phase relationships of crystalline oxidation products formed on unused CANDU fuel exposed to aerated steam and aerated water near 200. degree.C AU Taylor, Peter; Wood, Donald D.; Owen, Derrek O.; Hutchings, William G.; Duclos, A. Michael CS Whiteshell Lab., At. Energy Canada Ltd., Pinawa, MB, ROE ILO, Can.

SO At. Energy Can. Ltd., [ Rep.) AECL (1991), AECL-10476,71 pp.

CODEN: AECRAN;ISSN: 0067-0367 DT Report LA English AB This report considers the findings from dry, moist-and wet-air oxidn. expts. on unused UO2 fuel specimens at 200-225. degree.,

performed in support of the Dry Storage Program for used CANDU fuel.

The presence of liq. water, or unsatd. steam, adds to the complexity of air oxidn. of UO2. The following processes haw been identified by using a combination of optical and SEM and x-ray diffraction to 14 Toltech' Uterature Search Serwce Effect of Steam on Nuclear Fuel Componente May 6,1997

85/06/97 17:15t34 -> 518 841 9774 P:g2 816 detect oxidn. products, and are discussed in this report: (1) oxidathe dissoin. of U(VI) and pptn. of hydratc.d UO3;(2) back-

• • • redn*** . of dissolwd U(VI) and pptn. of 0308 on the UO2/U307 surface;(3) solid-state surface and grain-bouncery oxidn. of UO2 to

. beta.-U307; and (4) preferential dissoin. of UO2 grain boundaries, sometimes followed by the filling of the resulting gap with higher U oxide (s). Although moisture thus adds greatly to the variety of oxidn. reactions that can occur on UO2 surfaces, it does not appea r to promote swelling and spalling of the fuel, in spite of the large increase in molar wlume assocd. with formation of the hydrated phases. This conclusion is qualified, however, since variabilrty in the reactivity of fuel specimens, particularly with respect to grairvboundary oxidn., makes it difficult to distinguish moisture effects. With unused fuel, grain-bcundary alteration to U307 is the primary process by which oxidn. penetrates the fuel near 200. degree.. Reactions inw!ving water proceed on the specimen surface, and can also follow oxidized grain boundaries and (presumably) open raorosity.

O L4 ANSWER 22 OF 43 CA COPYRIGHT 1997 ACS AN 108-194409 CA Tl Properties of reactor fuel rod materials at high temperatures:

final summary report - sewre core damage property tests program AU Prater, J. T.; Courtright, E. L.

CS Battelle Pac. Northwest Lab., Richland, WA, USA SO Report (1987), PNIA164; Order No. NUREG/CR.4891,42 pp. Avail.:

NTIS From: Gov. Rep. Announce. Index (U. S.) 1987,87(23), Abstr. No.

755,198 DT Report LA English AB The report summarizes work sponsored by the U.S. Nuclear Regulatory Commission Division of Accident Evaluation to investigate those phys. properties that are needed to predict the behavior of fuel-rod assemblies during a lossofcoolant accident. The results include a detn. of the oxidn. kinetics of Zirealoyand Zirealoy-U oxide mixts.

in steam and steam-H gas mixts at 1300-2400. degree., viscosity measurements of Zr-oxide mixts. at 1800-2100. degree., an est. of the heat of reaction for the dissoln. of U oxide by molten Zr at 2000. degree., the thermal diffusivity measurements on prersacted Zirealoy-U oxide mixts. at 800-1500.deg ree..

CC 71-5(NuclearTechrology)

IT Nuclear reactors (accidents, kinetics of oxidn. of *** fuel *** rod material at high temp. in relation to)

IT ***Stea m***

(kiretics of oxidn. of Zircaloy and *** uranium"* oxide in presence of) l 15 Toltech' Uterature Search Service Effect of Steam on Nuclear Fuel Components May 6,1997

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) O L4 ANSWER 27 OF 43 CA COPYRGHT 1997 ACS

! AN 105:86927 CA Tl Oxidation of uranium dioxide by high-pressure steam AU Olander, D. R.

i CS Dep. Nucl. Eng., Univ. California, Berkeley, CA,94720, USA SO Nucl. Technol. (1996),74(2),215-17 CODEN: NUTYBB;ISSN:0029-5450 DT Journal LA English AB Thermodynamically, the O potential of pure steam increases as the pressure increases. As a result, high-pressure steam can produce more highly oxidized UO2 [1344-57-6] than can steam at atm.

l pressure. An O/U ratios as high as 2.60 can theor, be attained in steam at 150 atm and temps. near 1600 E. Oxidn. to this e.xtent can render the fuel nearly as important a source of H as the cladding in i sewre fuel damage accidents. Fuel oxidn. by steam, however, is endothermic and provides a heat sink rather than a heat source.

IT ***Stea m"*

(oxidn. of *** uranium *" dioxide by high-pressure) i 0 L4 ANSWER 29 OF 43 CA COPYRGHT 1997 ACS AN. 94:129058 CA Tl A model for release of fission gases and wlatile fission products 1

from irradiated uranium dioxide in steam environment  !

AU Cubicciotti, Daniel CS Electr. PowerRes. Inst., Palo Alto, CA,94303, USA SO Nucl. Technol. (1981),53(1),5-7 CODEN: NUTYBB;ISSN:0029-5450 DT Joumal l LA Eng4ish AB information conceming the release of fission gases and wlatile fission products from irradiated UO2 is important for modeling their behavior under accident conditions. There is evidence that fission

)

gas release is enhanced by the presence of, steam in the atm. in which UO2 is heated. Thus, fission gas release from a defected rod in a LWR under accident conditions would be larger than anticipated from data obtained on fission gas releases in intact fuel rods whose environment does not contain steam. The basis for the model is the well-established fact that the rate of sintering of UO2 is I significantly greater in a steam atm. than in an inert or

"* reducing"* atm. The model provides a basis forapplying data from intact fuel rod releases to defected rods and should be applicable to LWR accicent conditions.

IT "*Stea m***

(fission product re! ease from irradiated *" uranium *"

dioxide in, model for) l 16 l Teltech* 1.iterature Search Service l

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l 0 IA ANSWER 30 OF 43 CA COPYRIOHT 1997 ACS AN 93:139793 CA Tl Fluidized reactor "* reduction *** of uranium hexafluoride to uranium dioxide IN Welty, Richard K.

PA Exxon Nuclear Co., Inc., USA SO Cen.,10 pp.

CODEN: CAXXA4 Pi CA 1077680 800520 1 Al CA75-231248 750710 )

DT Patent l LA English  !

AB A 2-stage fluid bed process is described for conerting UF6 p783-81-5) to a sinterable UO2 [1344-57-6] of low fluoride conen.

Steam and UF6 are countercurrently introduced into contac't with a 1st fluid bed of seed particles of UO2F2 [13536844] solids at  !

658-680 degree. whersby UF6 is conarted to UO2F2 which is )

1 transferred to a 2nd fluid bed maintained at 730-780. degree.. The UO2F2 is conwrted to U308 [1344-59-8] by countercurrent contact with steam. The U308 solid is transferred to a 2nd reactor where it is *" reduced *" to UO2 by countercurrent contact with excess

• • • H2*** and steam at 450-500. degree.. The fluoride content of UO2 is <0.011% and the sinterability of the UO2 is enhanced.

IT *"Stea m***

( *"redn*" . of triuranium octoxide to *" uranium *"

dioxide by,2-staae fluid bed process for)

O L4 ANSWER 32 OF 43 CA COPYRIGHT 1997 ACS AN 87:13291 CA Tl Uranium dioxide IN Knudsen,Irving E.  !

PA Westinghouse Electric Corp., USA I

SO U.S.,8 pp.

CODEN: USXXAM l PI US 4020I46 770426 Al US 69443062 690718 DT Patent LA English i

AB A 3-stage pocess for conwrsion of UF6 into ceramic-grade UO2 powder is giwn. The process steps are:(a) reaction of H and steam a

with UF6 in a 1st fluidized bed at 475-600. degree. to form solid, intermediate products (UO2F2 and U308);(b) a 2nd reaction of the p

intermediate products with H and steam in a 2nd fluidized bed at 575-675.degros. to generate a 2nd group of products including UO2F2, U308, and UO2; and (c) a 3rd reaction of the 2nd group intermediate products in a 3rd fluidized bed at 575475. degree. to produce ceramic-grade UO2 powder.

IT *"Stea m*"

( *"redn*" . by, of *** uranium *" hexafluoride in fluidized beds, ceramic $rade *** uranium *" dioxide powder prepn. in relation to)

IT Nuclear reactor *** fuels *** and *** fuel *** elements

( *** uranium *" dioxide, manuf. of, by "* uranium ***

hexafluoride ***redn*" . in fluidrzed beds) i 17 Teltech' Lstorature Search Serwce Effect of Steem on phedeer Fuel Componente May 6,1997 I

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I O L4 ANSWER 34 OF 43 CA COPYRIOHT 1997 ACS AN 83:154236 CA

) Tl Mechanism and kinetics of the reaction of uranium hexafluoride with i *" hydrogen *" and steam in a fluidized bed i l

l AU Veryatin, U. D.; Yakhonin, l. F.

