- Email: [email protected]
The advisability of prototypic testing for space nuclear systems
Acta Astronautica 57 (2005) 404 – 414 www.elsevier.com/locate/actaastro
The advisability of prototypic testing for space nuclear systems Roger X. Lenard Little Prairie Services, USA Available online 23 May 2005
Abstract From October 1987 until 1993, the US Department of Defense conducted the Space Nuclear Thermal Propulsion program. This program’s objective was to design and develop a high specific impulse, high thrust-to-weight nuclear thermal rocket engine for upper stage applications. The author was the program manager for this program until 1992. Numerous analytical, programmatic and experimental results were generated during this period of time. This paper reviews the accomplishments of the program and highlights the importance of prototypic testing for all aspects of a space nuclear program so that a reliable and safe system compliant with all regulatory requirements can be effectively engineered. Specifically, the paper will recount how many non-prototypic tests we performed only to have more representative tests consistently generate different results. This was particularly true in area of direct nuclear heat generation. As nuclear tests are generally much more expensive than non-nuclear tests, programs attempt to avoid such tests in favor of less expensive non-nuclear tests. Each time this approach was followed, the SNTP program found these tests to not be verified by nuclear heated testing. Hence the author recommends that wherever possible, a spiral development approach that includes exploratory and confirmatory experimental testing be employed to ensure a viable design. © 2005 Published by Elsevier Ltd.
1. Background From October 1987 until 1992, the US Department of Defense conducted the Space Nuclear Thermal Propulsion program. This program’s objective was to design and develop a high specific impulse, high thrust-to-weight nuclear thermal rocket engine for upper-stage applications. The selected technology was the Particle Bed Reactor (PBR) as shown in Fig. 1. The PBR employs the high surface-to-volume ratio exhibited by small fuel particles to minimize the
E-mail address: [email protected]. 0094-5765/$ - see front matter © 2005 Published by Elsevier Ltd. doi:10.1016/j.actaastro.2005.03.056
temperature drop within the fuel to operating gas, in this case hydrogen. The Fuel particle is a variant of the TRISO (Triply Isotropic) fuel particle, except in the case of the PBR, the fuel particle outer coating was Zirconium Carbide instead of Silicon Carbide. The fuel particle was also substantially smaller in diameter, in the PBR case, the fuel particle is 0.4 mm in diameter. Fig. 2 shows some of the as-fabricated fuel particles. These fuel particles were placed between an outer flow control assembly and an inner porous liner called a frit. This configuration is displayed as a graphic called “Fuel Element” in Fig. 1. The inner frit was required to withstand the high-temperature exhaust −3000 K and was also required to withstand any thermal or
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
405
Fig. 1. Description of the Particle Bed Reactor.
Fig. 3. Non-irradiated SNTP particle fuel.
2. Developing the fuel particle
Fig. 2. Particle Bed Reactor concept.
mechanical shocks or stresses during startup, operation and shutdown of the rocket engine. The PBR engine maintained an epi-thermal fission neutron spectrum due to the presence of a polyethylene moderator, shown in Fig. 1 as a purple area. The polyethylene moderator is shaped as a right hexagonal parallelpiped with axial grooves along the exterior of the moderator to support hydrogen flow. The polyethylene is cooled by gaseous hydrogen from the turbo-pump assembly. The entire assembly, moderator, cold flow controller, fuel particles and hot frit are assembled into a fuel element that is later assembled in a hexagonal-closepacked assembly that forms the reactor core. A composite pressure vessel, thrust vector control and turbopump assembly are installed to form the entire engine as shown at Fig. 2. Specifications for the engine are shown in the figure as well [1].