CS USSR SO Jad. Energ. (1975),21G),262 CODEN: JADEAQ l DT Journal LA Russian AB ~Ihe title reaction was carried out in a Ni vessel with a fluidized

! bed composed of 200-700.mu.m UO2 particles. Pyrohydrolysis was carried out at 650-700. degree.. The temp. during defluorination was 700-800. degree.. At high temps. in a fluidized bed, which provides i

rapid exchange of mass and heat of the reacting phases, the process can be described by UF6 + 2H2O + "*H2*** .dblarw. UO2 + 6HF. A kinetic equation is also giwn for the process. i IT Nuclear reactor *** fuels *** and *** fuel"* elements '

(prepn. of *** uranium *** dioxide, by reaction of

• • • ura nium*** hexafluoride with *"hydregen*** and

• • • steam *" in fluidized bed) 0 L4 ANSWER 37 OF 43 CA COPYRIGHT 1997 ACS l

l AN 74:48904 CA I Tl Uranium dioxide IN Bode, James A.

PA United Nuclear Corp.

, SO Ger. Offen.,35 pp.

! j CODEN:GWXXBX PI DE 1949962 700917 PRAIUS 681004 DT Patent LA German AB UO2 is made from UF6 by hWrolyzing UF6 with steam in a fluid bed of UO2F2 at 370. degree. (by controlling the ratio UF6/ steam, the formation of solid U fluorides is prewnted), defluorinating the thus obtained UO2F2 down to 3000 ppm F with steam and H in a fluid bed, and increasing the H conen. in the gas to obtain UO2 with <100 ppm F.

O L4 ANSWER 39OF 43 CA COPYRIGHT 1997 ACS AN 6992177 CA Tl High-temperature aqueous corrosion of rnetallic uranium fuel elements with defacted cadding AU Cerrai, E.; Scaroni, A.

CS Lab. CISE, Milan, Italy SO J. Nuct Mater. (1968),27(3),264-74 CODEN:JNUMAM DT Joumal .

LA English i

AB Metallic U fuel elements clad with Zircaloy-2 with appropriate defects were placed in a loop able to produce water-steam mixts.,15 wt. % steam, at 285. degree. end 71 atm. At the beginning of the i reacton in which superheated steam came in contact with U at i

18 Toltech' Uterature Search Serwce Effect of Steam on Nuclear Fuel Componente May 6 1997

85/86 6/ 17:18:36 -> 518 841 9774 .P:se 828 360. degree., U oxides, mainly UO2, and ***H2" were fornwd as final products, and the ***H2*** diffused slowlywithin the 2 adjacent metals and against the steam flow entering the hole. As a consequence of temp. and pressure cycling, the plugs of UO2 formed gradually lost their protectiw effect. 10 references.

IT Nuclear reactor *** fuels *** , reactions (corrosion of metallic *** uranium *** , with defect claddings in water. " steam" mixts.)

O L4 ANSWER 41 OF 43 CA COPYRIGHT 1997 ACS AN 69:40557 CA Tl Behaviorof fission products released into a steam-air atmosphere from omrheated uranium dioxide previously irradiated to 20,000 Mw.-days / ton AU Parsly, L. F.; Row, T. H.

CS Oak Ridge Nat. Lab., Oak Ridge, Tenn., USA SO U. S. At. Energy Comm. (1967), ORNL-TM-1908(Pt.1) 44 pp. Awi!.:

Dep.; CFSTl.

From: Nuct. Sci. Abstr. 1968, 22(5),10012 CODEN: XAERAK DT Report LA English AB The results of the 1st run made in the Nuclear Safety Pilot Plant (NSPP) using Zircaloy-2 clad UO2 are presented. Fuel was melted in a plasma under *** reducing *** conditions with a steam 4ir atm, in the model containment wssel. Behavior of the fission products, which were released, was detd. Fission product deposition on the reactor containment wssel was measured. $

IT Nuclear reactor *** fuels *** , properties (release of gaseous fission products from irradiated

• • • uranium *** oxide (UO2), in air "* steam *" atm.) '

O L4 ANSWER 43 OF 43 CA COPYRIGHT 1997 ACS AN 67:69795 CA '

Tl Behavior of fission products released from synthetic high-burn-up uranium dioxide in steam a tmospheres (Nuclear Safety Pilot Plant runs 10-12)

AU Parsly, L. F.; Row, Thomas H.

CS Oak Ridge Nati. La b., Oak Ridge, Tenn., USA SO U. S. A. E. C. (1967), ORNL-TM-1698 79 pp. Awil.: Dep. Mn; CFSTI l From:Nuct Sci. Abstr. 1967,21(9), 15661 CODEN:XAERAK '

DT Report LA English AB Mixts. of stable and radioactiw tracer isotopes of fission products simulating UO2 irradiated to high bum-ups were released by melting stainless-steelclad UO2 to which the simulants had been added. ,

. Data are included on fission product release and distribution, airborne fission product conen. as a function of time, collection of condensate and of fission products in the condensate, deposition on surfaces, tests of recirculating filter systems, field tests of remote sampling devices for the loss.of-Fluid Test, and other pertinent details of the experiments. The furnace atm. had a major effect on the release and transport of 12 and Ru and less significant effects on Sr, Ba, and Co. Under " reducing" i 19  !

Teltech' Uterature Search Service Effect of Steam on Huoleer Fuel Componente May 6,1997

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.. ~

l conditions the release of Ru and 12 was suppressed, and wry rapid deposition of the 12 occurred in the containment wssel. No significant conen. of Ru were found in containment wssel atm.

6 samples. The amts. of fission products found in the containment wssel atm. samples were always less than assumed in the TID-14844 accident analysis criteria.

IT *** Steam" , uses and miscellaneous (fission product release from irradiated " uranium *** oxide (UO2)into) l IT Nuclear reactor *** fuels *** , properties 0 IA ANSWER 5OF13 CA COPYRIGHT 1997 ACS

' AN 114:194348 CA Tl Data summary report for fission product release Test \1-3 AU Osbome, M. F.;Iorenz, R. A.; Collins, J. L.; Travis, J. R.;

Webster, C. S.

CS Oak Ridge Natl.14b.. Oak Ridge, TN, USA SO Report (1990), ORNI/rM-11399; Order No. NUREG/CR-5480,70 pp.

Awil.: NTIS From: Gov. Rep. Announce. Index (U. S.) 1990,90(17), Abstr. No.

043,975 DT Report LA English AB Test VI-3, the third in a series of high-temp. fission product release tests in the wrtical test app., was conducted in flowing steam. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in . Belgium, which had been irradiated to a bumup of 42 mwd /kg. Using an induction furnace, it was heated under simulated LWR accident conditions to 2 test temps.,20 min at 2000 K and then 20 min at 2700 K. The cladding was completely oxidized during the test and wry little melting or fuel-cladding interaction had occurred. Based on fission product inwntories measured in the fuel or caled. by ORIGEN2, analyses of test i

components showed total releases from the fuel of 100% for 85Kr,5%

l for 106Ru,99% for 125Sb, and 99% for both 134Cs and 137Cs. Small '

release fractions for many other fission products were detected. In addn., wry small amts. of fuel material U and Pu-were released.

The total mass released from the furnace to the collection system was 3.17 g,78% of which was collected on the filters. The results from the test were compared with previous tests in the series and with a commonly used model for fission product release.

IT 10045-97-3, Cesium-137, properties 13967-48-1, Ruthenium-106, properties 13967-70-9, Cesium-134, properties 13983-27-2, Krypton 45, properties 14234-35-6, Antimony-125, properties RL: PRP(Properties)

(release of fission product, from irradiated *** fuel *** under simulated LWR accident conditions)

IT ' 744047-5, *** Plutonium *** , properties 7440-61-1, Uranium, uses and miscellaneous )

RL PRP(Properties)

(release of, from irradiated *** fuel *** under simulated LWR l; conditions)

I 4

20 l Tottech' Uterature Search Service Effect of Steam on hcieer Fuel Components May6,1997

85/86/97 17:28:14 -> 518 841 9774 hs2 822 O IS ANSWER 13OF 13 CA COPYRIGHT 1997 ACS AN 66:110579 CA Tl Water and steam corrosion characteristics of mixed plutonium-uranium oxide pellets AU Marley, Jack L.

CS Nucl. Maters and Equipment Corp., Apollo, Pa., USA SO AEC Access. Nos. (1966), NUMEC-3463-13,54 pp. Avail.: Dep. mn; CFSTi, S3 cy From: Nuct. Sci. Abstr. 1967, 21(2), 1927 q

CODEN: AECAA6 15 DT Report LA En0 lish AB Corrosion tests in neutral, degassed water, and steam and in mndified media were made with pellets of PuO2-UO2. The pellets which contained 1,5, and 15 wt. % PuO2 were prepd. byco-pptn. and by mech. blending of the mixed oxides. Substoichiometric and superstoichiometric pellets were propd. for the tests. No significant differences in corrosion behavior were obserwd in any of the tests. Pellets which were contained in intentionally defected Zircaloy-2 tubes were as corrosion resistant as pellets which were tested individually.

O L14 ANSWER 2 OF6 CA COPYRIGHT 1997 ACS AN 124:69190 CA TI The first PHEBUS fission product experiment FPTO. General aspects of the experimental sequence concoming the fuel bundle degradation and FP release AU Repetto, Georges: Von Der Hardt, Peter; Gonnier, Christian; Haess!er, Maurice CS CE-Cadarache France, institut de Protection et de Surete Nucleaire (IPSN), St-Paul-Lez-Durance,13108, Fr.

SO Proc. Top. Meet. Saf. Oper. React. (1995),529-36 Publisher.

American Nuclear Society La Grange Park, Ill.

CODEN: 61XYAW ,

DT Conference LA English AB The PHEBUS fission products program is a LWR source term research program initiated by IPSN and the Joint Research Center of the European Commission. The objectrve is to study, during a PWR sewre accident, the degrdn. phenomena of a bundle of fuel rods and the

behavior of the fission products released. The first test, FPTO, was performed in Dec.1993. Objectiws were fulfilled and show a core degrdn. progression in the late phase far beyond any other integral effect expt. that has been performed up to now

PBF SFD, PHEBUS CSD, CORA, FLHT, LOFT-FP-2.

! iT 10198-40-0, *** Cobalt *** -60, occurrence 13967-71 4, l

Zirconium-95, occurrence 14391-76-5, Silwr-110, occurrence RL: OCU (Gccurrence, unclassified); OCCU (Occurrence)

(LWR PHEBUS fission product expt. FPTO and general aspects of the expt!. sequence concerning *** fuel *** bundle degrdn. and fission product release in relation to) 21 Teaterdi' Literature Search Service Effect of Steem on lepoleer Fuel Componente i

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i O L14 ANSWER 6 OF 6 CA COPYRIGHT 1997 ACS AN 106:184535 CA l

TI Corrosion products in the W%TR-40 primary coolant circuit AU Venz, H.

j CS VE Kombinat Kemkraftwerke "Bruno Leuschner" Greifswald, Greifswald, i

DDR-2200, Ger. Dem. Rep.

l SO Kernenergie(1987),30(1),318 CODEN: KERNAQ: ISSN: 0023-0642 l DT Journal IA Oerman AB Anal. of the deposited materialin the primary coolant circuit in

• "nuc!aar*" reactors of type WWER-440 showed that high corrosion promotes the contamination. In the region of the steam generators the metal ions formed by the corrosion process are transformed into colloids, which are transported with the coolant and deposited on the surface of the fuel elements. The out of-core areas were mainly contaminated by the solutes which were incorporated in the corrosion layers.

IT 10198-40-0, *** Cobalt *** 60, occurrence 13966-31-9, Manganese 54, occurrence 13981-389, *** Cobalt *** 58, occurrence 14392424, Chromium 51, occurrence 14596-12-4, fron 59, occurrence RL: OCCU (Occurrence)

(in corrosion product contaminant in primary cooling system of WWER-440)

O L17 ANSWER 5 OF 7 CA COPYRIGHT 1997 ACS AN 100:181830 CA TI Thermal stabiltty of cesium uranate (Cs2UO4) phase at high temperature AU Dharwadkar, S. R.; Shyamala, M.; Chattopadhyay, G.;

Chandrasekharaiah, M. S.

CS Chem. Div., Bhabha At. Res. Cent., Bombay,400 085, India SO Trans. Indian inst. Met. (1983),36(4-5),295-7 CODEN: TilMA3;ISSN:0019493X DT Journal LA English AB The high-temp. thermodn. properties of the Cs-U4) system haw considerable practical applications, in the anal. of oxide fuel performance of *** nuclear *** reactors. The thermaland thermodn. stability data were crit, re-ewluated, and sewre inconsistencies were found. For example, the reported sequence of Cs uranates in air is inconsistent with the free enthalpy data reported for these phases. Therefore, a systematic reinwstigation of the thermodn. stability of Cs uranates was initiated. Sewral rnethods of p.epg. pure, monophasic Cs2UO4(s) were attempted, but the hygroscopicity of the product necessitated in-situ generation of the Cs2UO4 before such property measurement. It was exptl. demonstrated that Cs2UO4 is sta ble in dry air or Ar up to 1225 K, which is in complete disagreement with the results of E. H. P. Cordfunke, et al.

(1975). Transformation of Cs2UO4 to Cs2U207 is possible at 900 K, l

only if some moisture (water pressure = 10-4 atm) is also present.

This role of moisture in the reaction of Cs2UO4 was shown exptl.;

the reaction is: 2Cs2UO4 + H20(g) = Cs2U207 + 2CsOH(g).

22 l , Toltech' Uterature Search Serwce Effect of Steam on Nucieer Fuel Componente May 6,1997

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0 L17 ANSWER 6 OF 7 CA COPYRIGHT 1997 ACS AN 92:205542 CA Tl Fission product release from LWR fuel defected in steam in the temperature range 500 to 1600. degree.C AU Lorenz, R. A; Collins, J. L.; Osborne, M. F.; Malinauskas, A P.

CS Oak Ridge Natl. Lab., Oak Ridge, TN, USA SO Report (1979), CONF-790935-3,9 pp. Awit.:NTIS From: Energy Res. Abstr. 1979,4(24), Abstr. No. 56390 DT Report LA English AB For practical purposes, the release of Cs and I from L%3 fuel rods in the temp. range 500 to 1600. degree. originates from 3 sources:  !

(1) the gap inwntory with release by both burst (vented gas) and diffusion; (2) grain boundary with release by tunnel fomsation; and (3) UO2 matrixwith release by solid state diffusion. The chem.  ;

j behavior of released I and Cs (at least after contact with steam) is 4 predominantly that of Csl [7789-17-5) and CsOH [21351-791].

Fission gas release is the sum of the plenum inwntory, gas embedded in the fuel and cladding surface layers and that released by tunnel formation, and solid state diffusion from the UO2 matrix. A small amt. of large particle-sized fuel dust is ejected at time of rupture.

IT "**Stea m***

I (fission product release to, in LWR)

O L17 ANSWER 7OF7 CA COPYRIGHT 1997 ACS AN 92:49003 CA Tl Fission product source terms for the light water reactor loss <f-coolant accident AU Lorenz, R. A; Collins, J. L.; Malinauskas, A P.

CS Chem. Techno!. Div., Oak Ridge Nati. Lab., Oak Ridge, TN, 37830, USA SO Nucl. Technol. (1979),46(3),404-10 l CODEN: NUTYBB;ISSN: 0029-5450 DT Journal l LA English AB Models for Cs and I release from LWR fuel rods defected in steam were formulated, based on exptl. fission product release data from several types of defected LWR fuel rods. The models were applied to a PWR undergoing a loss <f<oolant accident temp. transient. Caled.

hI totalI and Cs releases were 0.053 and 0.025% of the total reactor inventories of these elements, resp., with most of the release l occurring at the time of rupture. These values are .apprx.2 orders l of magnitude less than those used in WASE1400, the Reactor Safety l Study.

IT *"Stea m"*

l (release of fission products from reactor *" fuel *" rods defected in, model for)

I l

l l

23 Toltech' Uterature Search Service Effect cf Steam on feucdear Fuel Componente May 6,1997 l

05/06/W 17:22:37 -> 518 841 W74 P;g2 025 O SEARCH STRATEGY IN CA FILE 'CA' ENTERED AT 12:48:19 ON 06 MAY 1997 L1 65544 SEA URANIUM /IT I2 11180 SEA L1 AND FUEL'AT L3 98 SEA L2 AND STEAM?/IT,ST L4 43 SEA L3 AND (REDUC? OR REDN OR REDOX OR HYDROGEN OR H2 OR 1333-744)

L5 13 SEA PLUTONIUM /IT,ST AND FUEL?/IT,ST AND STEAM?nT.ST IS 13 SEA L5 NOT L4 L7 36 SEA STRONTl?/IT,ST AND FUEL?/IT,ST AND STEAM?M,ST IB 16 SEA L7 AND (REDUC? OR REDN OR REDOX OR HYDROGEN OR H2 OR 1333-744)

L9 14 SEA L8 NOT(L4 ORIE)

L10 22 SEA L7 NOT L9 L11 3 SEA L10 AND NUCLEAR?

L12 71 SEA L4 ORIE OR L9 OR L11 L13 77 SEA COBALT /IT,ST AND FUEL?/IT,ST AND STEAM?/IT,ST L14 6 SEA L13 AND NUCLEAR?

L15 19 SEA CESIUM?MT,ST AND FUEL/IT,ST AND STEAM?/IT,ST L16 15 SEA L15 AND NUCLEAR?

L17 7 SEA L16 NOT (L12 OR L14)

L18 82 SEA L12 OR L14 OR L17 SAVE L18 MAYTGAI/A O SEARCH STRATEGY IN SCITECH DATABASES Set items Description SO 151 *REEP*

S1 2056055 (URANIUM OR PLUTONIUM OR STRONTIUM OR COBALT OR CESfUM OR U OR PU OR SR OR CO OR CS)

S2 230181 S1 AND NUCLEAR?m.DE,lD S3 2740938 (STEAM? OR HYDROGEN OR fl2 OR REDUC? OR GAS OR GASES OR GAS-EOUS)m,DE,lD .

S4 37930 S2 ANDS 3 SS 23015 S4 AND (HYDROGEN OR H2 OR REDUC? OR STEAM?)m,DE,lD S6 5227 SS AND (HYDROGEN OR H2 OR REDUC' OR STEAMNil 1

S7 92 S6 AND (STRONTIUM OR SR)

S8 85 RD (unique items) i S9 l 19 S8 AND (STROhTIUM OR SR)m i S10 5971 S5 AND FUEL? '

S11 1466 S10 ANDSTEAM?