The fuel form for the PBR was substantially new. The development of a robust fuel form for Nuclear Thermal Propulsion has been an extremely difficult task. The program decided to create a particle fuel fabrication pilot line at [then] Babcock & Wilcox in Lynchburg, Virginia. We employed a sol-gelation process to convert uranium oxide to uranium carbide (UC). This formed the kernel of the fuel particle. The nominal diameter of the internal fuel kernel was −220 m. The fuel kernels were transferred to a chemical vapor deposition furnace where a thin layer of low-density graphite was placed on top of the UC kernel. The low-density graphite layer is designed to capture fission products generated in the power generating process. On top of that layer a densified layer of pyrolitic graphite was deposited, flowed by a layer of zirconium carbide. A series of un-irradiated fuel particles is shown in Fig. 3. A magnified view of a cross-sectioned fuel particle is shown in Fig. 4. This particle has undergone a nuclear heating test
406
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
Fig. 4. Fuel particle after nuclear in-core heating tests.
and has been cycled to high temperature a number of times. You can note the separation of the lowdensity graphite layer from the fuel kernel. In this figure, one can also note that some of the UC kernel has migrated into the ZrC coating. One can note the consequences of repeated high-temperature operations and substantial numbers of thermal cycles. As may be noted, there is significant grain growth in the heated sample. This grain growth continues until the grain boundaries become columnar in the radial direction. Each grain boundary is a potential source of uranium leakage or carbon depletion into the hot hydrogen fuel. Fuel particles were subjected to a number of tests in both in-core and ex-core heating tests. Results of ex-core heating tests showed good performance of the fuel particles, however, for in-core testing the results were dramatically different. Fig. 5 shows the particle heating test (PHT) apparatus. There were three zones to the test apparatus, each experiencing a different neutron fluence, hence operating power level. Since
Fig. 5. Particle heat test (PHT) in-core test column.
cooling was constant for all particle containers, the temperature at each location was different. Temperature ranged from 2100 to 3100 K. The tests showed that after repeated nuclear heated operations, some of the fuel particles sloughed off layers of materials, including the zirconium carbide coating.
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
407
During the course of the nuclear heating in the Annual Core Research Reactor (ACRR), some of the fuel particles simply turned into dust. This required that the program concentrate on developing a sound fuel form. Several developmental pathways were taken resulting in a mechanism to identify unacceptable particles and sound ones. These processes were used to generate fuel particle coatings with acceptable performance and lifetime for the next series of tests, called the particle in-pile element (PIPE) series of tests.
3. PIPE testing program 3.1. Background The PIPE testing program was designed to take the SNTP system beyond isolated fuel particle testing. The next step was to integrate fuel particles into a stand-alone fuel element that could be tested in the Sandia National Laboratories Annular Core Research Reactor (ACRR). The ACRR exhibits properties highly suited to stockpile stewardship verification. The ACRR is a swimming-pool reactor with a very unique (UO2 –BeO) fuel that testing. It has very high heat capacity, high thermal conductivity and enables very high power (70,000 MW)/high energy (300 MJ) pulses. It can also operate at a steady-state power level of 2 MWth or at 4 MWth for short periods of time. The ACRR has a dry central cavity that can accommodate an experimental package that is up to 9 in–225 mm in outside diameter. The PIPE experiment employed this large experiment package size to develop two different test articles. 3.2. PIPE description The PIPE experimental package featured a fullscale, half-length pre-prototypic fuel element shown at Fig. 6. The fuel element was a radial in-flow fuel element with a bed thickness of approximately 2 cm. The outer channel retainer was fabricated from sintered stainless steel to provide a low-pressure drop hydrogen flow path. The inner hot frit was made from rhenium, external source of neutrons to enable fissions to occur within the fuel bed. This heated the hydrogen flowing over the bed. The ACRR was operated in an extended pulse mode being heated in the fuel volume.
Fig. 6. PBR in-pile experiment (PIPE) test assembly.
The ACRR provided an to provide the required neutron fluence. The experimental setup was designed so that the entire capsule was self-contained. Hydrogen gas was circulated through a small pre-cooled pebble bed and into the fuel element. Heated gas was diagnosed and returned to the pebble bed. This allowed for only a short-duration test sequence.