S12 25 S11 AND (STRONTIUM OR SR)

S13 25 S12 NOTSO S14 22 RD (uniqus items)

S15 257102 S3 AND (NUCLEAR? OR REACTOR)m,DE.lD S16 56115 S15 AND FUEL'm,DE,ID S17 14517 S16 AND (URANIUM OR UR)m,DE,lD S18 6586 S17 AND (STEAM? OR HYDROGEN OR H2 OR REDUC')m,DE,lD S19 1200 S18 AND (STEAM? OR REDUC?)m S20 1200 S19NOTSO S21 32 S20 AST)(URANIUM 00XIDEORUO2)

S22 28 RD (unique items)

S23 10 NUCLEAR 0 FUEL'(L) REDUC?

24 Tottech' Uterature Search Serwce Effect of Steam on insclear Fuel Componente May 6,1997 l

l

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I S24 10 S23 NOTSO S25 10 RD(uniqueitems)

S26 3550 S16 AND (PLUTONIUM OR PU)m.DE.lD NOTSO l S27 268 S26 AND (STEAM? OR REDUC?)m t l S28 6 S27 AND (PLUTONIUM 00XIDE OR PUO2) t S29 6 RD(uniqueitems) l S30 121 S27 NOTS 19 S31 262 S27 NOTS 29 ',

S32 130 S31 AND (PLUTONIUM OR PU)m S33 18 S32 AND(HYDROOENORH2)

S34 17 RD(uniqueitems)

! S36 244 S31 NOTS 33 s' l S36 14 S35 AND PLUTONIUM 0 DIOXIDE S37 13 RD(un%ueitems) i S38 1084 S16 AND (COBALT OR CO OR 60C0 OR CO60)m,DE,lD S39 455 S38 AND OIYDROOEN OR H2) i S40 83 S39 AND(STEAM' OR REDUCS)m,DE ID S41 83 S40 NOTSO S42 82 RD(uniqusitems) l St3 i 38 S42 AND (FUEL? OR COBALT OR CO OR SOCO OR 60CO)m i S44 32 S42 AND (COBALT OR 60C0 OR CO60) NOTS 0 l

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25 l Teltech' Uterature Search Service Effect of Steam on Nuclear Fuel Componente May 6,1997 l f

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a r3 8 TELTECH *"

...the knowledge to compete **

E Network of Experts - Confer with technical experts who can answer your questions.

E Uterature Searching - Find mission-critical technical and business information in print.

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If transmission is incomplete, oryou wish to use any of Teltech's other knowledge services, please call 16043678358.

I l

l DATE i05/07/97 TIME  : 15:02:28 TO  : Ms. Amy Schumacher l

FAX  : 1509-735-6920 FROM  : Tina Mattia HESSAGE : i l

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1 O L24 ANSWER 9OF20 CA COPYRIGHT 1997 ACS AN 112:147508 CA Tl Release of wlatile fission products from irradiated LWR fuel: mass spectrometry studies: final report l AU Johnson, l.; Johnson. C. E.

CS Electr. Power Res. Inst., Palo Alto, CA, USA SO Report (1989), EPRI-NP-6328,75 pp. Avail.: RRC, P.O. Box 50490, Palo Alto, CA,94303 6

From: Energy Res. Abstr. 1989,14(13), Abstr. No. 26552 DT Report -

LA English N b @h l

AB Quadrupole mass spectrometry has been used to study the chem. form and release rate of wlatile fission products effusing from an l irradiated LWR fuel sample. Expts, up to a temp. of 1900. degree. \- Ce

' have identified Kr, Xe, Cs, and Te as the species released from the fuel. A weak signal for at. I was obsd. at 1050. degree.. Release d

j $

rates were measured under reducing conditions, as would be expected -

! when an excess of gaseous H is present, and oxidizing conditions, as would be expected in a pure *** steam *** environment. The main I) conclusions from these studies are (1) the release rates of Kr and j Ke from irradiated fuelare less than the previous ests. by 03 I

(Eidizing conditions) to 3 (reducing conditions orders of T

l rnagnitude, (2) the release rates of CQ, anhre significantly C '

l lower than those of Kr and Xe; underoxidizing conditions, the I

l release rate of Cs is zero, (3) the low release rates of Cs and I

! i I

reduced the conen. of Csl below the detection limit of the mass spectrometer. In complementary expts, on nonradioactiw material, ,

the release of To was hindered

' dissoln. in the Zircaloy cladding. l Above 1300. degree., gaseous nT as obsd.; its formation is attributed to reaction of the Te with the Sn in the Zircaloy.

hk, l

IT ***1333-74-0*** , *** Hydrogen *** , properties l RL: PRP (Properties) l (fission product release from irradiated LWR *** fuel *** in '

presence of excess gaseous, mass spectrometry studies of)  !

l IT 7439-90 9, Krypton, properties 7440-46-2, *** Cesium *** ,

l properties 7440 63-3, Xenon, properties 7553-56-2, lodine, properties 13494-80-9, Tellurium, properties l RL: PRP(Properties)

(release of fission product, from irradiated LWR *** fuel *** ,

j mass spectrometry studies of)

! O L24 ANSWER 11 OF20 CA COPYRIGHT 1997 ACS AN 109.81582 CA TI Chemical state of tellurium in a degraded LWR core AU Imoto S.; Tanabe, T. .

CS Fac. Eng., Osaka Univ., Osaka,565, Japan SO J. Nucl. Mater. (1988),154(1),62-6

! CODEN: JNUMAM: ISSN: 0022-3115

! DT Journal IA English AB Changes of the chem. state of Te in the heatup stage of a sewre fuel damage accident were estd. thermodn. According to equil.

calens. with the SOICASMIX-PV code, Te exists as Cs telluride, as 4 the element or possibly as PdTe dunng normal operation, in the 16 Toltech' Literature Search Service Effacts of Steam on Nuclear Fuel Components: Part 2 May 7,1997 i

i m Y 07 '97 13:30 i

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. . . . . .uasueiar As.e.A.Au -r bt19 (.15 bV4tl P:ge Big 1

l heatup stage of an accident, elemental Te is absorbed in the Zirealoy cledding by formation of ZrTex (x = 1-2). Ce telluride does not react with Zr own under the low 0 potentials fawring the Zr/UO2 reaction. Te is also absorbed in 0-stabilized .alpte.-Zr. ,

. The stability of Cs2Te in the *** steam *** /H atm. is discussed.

IT 778349-7, *** Hydrogen *** te!!uride (H2Te) 12037-94-4, M

Palladium telluride (PdTe) 12191-06-9, *** Cesium *** telluride (Cs2Te) 32321-654, Zirconium ditelluride 35065-56-6, Zirconium telluride (ZrTe) 39294-10-5, Zirconium telluride (ZrTe3)

RL PROC (Process)

(in LWR degraded core) l l

l l

l l.

l' i

i l

1 l

1 l

17 Toltech' Uterature Search Service f Effects of Steem on Nuclear Fuel Components: Part 2 l May 7,1997 i

l MRY 07 '97 13:30 por:e to

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\ 'l O L24 ANSWER 17 OF 20 CA COPYRIGHT 1997 ACS AN 89:187636 CA TI Chemical thermodynamics of the system *" cesium ***

-uranium-zirconium *** hydrogen *** -iodine-oxygen in the light water *** reactor *" fuel-cladding gap AU Besmann, Theodore M.; Lindemer, Torrence B.

CS Chem. Technol. Div., Oak Ridge Natl. Lab., Oak Ridge, Tenn., USA SO Nucl.Technol. (1978),40(3),297-305 CODEN: NUTYBE; ISSN: 0029-5450 DT Journal IA English AB Equil. thermodn. calens. were performed on the Cs-U-Zr-H-i-O system that is assumed to exist in the fuel <ladding gap of light water

• " reactor" fuel under in *" reactor *** , " steam"* ,

and 50% *** steam" -50% air conditions. The in *** reactor ***

O potential is assumed to be controlled by either UO2+x + Cs2UO4 or i Zr + ZrO2. The important condensed phases in *** reactor"* are g UO2+x, Cs2UO4, and Cst, ar'd the major gaseous species are Cs, Cs2, T Cst, and Cs212. The presence of "* steam *** does not alter It these species, sithough CsOH also beenmes a major gaseous species.

In a 50% "* steam *** -50% air mixt., the equil. condensed phases are U308 or UO3 and Cs2U15046. Under a nonequil. situation where Zr metal can react with 1, Zr13 or liq. Zr:2 is present, a nd the gascous species Zrl3 and Zrl4 haw large partial pressures.

IT "*Nuclea r*** ***reac to r*** *** fuels *** and *** fuel *"

elements

( *** cesium *** -uranium-zirconium *** hydrogen ***

-iodineoxygen system chem. thermodn. in gap between cladding anB)

O L24 ANSWER 18 OF 20 CA COPYRIGHT 1997 ACS AN 89:81725 CA TI Chemical thermodynamics of the system *** cesium"*

-uranium-zirconium *** hydrogen *** -io' dine-oxygen in the LWR fuel < lad gap AU Besmann, T M.;Lindemer, T. B.

CS Oak Ridge Natl. Lab., Oak Ridge, Tenn., USA SO Report (1978), ORNUTM-6130,25 pp. Amil.: NTIS From: Energy Res. Abstr. 1978,3(11), Abst. No. 26211 DT Report IA English AB Equil. thermodn. calens. were performed on the Cs-U-Zr-H40 system that is assumed to exist in the fuel < lad gap of light water

• • • reactor *** fuel under in *** reactor *** , "* steam"* ,

and 50% "* steam" -50% air conditions. The in. "* reactor"*

O potential is assumed to be controlled by UO2+x rather than Zr +

ZrO2. Thus, the important condensed ph?ses present are UO2+x, Cs2UO4, and Cst, a nd the major gaseous species are Cs, Tand Cs212. The presence of *** steam *** does not alter the species present,although CsOH also becomes a maior caseous soecies. In a 50% *** steam" -50% air mrxt., the condensed phases U308 or UO3, Q $

Cs2U15046, and Zrl3 or liq. Zrl2 are present at equil., and the gaseous species Zr!2 Zr!3, and Zrl4 haw large partial pressures.

IT "1333 74 0*" , properties 7440-46 2, properties 7440-61-1, g ~,[j dh Y

18 Toltech' Literature Search Service Effects of Steam on Nuclear Fuel Componente: Part 2 May 7,1997 f%Y 07 ' 97 132 3t oom ' o

u uun as ss.u. 22 > dos rs3 b34n F ge tit.tf properties 7440-67 7, properties 7553-56-2, properties 7782 44-7 properties RL: PRP(Proporties)

(chem. thermodn. of system in *** nuclear *** ***reac tor ***

• • • fuel" -clad gap contg.)

O L28 ANSWER 1 OF33 CA COPYRIGHT 1997 ACS AN 124:69145 CA Ti Application of *** hydrogen"* waterchemistry to moderate

• • • corrosiw" circumstances around the "* reactor" pressure wssel bottom of Boiling Water " Reactors"*

AU Uchida, Shunsuke; lbe, Eishi; Ohsumi, Katsumi CS Energy Research IAboratory, Hitachi, Ltd., Ibaraki, Japan SO Chim. React. Eau Actes Conf. Int. (1994), Volume 2,394-7 Publisher:

Societe Francaise d'Energie Nucleaire, Paris, Fr.

CODEN: 61SMA5 DT Conference IA English AB Application of ***hWrogen*** waterchem, to moderate corrosiw circumstances is a promising approach to preserw structural integrity of major components and structures in the primary cooling system of BWRs. The benefits of HWC application are usually accompanied bysewral disadvantages. Afterenlucting merits and demerits of HWC application, it is concluded that optimal amts. of

• • • hydrogen"* injected into the feed watercan moderate corrosiw circumstances, in the region to be presened, without serious disadwntages.

IT "1333-744*** , *" Hydrogen *** , uses RL: NUU (Nonbiological use, unclassified); PEP (Physical, engineering or chemical process); TEM (Technical or engineered meterial use); PROC (Process); USES (Uses)

( " hydrogen *" water chem. application to moderate

• • • corrosiw'** circumstances around the BWR pressure wssel bottom) i l

l 19 Teltech' Literature Search Service Effects of Steam on Nuclear Fuel Components: Part 2 May 7,1997

• # ?__?_?_________ __

~

ussu0 as su.s.c.ns -> un's t.13 kun V:g3 klZ1 x

O L28 ANSWER 5 OF 33 CA COPYRIGHT 1997 ACS AN 121:165623 CA TI The role of ZrxlyC compounds in minimizing stress ***corrcsion*"

cracking in fuel cladding AU Chan, P. K; frving, K G.; Mitchell, J. R.

CS System Chem. Corrosion Branch, Chalk Riwr Laboratories Chalk River, ON, K0J IJ0, Can.

SO Conf. Proc, . Int. Conf. CANDU Fuel Perform., 3rd (1992), 7-24/1-40.

Editor (s): Boczar, Peter G. Publisher: Can. Nucl. Soc., Toronto, Can.

CODEN: 60DHAC DT Conference IA English AB Some evidence was found that stable ZrxtyC and CsZrxlyC compds. can be formed in samples of oxidized CANLUB-coated Zircaloy-4 cledding exposed to 12 and Cal, resp., at 320. degree.. Powder samples obtained from the surfaces of the irradiated CANLUB-coated fuel cladding were also examd. by XPS. The results confirmed the authors' previous finding that the deposited I, in an irradiated element (BDL-406-AAH), was present as a stable metal iodide

((Cs)Zr-14) cluster compd. This was shown by correlating Zr 3d,1 3d and C 1s binding energies with those obtained from well-characterized Zr6112C stds. The vapor pressure of I abow the cluster compds., measured by a mass spectrometer at 10-8 torr, did not change with temp, up to 325. degree., l'reliminary kinetic expts., to measure the reaction rate of 1311 with CANLUB, indicated that CANLUB<oated Zircaloy-4 cladding can bind chem. with 1311 to {

form stable compd.(s)(such as Zr-I-C) after.apprx.35 ***h"*

at 320. degree.. These are important findings in the program to det.

the role of (Cs)Zr-1-C compds. in the sheath-protection mechanism afforded by CANLUB.

O L28 ANSWER 6OF33 CA COPYRIGHT 1997 ACS AN 121:67757 CA j

Tl The effect of gaseous Cs-Te mixtures on nickel AU Pulham, Richard J.; Richards, Martyn W. ]

l CS Chemistry Department, University of Nottingham, Nottingham, NG7 2RD.UK SO J. Nucl. Mater. (1994),210(1-2),30-3 CODEN: JNUMAM;ISSN: 0022-3115 DT Journal IA English AB Ni capsuies which were used to contain Cs-Te mixts with Cs/Te ratios of 1:1, at 948 K for 168 ***h*** during corrosion tests on *** nuclear"* fast *** reactor *** fuelcladding materials were themselves sewrely embrittled by the vapor of the !iq. mixts.

The embrittlement is attributed to traces of the gaseous species y CsTe(g)and CsTe2(g)which penetrate the Niintergranularly. The I driving force comes from the reaction CsTe2(g) + CsTe(g) + 3Ni(s) hkV d"2.

.fwdarw. Ni3Te2(s) + Cs2Te(s), for which . DELTA.rH [Cs(g). Te(g), hMD Ngg Ni(s),948 K] = -515 kJ/mol. The vapor abow certain Cs-Te liq. g Q{ttEft(gCA W h.

mixts, may be as chem. deleterious to cladding alloys as are the liqs. themselvss.

4 20 Yeltesh' Uterature Search Service Effects of Steam on Nuclear Fuel Componente: Part 2 May 7,1997

I l

0 L28 ANSWER 11 OF 33 CA COPYRIGHT 1997 ACS AN 117:78473 CA Tl Release of fission products and actinides from used CANDU fuel fragments under *"raducing*** conditions at 100. degree.C AU Stroes-Gascoyne, S.; Johnson, L. H.; Se!!inger, D. M.; Wilkin, D. L.

CS Whiteshell Lab., Pinawa, MB, ROE ILO, Can. '

SO At. Energy Can. Ltd., [ Rep.) AECL (1992), AECL 10574. 43 pp.

CODEN: AECRAN;ISSN: 0067-0367 DT Report LA English AB Used fuel leaching under rnildly reducino conditions (Ar.3% i

• • • H2*" atm.) at 100. degree. was sTatted using unoxidized fuel fragments and 2 leaching soins.: distd. deionized water (DDH2O) and Std. Canadian Shield Saline Soln. (SCSSS, I = 1.37). Conen. lewis ([gggtr(p.

of 137Cs,90Sr,99Tc,238U,239+240Pu, and 238Pu/241Am were 1-2 '

orders of magnitude higher in SCSSS than in DDH20. Differences in k-1 bicarbonate conen., ionic strength, and radiolysis effects are considered as possible explanations for these obserwtions. The use  !

of 90Sr conens. as a measure of fuel matrix dissoin. and the congruency of actinide dissoln. are discussed with respect to the origin (i.e., gap, grain boundaries or matrix) and the dissoln.

behavior (soly, and dissoln. kinetics) of radionuclides iri soln.

IT "*1333-74-0*" , "* Hydrogen"* , reactions

RL

RCT(Reactant)

( ***redn*** . by, in actinide and fission product release from i

l used CANDU *** fuel *" fragments at high temp.)

l l 0 L28 AtjSWER12OF33 CA COPYRIGHT 1997 ACS AN 116:160863 CA TI Chemical reactions of "* cesium *** , tellurium and oxygen with

fast breeder *** reactor *** cladding alloys. Part V-The

"* corrosion"* of 12R72HV steel by ,*" cesium *** telluride i

! (Cs2Te) under constant oxygen potentials '

AU Pulham, R. J.; Richards, M. W.

CS Chem. Dep., Univ. Nottingham, Nottingham, NG7 2RD, UK SO J. Nucl. Mater. (1992),187(1),39 42 j

CODEN: JNUMAM;ISSN: 0022-3115 DT Journal l IA English 9 0 01314 l AB The depth and nature of the corrosion of 12R72HV steel by C at t 948 K after 168 "*h*" in sealed capsules was detd. The opth of corrosion increases from 60 to .gtorsq.140.mu.m upon increasing k

the 02 potential from . DELTA..hivin.GO2 = -417 (Mo/ moo 2 buffer) to 307 kJ moi-102 (Ni/NiO buffer). At both 02 potentials, the depth l

of corrosion of alloys increases in the order PE16 < M316 < 12R72HV,

! The corrosion took the form of alternate dark and light colored layers. Anal. of the layers by SEM/EDAX was consistent with the '

dark layers being the compds. Cs3CrO4, Cr2Te3, and Cr203 sepd. by j light layers of Fe/Ni. This type of corrosion is common to all 3 i

alloys (12R72HV, M316 and PE16) but is pa rticularly well illustrated with 12R72HV.

21 Tottech' Uterature Search Service Effects of Steam on Nuclear Fuel Components: Part 2 May 7,1997 l%Y 07 '97 13:33 o" "  !