408
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
Fig. 7. Predicted PIPE hydrogen outlet temperatures.
The program proposed that this test was simply a proof-of-principle test that would allow sufficient information to be generated and analyzed that a more realistic fuel element would be designed. Results from the PIPE-1 experiment were initially promising, in that high-temperature gas was found to flow at the predicted rate with a power density of 3 MW/l, which was at least as high as the highest power density achieved in the NERVA program, but due to limitations of the ACRR, the run was limited to approximately 10 s overall. The predicted performance, which was reasonably closely matched by actual test data is shown in Fig. 7. A second PIPE test conducted in the ACRR was not as successful due to a different material fabrication process used for the rhenium hot frit. The hot frit apparently experienced some initial failures at approximately 2800 K gas temperature and melt began to occur in the hot frit. As a result there was an increase in overall gas output temperature to approximately 3000 K. A post irradiation examination of both fuel elements showed only minor damage in the PIPE1 experiment, but major damage in the PIPE-2 experiment. However, sufficient data had been gathered to proceed with a more robust design of the cold and hot frits as well as continue development of the ZrCcoated fuel particles. In each case, the fuel element was heated thermally by an external source with few problems encountered, therefore, the differences between internal nuclear heating and external thermal heating were once again brought to the forefront of the testing program.
Fig. 8. Particle bed reactor critical experiment.
4. Critical experiment As technology began to advance for the PBR, it was necessary to understand the overall design. Many Monte Carlo neutronics calculations were performed on the SNTP core, which varied from 19 to 37 or to as high as 61 fuel elements in the core depending upon reactor power level and fuel element size. A companion effort attempted to determine pressure drop and hot channel factors associated with a PBR system. However, in order to ensure a reliable design, the MCNP neutronics code runs needed to be benchmarked [2]. To accomplish this, the PBR Critical Experiment was designed and fabricated [3]. There was widespread controversy over this decision at first, because it was believed that analytical tools had achieved a level of performance that verification tests were unnecessary. From start to first operation, the PBRCX took 18 months and cost $19M, including all safety and environmental compliance costs. The development and operation of the PBRCX was a major programmatic achievement. A photo of the PBRCX is shown at Fig. 8. After the initial critical
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
409
Fig. 9. Peek-a-Boo shutter control worth levels. Fig. 10. Platelet Technology Cold Screen.
operation, a wide range of experiments were performed to verify code predictions. During the course of the program, the program manager invented a new type of internal controller called the Peek-A-Book (PABOO) shutter controller. Since it was necessary to determine the utility of such a controller, tests were performed on this and other important benchmark criticality measurements. We determined that the PABOO shutter controller provided the necessary reactivity shift to control the PBR rocket engine. Results of reactivity measurements on the PABOO controller are shown in Fig. 9 [4]. We then decided to measure the temperature coefficient of reactivity. This proved to be a watershed event. In retrospect, those who had opposed the PBRCX on grounds that analytical tools were sufficiently advanced saw that unanticipated interactive effects of the design cannot always be effectively modeled. We discovered that the PBR rocket engine, while having a goal of a negative temperature coefficient in reality exhibited a positive temperature coefficient over a modest temperature range [5,6]. This was a completely unexpected event and resulted in a short stand down for safety purposes when it was discovered that the CX could exhibit a positive temperature coefficient, when all the safety documentation showed (from analysis) that the temperature coefficient would be negative [7]. The results from the PBRCX definitely show the importance of prototypic testing in order to obtain results usable for sound engineering design.
4.1. Advanced fuel elements Based on neutronics, thermal-hydraulic and mechanical analyses, the program acknowledged the need for an advanced fuel element.