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L28 ANSWER 14 OF33 CA COPYRIGHT 1997 ACS AN 114:90278 CA Tl Fission product release and fuel behavior of irradiated light water

• • • reactor *** fuel under sewre accident conditions: the ST-1 Experiment AU Allen, Michael D.; Stockman, Harlan W.; Reil, Kenneth 0.; Grirnley, Arthur J.

CS Sandia Natl. Lab., Albuquerque, NM,87185, USA SO Nucl. Technol. (1990), 92(2), 214-28  :

CODEN: NUTYBB;ISSN: 0029-5450 DT Journal LA English AB High burnup UO2 *" reactor" fuelwas hcated in-pile at

.apprx.2490 K in a redmina atm, (3% "*H" in Ar) for 16 min.

Fission-product aerosols and vapors released from the fuel were collected on a series of sequentially opened filters; the fractions of the original fuelinwntory collected on the filters were fCs u 0.56, il = 0.38, fBa = 0.078, fSr = 0.053, fEu = 0.064, and fTe <

i 0.002. The measured release rates for nonwlatile fission products were much higher than those predicted by existing release codes, whereas the Te release rate was much lower. Post-test examn. of the gy)C6 fuo! indicates extensiw fuel / clad interaction, fuel swelling, and g60C fcp infi!tration of the fuel by a Zr-rich metallic melt; this melt kept

{ fgM b

O potentials in the fuel wry low. The low 0 potentials and fuel disrrption may account for the discrepancy between release codes and pqr# ~ '

the test release results.

IT ***1333-74-0*** , *** Hydrogen *" , properties RL: PRP (Properties)

(fission product release and *" fuel" behavior of irradiated LWR *** fuel *** under sewre accident conditions in

• • • roducing*** atm. of) l l

l i

l 1

22 Toltech' Literature Search Service Effects of Steam on Nuclear Fuel Components: Port 2 May 7,1997 l

11rW 07 '97 13: 33 "0" " t

-. ~ --- . _ _ . . . . .. . .- . . ~ _ ~ .

Cui u s / U s 13 4 ct .'t's -/ bCV (.53 bV4D Rge tjZ4 4-4 O L28 ANSWER 15 OF 33 CA COPYRIGHT 1997 ACS AN 113:140510 CA TI Chemical reactions of *** cesium *** , tellurium and oxygen with fast breeder *** reactor *** cladding alloys. Part IV- the j

• • • corrosion *** of ferritic steels AU Pulham, R. J.; Richards, M. W.

CS Chem. Dep., Univ. Nottingham, Nottingham, NG7 2RD, UK 5

SO J. Nucl. Mater. (1990),172(3),304-13 CODEN: JNUMAM;ISSN: 0022-3115 DT Joumal LA English AB A study of the corrosion of the steels FV448 and DT2203Y05 by CsKe mixts. in sealed capsules contg. O buffers at 948 K after 168h shows that the oxide dispersion-strengthened steel is more resistant to corrosion than is FV448 at Cs6e ratios of 1:1 and 2:1. ]

Both steels generally corrode ownly and show more resistarce to the 1

more damaging intergranular penetration than do PE16 and M316 '

' alloys. Corrosion is rnost sewre at 1:1 compns. irresp. of O potential, and the corrosion products are Cs2Te + transition metal tellurides. The corrodants Cs2Te, Cs2Te + Cs, and Cs are inert to . .

c '

FV448 in the absence of 02, but corrosion increases with increasing O potential. At low potentials, the dominant corrosion products are '

Cs chromates + Cr2Te3, and these are augmented by CsFeO2 + Fete 0.9 '

at higher potentials. The various types of corrosion are summarized.

O L28 ANSWER 16OF33 CA COPYRIGHT 1997 ACS AN 11Q:122277 CA >

TI Chemical reactions between "* cesium *" , tellurium and oxwen ,

with fast breeder *** reactor *** cladding alloys. Part II .the

• • • corrosion *** by *** cesium *** -oxygen mixtures AU Pulham, R. J.; Richards, M. W.

CS Chem. Dep., Univ. Nottingham, Nottirg' ham, NG7 2RD, UK i SO J. Nucl. Mater. (1990),172(1),47-53 CODEN: JNUMAM; ISSN: 0022 3115 DT Journal LA English AB The corrosion of the alloys PE16, M316 and FV448 by liq. Cs in sealed capsules was studied after 168 h at 948 K in the presence of a buffer of Ni/NiO which set the thermodn. O potential at . DELTA..hivin.GO2 = -307 kJ mol 102. At this relatiwly high potential, the Cs is converted to a liq. phase contg. ca. 33 mol% O at 948 K, and all3 alloys are severely corroded with the degree decreasing in the order M316.mchgt. FV448 > PE16. With 316 the corrosion is entirely intergranular causing sewre longitudinal disruption between the grains. With PE16, the corrosion is intergranular ahead of layering. With FV448, there is a combination of intergranular and transgranular corrosion on an ewn front. The chem. of the corrosion is dominated by the formation of Cs chromate '

and ferrite.

23 Tottech' l.iterature Search Service Effects of Steam on Nuclear Fuel Components: Port 2 May 7,1997 r%Y 87 '97 13: 34 PON "?d

__ -. - --- - ~ ~~

=4-O L28 ANSWER 17 OF33 CA COPYRIGHT 1997 ACS AN 113:122269 CA TI Chemical reactions between *** cesium *** , tellurium, and oxygen with fast breodor *** reactor *** cladding alloys. Part lil-The effect of oxygen potentialon the *** corrosion *** by i

• • • cesium *** -tellurium mixtures AU Pulham, R. J.; Richards, M. W.

CS Chem. Dep., Univ. Nottingham, Nottingham, NG7 2RD, UK SO J. Nucl. Mater. (1990),172(2),20619 CODEN: JNUMAM;ISSN: 0022G115 DT Journal LA English AB The corrosion cf the alloys PE16 and M316 by Cs:Te mixt. (1:1,2:1 and 4:1) was studied in sealed capsules under partial pressures of O set by metal / metal oxide couples at 948 K for 168 h. The alloys suffered severe integranular corrosion by the 1:1 mixts. and corrosion seemcc independent of O potential. The 2:1 and 4:1 mixts.

produced a matrix (alternate layering) type fo corrosion and the '

depth of corresion increased with increasing O potential. At the highest potential the 4:1 mixt. differentiated the 2 alloys; PE16 suffered a combination of intergranular and matrix attack whereas l M316 was penetrated throughout (mainly by Cs/0) integranularly.

Generally Cs, Te, Cr and O were found assocd. (as a mixt. of C ,

chromate and C telluride), and PE16 was more resistant that M316 steel to corrosion.

l

• i i

. 1 l

l 24 Teltech' Uterature Search Service Effacts of Stoem on Nuclear Fuel Components: Port 2 May 7,1997 f%Y 07 '97 13:34 Do" " )

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us,ce,3r 16.s.s;as~ -i ~~ ~ 5[(.sd bh4ts hoe if26

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0 L28 ANSWER 18 OF33 CA COPYRIGHT 1997 ACS AN 113:86704 CA TI Chemical reactions of *** cesium *" , tellurium, and oxygen with fast breeder ***rcactor*** cladding alloys. Part I- the

"* corrosion *" bytellurium AU Pulham, R. J.; Richards, M. W.

CS Chem. Dep., Univ. Nottingham, Nottingham, NG7 2RD, UK SO J. Nucl. Mater. (1990),171(2-3),319-26 CODEN: JNUMAM;ISSN: 0022-3115 '

DT Journal l IA English l

AB The corrosion of the alloys PE16, M316, and FV448 by liq. Te in sealed capsules was studied after 168 h at 948 K both with and without a bufferof Mo/ moo 2. This buffer sets the thermodn. O l

potential at . DELTA..hivin.GO2 = -417 kJ mol-102. All 3 alloys were sewrely corroded and the extent was in the order M316 > PE16 = '

FV448. O diminished the extent of corrosion but did not change the

• order. All 3 alloys carried 2 layers of corrosion products on top l

' of a damaged metal surface. The outer layer contained Fe1.5 Nil.5Te2 for PE16 but Fe2.25Te2 for M316 and FV448. The inner layer contained Cr2Te3 for all 3 alloys. The damaged metal surface was largely Cr2Te3. The substrate alloys (PE16 and FV448)were ,

penetrated intergranularly by Te, but M316 exhibited an een band of l Cr-depleted steel. The results are explained by the diffusion of Te into the alloy and the crffusion of the alloy components in the i

opposite direction. The corrosion is exacerbated by dissoin. of i

l alloy into the liq. Te from which the metal tellurides subsequently ppt.

1 l

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25 Toltech' Literature Search Service Effects of Steem on Nuclear Fuel Components: Port 2 May 7,1997 l f AY 07 '97 13: 35 por.c ?c

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awruhr hub.d ~ -> Sd3 Ub b96 pag 582i O L28 ANSWER 23OF33 CA COPYRIGHT 1997 ACS AN 110:103504 CA Tl Hot *** corrosion *** -induced cracking of fuel pin cladding tubes of chromium-nickei stesls AU Dienst, W.; Hofmann, P.; Liesenfeld, U.

CS Inst. Mater. Festkoerperforsch. I, Kernforschungszent. Karlsruhe, Karlsruhe, D-7500/1, Fed. Rep. Ger.

SO J. Nucl. Mater. (1988),160(1),1-9 i CODEN: JNUMAM;ISSN: 0022-3115 DT Journal LA English AB Siress rupture expts. were conducted in tube burst and tensile tests under the influence of S, Ss, Te, I and Cs2Te, esp. in combination with NiO whch was employed to establish a relatiwly high O

, potential. The stress lewis were chosen in such a manner that a rupture life of .apprx.1 h was achiewd (without the influence of the reactive species) at 700,800,900 and 1000. degree., and of about 100 and 500 h at 700. degree..

The cladding tube materials tested were X10CrNiMoTiB1515 (1.4970), .

X5NiCrTi2510 (RGTX1), X10NiCrMoB2510 (RGTX2) a nd X17CrMoVNb121 (1.4914). Effects and phenomena that could be due to stress corrosion cracking were obsd. mainly in 1 h tests at 700. degree., under the influence of I + NiO and Cs2Te + NiO There was a sewre redn. in time-to-failure and elongation at fracture.

The a usteniti: steels RGTX1 and 2, which are particularly poor in Cr

(.apprx.10 wt%), were most sensitiw to stress corrosion cracking.

The martensitic-ferritic steel 1.4914 appeared to be completely insensitive. Grain bourdary attack and sepn. seemed to be responsible for the effect of stress corrosion cracking.

O L28 ANSWER 24 OF33 CA COPYRIGHT 1997 ACS AN 109:117959 CA Tl Stabilization of radioactim clad IN Watanabe, Takeshi; Suzuki, Kazunori PA JGC Corp., Japan SO Jpn. Kokai Tokkyo Koho,3 pp.

CODEN: JKXXAF PI JP63127199 A2 880531 Showa Al JP86-273667 861117 DT Patent LA Japanese AB ** *N uclea r*** *** reactor *** spent fuelclad mainlycomposed of Fe oxides contg. radioelements is heated to 800-1300. degree. in a reducing enpronment to make a sintered or molten product. In this

! method the radioelements such as 60Co 54Mn and 63Ni are contained in Fe matrix as a solid soin. Thus, a mixt. of . alpha.-Fe2O3 (20.8 g), coo (0.5 g), and C (6.0 g) was heated at 1100. degree. for 40 h to giw a 17.2 g half-molten mass contg.15.2 g Fe-Co solid soln.

' IT *"1333.74-0*** , *** Hydrogen *** , uses and miscellaneous RL: USES (Uses)

( *** reducing"* agent, in radioactiw waste clad treatment) 26 Toltech' Literature Search Service Effects of Steam on Nuclear Fuel Components: Port 2 i

l May 7,1997

m Y 07 '97 13

36 "*

s-- __ _ m. m-M-1 aJ-s _ J- ~ X A e - m-- --.LA -M -.,sa,. .a,.. _J E+g 3 03/ G r/ 'J( 13 4f*Er -? bt!3 (.15 b34tl hee BZ8 es

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l 0 L28 ANSWER 25 OF 33 CA COPYRIGHT 1997 ACS AN 107:207215 CA Tl Process and apparatus foroxidizing or *** reducing *** disso!wd

. substances )

IN Kitamori, Takehiko; Nishi, Takashi; Fukasawa, Tetsuo; Fujimori, Haruo; Sasahira, Akira; Ozawa, Yoshihiro; Suzuki, Kazumichi; Yusa, Hideo PA Hitachi, Ltd. , Japan SO Eur. Pat. Appl.,18 pp.

CODEN:EPXXDW 1 Pl EP233498 A2 870826 )

i DS R: BE, DE, FR, GB l

Al EP87-100807 870121  !

! PRAIJP8610156 860122 '

l DT Patent IA English AB The title process and app., useful for *** nuclear"** fuel

reprocessing, inwlw adding a particulate semiconductor  ;

i photocatalyst and a H20-sol. e acceptor or donor for the i photocatalyst to a soln. and irradiating the photocatalyst with an  ;

electromagnetic waw having an energy high enough to excite the j photocatalyst and having a wawlength outside the absorption region 1 of a precursor substance capable of producing an oxidn.-or l t

l redn.-inhibiting substance. Thus, a simulated waste soin. contg.

i 0.1 g Ru/L was mixed with TiO2 powder having Pt on the surface and ETOH as e donor for preventing the rewrse reaction. When the TrO2 was irradiated with a Xe lamp with available light intensity

.apprx.200 mW in soln., almost all of the Ru was deposited on Pt in

.apprx.1 h bycutting off the light with wawlength <330 nm.

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1 27 Toltech' Literature Search Service Effectk of Steam on Nuclear Fuel Components: Part 2 May 7,1997

! MAY 07 '97 13: 36 co'c *

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• O L28 ANSWER 26 OF 33 CA COPYRIGHT 1997 ACS AN 107:185478 CA Tl Stress "* corrosion *" cracking of zirconium alloys AU Cox, B.

CS Chalk Riwr Nucl. Lab., Atomic Enerl af Cenada Ltd., Chalk Riwr, ON, K0J 1J0, Can.

SO Langmuir(1987),3(6),867-73 CODEN: LANGD5;ISSN: 0743-7463 DT Journal .

LA English OS CJACS AB The environments in which Zr alloys suffer stress corrosion cracking are summarized and the conditions leading to delayed hWride cracking identified. The 2 processes are clearly distinguishable fractog., and these differences are illustrated. The mechanisms of stress corrosion (under conditions where hWride cracking is absent) are discussed in terms of the important steps in the process. These rate-controlling steps are examd. for their relevance to the ~

pellet < lad interaction (PCI) phenomeron that can lead to the failure of "* nuclear *** "" reactor *" fuelcladding. The

• actiw fission product species causing PCI failures are discussed; between I and Cs/Cd as culprits, the choice is 1.

IT "*1333-74 0"* , *" Hydrogen *" , properties RL PRP(Properties)

(soly, of, in zirconium, stress *" corrosion *** cracking of zirconium alloys in relation to) -

O L28 AN,SWER27 OF33 CA COPYRIGHT 1997 ACS AN 107:29754 CA Tl "* Corrosion *** deposits and their removal from heat-transfer surfaces of RBMK-type *** reactors ***

AU Varovin, l. A.; Eperin, A. P.; Konstantinov,*

E. A.; Sedov, V. M.;

Senin, E. V.; Filippov, E. M.

CS USSR SO At. Elektr. Stn. (1985),8,84-6 CODEN: AESTDA DT Journal

! IA Russian AB The chem. compns. of corrosion products in loose deposits with regard to structural materials of stainless steel, C steel and Zr alloy (2.5% Nb) are tabulated for a RBMK-1000 "* reactor *" .

The Fe is found in the form of . gamma.-Fe2O3 and Fe304. The sp.

activity of the following radionuclides on the inner surfaces of the

"* reactor"* circuit is giwn: 51Cr,54Mn,58Co,59Fe,60Co, l 64Cu,65Zn,95Zr + 95Nb,106Ra, and 124Sb. The decontamination is carried out by using an oxalic acid soln. followed by treatment with H202 to form oxalate complexes of Fe(Ill) which dissolw easily.

Thus, the pre-servicing decontamination of the RBME-1000

• • • reactor *" is economically expedient operation.

1 28 Toltech' Literature Search Service Effects of Stoem on Nuclear Fuel Componente: Port 2 May 7,1997 MAY 07 '97 13:37 com M

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• O L28 ANSWER 28 OF 33 CA COPYRIGHT 1997 ACS AN 106:57498 CA Tl The *** corrosion *** of spent uranium dioxide fuelin synthetic groundwater AU Forsyth, R. S.; Werme, L. O.: Bruno, J.

CS Studsvik Energiteknik A.B., Nykoeping, Swed.

SO SKB Tech. Rep. (1986),85-16,33 pp.

CODEN: STRPEP DT Report IA English AB in connection with spent fuel disposal, teaching of high burnup BWR fuel for up to 3 yr showed that both U and Pu attain satn. rapidly at pH 8.1, giving wlues of 1-2 mg/L and 1.mu.9 /L, resp. The leaching rate for 90Sr decreased from about 10-b/d to 10-7/d, but was always higher than the rates for U, Pu, Cm, Ce, Eu and Ru.

Congruent dissoin. was only attained at pH wlues of .apprx.4. When reducing conditions were imposed on the pH 8.1 groundwater by means of ***H2*** /Arin the presence of a Pd catalyst, significantly lower leach rates were attained. Tha hypothesis that alpha radiolytic decompn. of water is a driving force for UO2 corrosion ewn under reducing conditions has been examd. in teaching tests on low burnup (Iow alpha dose-rate) fuel. No significant effect of alpha radiolysis under the exptl. conditions was detected.

Thermodynamically the caled. U solubilities in the pH range 44.2 generally agreed, well with the measured ones, although assumptions made for certain pararneters in the calens. limit the validity of the results.

l 29 Toltech' Literature Search Service Effects of Steam on Nuclear Fuel Components: Part 2 May 7,1997 l

l MAY 07 '97 13:37 Pom v

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O L28 ANSWER 29 OF 33 CA COPYRIGHT 1997 ACS AN 104:194963 CA Tl Stress *** corrosion *** cracking of Zircaloys in unirradiated and irradiated *** cesium *** iodide AU Cox, B.; Surette, B. A.; Wood, J. C.

CS Chalk Riwr Nucl. La b., At. Energy Canada Ltd., Chalk River, ON, K0J 1J0,Can.