4.2. Advanced cold inlet screen The new fuel element would replace the sintered stainless-steel entry screen with an aluminum flow engineered screen employing the Honeywell stacked plate process. This process is commonly known as platelet technology, because thin films of material are photo-etched, cleaned and then stacked upon each other. The final stack is then diffusion-bonded together to form an integrated whole. The SNTP program’s advanced cold screen is shown in Fig. 10. The engineered screen allowed the flow to be enabled in relation to the way power was generated in the fuel bed. Since the reactor exhibited both radial and axial power profiles, in order to maintain a uniform outlet temperature, flow needed to be managed over the entire packed bed. The flow engineered screen allowed this to occur. The cold screen also contained a compliant layer that, at least in theory, would accommodate bed expansion and contraction as the fuel particles were heated or cooled with nuclear heat.
410
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
Fig. 11. Niobium carbide-coated graphite hot frit.
or NET-0. Performing a complete dry-run enabled our team to determine problems with equipment, procedures, safety considerations and in general ensure that the test would proceed smoothly. As will be noted below, some important findings were discovered during the pre-test.
4.3. Hot frit 4.5. NET test capsule description The problems associated with the PIPE-2 experiment required that the program use materials that would not form lower temperature eutectics thereby enhancing probability of failure of the fuel element. For this reason, two developmental pathways were selected. The first pathway was a laser-drilled POCO graphite that would be CVD coated with niobium carbide. A photo of a NbC-coated graphite hot frit is shown at Fig. 11. This hot frit was extracted from the nuclear element test (NET) experiments (to be described later); this particular hot frit was fractured during initial power operations of the NET experiment. While the NbC-coated hot frit performed well from a compatibility perspective, there existed more serious design issues associated with the PBR design itself, some of which were revealed in the NET. 4.4. Nuclear element tests The advancement of component technology was sufficiently compelling that the program determined it was a fortuitous time to integrate a fuel element comprising components that had been specifically developed to counteract issues associated with the PIPE test series. To this end, a NET series fabrication development and test pan was generated. The fuel and the end fittings would be provided by Babcock & Wilcox, the hot frit by Brookhaven National Laboratories, and compliant layer cold screen by Garret Fluid System Division, and the entire unit would be integrated by Babcock and Wilcox. Since nuclear facility operations time is very expensive, the project determined that a precursor test run would be accomplished with simulated fuel particles—this would eliminate considerable security costs during shipment and at Sandia National Laboratories. The fuel element would be full-scale, but half-length. It would be fitted into the identical test device to perform a dry run on the experiment. This fuel element test was called Nuclear Element Test-0
The NET test would be a substantial increase in thermal power (1 MW versus) 3 kW for PIPE. Also, a higher power density in the bed was anticipated ∼3–5 MW/l. This required a substantial amount of cooling if the test objectives were to be achieved. The Net test capsule is shown in Fig. 12. The capsule contains a mounting location for the fuel element assembly, a fan drive motor and fan, cooling channels, a thermal heat sink instrumentation access and various safety related items. The important feature of this capsule was that the heat sink could be cooled to liquid nitrogen temperatures before the test. This would more closely replicate the actual inlet conditions of the hydrogen when the rocket engine would be in operation. The motor and fan drive were of a speciality design built by Skurka engineering. This motor was a several kilowatt motor due to the power requirements to drive the hydrogen through the circuit. The drive motor contained special magnets that, during the pre-test formed a metal hydride and turned to powder, so less capable, but hydrogen compatible magnets were required. This entire capsule was then placed inside a containment capsule that would be placed within the central cavity of the ACRR. Fig. 13 shows the NET fuel element prior to testing. 4.6. NET tests The uranium fueled NET element was shipped to Sandia National Laboratories for testing. The NET-1 fuel element is shown in Fig. 13 in its ready-to-install configuration. The NET capsule was sealed and placed into the ACRR for testing. The initial tests were simply to verify that the fuel element was in the correct position and a series of low power tests were run. After the initial modest power run was conducted, and a criticality check was performed. In this test, it was noted that the criticality configuration of the ACRR was modified slightly. Because of this, the NET
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
411
Fig. 13. Instrumented Nuclear Element Test Element.