SO J. Nucl. Mater. (1986),138(1),89-98 CODEN: JNUMAM; ISSN: 0022-3115 DT Journal IA English AB Unirradiated split-ring specimens of Zircaloy fuel cladding, coated with Csi, cracked when stressed at elevated temps. The specimens were reexamd. fractog. and metallog. The cause of cracking was stress corrosion (SCC) and not delayed hydride cracking (DHC).

Further specimens were cracked at 350. degree. by a soin. of Csl in a fused mixt. of nitrates of Ru, Cs, Sr, and Ba by a similar mechanism. Csl dissolved in a fused molybdate melt was not stable at 400. degree., and rapidly swived I, leaving a melt that was incapable of causing SCC. Irredn. of stressed split ring specimens of Zircaloy fuel cladding in a . gamma.-irradiator of 106 R/ h and in the U-5 loop in the NRU *** reactor *** at an estd.109 R/

h caused SCC when the specimens were packed in dry Csl ,

powder. Care had to be teken to dry the Csl, otherwise cracking occurred by a DHC mechanism from ***H*** absorbed from residual moisture in the Csl. The crack surfaces obtained with dry Csl were typical of I-induced SCC rather than Cs-induced metal vapor embrittlement. Thus, if a transport process is provided for the iodide to obtain access to the Zrsurface, Cslis capable of causing SCC of Zircaloy. This transport process might be ionic diffusion in a fission product oxide melt in the fuel-clad gap, howewr, radiolysis of Csl to form a volatile I species in a radiation field is the more probable explanation of pellet-clad interaction failures.

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30 Toltech' Literature Search Sersice Effects of Steam on Nuclear Fuel Components: Part 2 May 7,1997 MAY 07 '97 13:38 Do * ""

auscesya so.sv.cm -s ucy t.23 toca y g2 u3z

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d O L28 ANSWER 30OF33 CA COPYRIGHT 1997 ACS AN 103:77793 CA TI The coupled *** kinetics *** of grain growth and fission product behaviorin *** nuclear *** fuel underdegraded< ore accident conditions AU Rest, J.

CS Mater. Sci. Technol. Div., Argonne Natl. Lab., Argonne, IL,60439, i USA 4 SO J. Nucl. Mater. (1985),131(2-3),291-302 CODEN: JNUMAM;ISSN: 0022-3115 DT Journal IA English l'

AB The theor, FASTGRASS-VFP model was used in the interpretation of fission gas, I, and Cs release from (1) irradiated high-burnup LWR fuel in a flowing stream atm. during high-temp., in-cell heating tests, and (2) trace-irradiated LWR fuel during sewre-fueldamage (SFD) tests. A theory of grain boundary sweeping of gas bubbles was included within the FASTGRASS-VFP formalism. This theory considers the interaction between the moving grain boundary and 2 distinct size classes of bubbles, those on grain faces and on grain edges, and provides a means of detg. whether gas bubbles are caught up and moved along by a moving grain boundary or whether the grain boundary is only temporarily retarded by the bubbles and then breaks away.

In addn., as FASTGRASS-VFP provides for a mechanistic calen. of intra- and intergranular fission product behavior, the coupled calen. between fission gas behavior and grain growth is kinetically comprehensiw. Results of the analyses demostrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a graingrowth/ grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. The effect of fuel oxidn. by *** steam *** on fission product and grain growth behavior is also considered. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fissinn product behavior in trace-irradiated and in high-burnup fuel are highlighted.

31 Teltech' Literature Search Service Effects of Stoem on Nuclear Fuel Components: Port 2 May 7,1997 MAY 07 '97 13:38 NV

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us/mif is.. sci.co -i ~ dei s'sshu F M H33 1

.a t e, O L28 ANSWER 31 OF 33 CA COPYRIGHT 1997 ACS AN 99:147922 CA Tl The effect of pile radiation, high helium pressure,

• • • corrosion *** and gaseous HTR<oolant impurities on the diffusion of *** cesium *" in a graphitic matrix AU Hensel, W.; Hoinkis, E.

CS Hahn-Meitner-Inst., Berlin, D-1000/39, Fed. Rep. Ger.

SO Hahn-Meitner-inst. Kemforsch. Berlin, [Ber.] (1983), HMI-B 372, '

Transp. Fission Prod. Matrix Graphite,73-6 CODEN: HMIKAT; ISSN: 0440-0836 DT Report IA English l

AB Cs4iffusion behavior changed from a diffusion-trapping process in an as-received graphitic matrix A33 to a classical diffusion process in an oxidized (.gtorsq.0.3 wt.%) matrix. Diffusion coeffs.

t increased with increasing extent of corrosion. Csdiffusion coeffs. I were independent of gaseous HTR-coolant impurities ( "*H2"* ,

CO, CO2, CH4) and of ***H*** with a partial pressure of 0.37 torr. A fast n flux of 2.4. times.1013 cm-2-s-1 (E > 1 MeV) in the presence of 4 bar He had no effect on the transport of Cs.

IT ***1333-744*** , uses and miscellaneous RL USES (Uses)

(HTR coolant impurity, effect on *** cesium *" diffusion in graphitic matrix)

O L28 ANSWER 320F33 CA COPYRIGHT 1997 ACS AN 93:33511 CA Tl Model for *** kinetics *** of *** cesium *** -graphite sorption AU Zumwalt, L R.; Eazi, N.1.

CS North Carolina State Univ., Raleigh, NC, USA SO Trans. Am. Nucl. Soc. (1980),34,215-16 CODEN:TANSAO; ISSN: 0003418X DT Journal IA English AB The proposed model, with an activated sorption site no. distribution decreasing exponentially with energy, fits obsd. Csgraphite (

• • • H*** -451) power sorption-desorption kinetics quite well. At ,

low equi!. wpor pressure of Cs (<1 Pa), the time to reach equil. is

>105 s (i.e. days).

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Tottech' Literature Search Service Etfacts of Stoem on Nuclear Fuel Componente: Port 2 May 7,1997 1

j NY 07 '97 13: 39 'OC# "

4 +4 -  %. -.i--..a_+-h e -.e- -+5 4 - - - +---,a 4-a;i-. 4 -Aa, .4u a--4 & 2 A-- - -41 ..i as/ vin As.sa.s s -> dus ud bw.u Pisa 834

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O L28 ANSWER 33OF33 CA COPYRIGHT 1997 ACS AN 91:200852 CA Tl influences of *** cesium *** and *** cesium *** oxide on iodine stress *** corrosion *** cracking of Zircaloy-2 in outof-pile and in pile conditions AU Une, Katsumi CS Toshiba Res. Dev. Cent., Toshiba Corp., Kawasaki,210, Japan SO J. Nucl. Mater. (1979),87(1),207-10 CODEN: JNUMAM;ISSN: 0022-3115 DT Journal LA English AB A cold-worked Zircaloy-2 [11068-94-3] cladding tube was used for the in-pile (B . times.1016 n/m2-s for 8 24 h) and outof-pile expts. The i stress-corrosion cracking of Zircaloy in the outof-pile expts depended on the Cs and Cs2O [2028100-9]

conens.12 reacts with Cs and Cs2O to form Csl which ties up the I and suppresses the stressoorrosion cracking. Ewn under irradn.

conditions, if enough Cs2O is present, there is insufficient conen.

I decompd. from Csl by radiolysis to cause stress <orrosion cracking.

l 0 SEARCH STRATEGY FILE 'CA' ENTERED AT 13:58:58 ON 07 MAY 1997 L1 244005 SEA (CESIUM OR COBALT?)m,lT,ST l L2 3732 SEA L1 AND FUEL?/IT,ST L3 1667 SEA L2 AND (NUCLEAR? OR REACTOR? OR 70/CC,SX OR 71/CC,SX) l L4 , 61 SEA L3 AND STEAM?

l L5 29 SEA L4 AND (HYDROGEN OR H2 OR 1333-744) l ACT MAYTGAL/A L23 82 SEA L17 OR L19 OR L22 FROM PREVIOUS SEARCH l L24 20 SEA L5 NOT L23 l L25 327 SEA L3 AND (STEAM 9 OR HYDROGEN OR H2 OR H OR 1333-74-0) l L26 286 SEA L25 NOT(L5 OR L23)

L27 43 SEA L26 AND (CORROS? OR KINETIC? OR REDOX OR REDN OR REDU C')/TI,fT,ST L28 33 SEA L27 AND (NUCLEAR? OR 70/CC SX OR 71/CC.SX)

L29 53 SEA L24 OR L28 SAVE L29 MAYTGAL2/A l

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33 htech' Literature Search Service Effects of Steem on Nuclear Fuel Components: Part 2 May 7,1997 1

t%Y 07 '97 13:40 DON ' 8

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%6 6 g, t O SEARCH STRATEGY Set items Description S1 35376 (CESIUM OR COBALT? OR CS0137 OR 137CS OR CS137 OR CO60 OR COO 60 OR 6OCO)

S2 1545 S1 AND FUEL?

S3 1024 S2 AND (NUCLEAR? OR REACTORDS,DE,lD AND FUEL?m,DE,tD l S4 444 S3 AND (RSSION0 PROD?) - l S5 160 TI,DE S4 AND (CORROS? OR REDOX OR REDUC? OR HYDROOEN OR STEAW)/- ,

S6 168 SS ANDSim,DE S7 61 S6 AND STEAM? i S8 8 S7 AND KINETIC?

l S9 96 S3 AND STEAWM,DE S10 28 S9 AND HYDROOEN i

S11 31 S8 OR S10 '

S12 l

18 S11 NOT(WASTE? OR RECOVERY OR REPROCESSD Temp SearchSaw "TD535" stored ('

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