Fig. 14. Damaged hot frit.
capsule was removed from the reactor and procedures were developed to determine if there was any fuel relocation in the test assembly or a failure of any kind [8]. Unfortunately, it was found that the fuel element had failed. One can note the failure of the hot frit in Fig. 14, and also in Fig. 11. Eventually, after a thorough postmortem, two causes were discovered for the failure. 1. The fuel element initially failed at the compositemetal interface boss due to uncompensated-for thermal expansion mismatches; and 2. The fuel particle bed compliant liner did not prevent bed resettling resulting in large thermal stresses during cyclical runs of the fuel element.
Fig. 12. Nuclear element test assembly with attached fuel element.
This first failure can be readily resolved through more detailed design and attention to the expansion joint interface, however, the second problem will be more recalcitrant. It is generally understood that packed-bed geometries can only accommodate vertical growth of a few times the particle bed thickness. This means that in some fashion or other, either the particles in the bed will need to be fused together to form a more coherent package, or the bed will have to be axially sectioned. There does not appear to be an easy mechanism to implement the first solution, and the second solution would require a major redesign of the fuel element, and it is not clear that a flow pathway that is not heated would not form, resulting in lowered exit temperatures and the potential for unanticipated hot channel factors.
412
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
Unfortunately, the SNTP program was cancelled prior to the proposed modifications to the fuel element to be implemented, although a complete after-action report and analysis was performed.
5. Prototypic fuel element ground testing and environmental impact statement The SNTP program continually looked forward to the time when prototypic testing would be required in order to validate the fuel element to its full capability. The ACRR, while acceptable for initial testing is insufficient to provide the required thermal neutron fluence of 1015 n/cm2 s. Other than in a short pulsed duration mode, the ACRR would not achieve that neutron fluence level. Therefore, a location for ground testing fuel elements, and eventually, a complete nuclear rocket system would be necessary. After a thorough site survey, the program selected the Nevada Test Site, the location of the original NERVA rocket engine test program, although the SNTP program would be tested at a completely different site at the NTS. The Saddle Mountain Test site was selected as appropriate to the program’s needs. PBR in-pile element tester (PIPET): The program identified the need for a subscale reactor, called PIPET as the ideal solution to the initial fuel element test problem. PIPET would be a smaller ∼7 fuel element reactor that would provide the necessary neutron fluence for an extended period of time essential to test the prototypic fuel element. A concept drawing of the PIPET reactor is shown in Fig. 15. The PIPET reactor design was never completed beyond the preliminary design stage, sufficient to complete Title II design of the Saddle Mountain Test Station (SMTS) depicted conceptually below. 5.1. Saddle Mountain Test Station The SMTS was a completely integrated test facility that would allow testing of fuel elements and fully integrated nuclear rocket assemblies such as the one depicted in Fig. 1. The SMTS concept drawing is shown in Fig. 16, below. The SMTS contained a large set of pressurized tanks containing nitrogen, hydrogen and helium, as well as provisions for liquid nitrogen and liquid hydrogen.
Fig. 15. PBR In-Pile Fuel Element Test Reactor (PIPET).
Fig. 16. Saddle Mountain Test Station Layout.
The entire facility was located at the Nevada Test Site, which is a controlled access site exhibiting adequate security for most of the STNP needs. However, due to program security and the quantities of Category I special Nuclear Materials, additional security was required. Therefore, a special, two-row security fence with the latest Intrusion Detection Alarm systems was installed. Fully enriched fuel elements or rocket engine assemblies arrived from the manufacturer at the shipping/receiving assembly building entrance, where QA and other inspections would be performed. If assembly was required, it was performed here. Typically, to minimize shipping requirements, fuel elements for a complete rocket engine would be sent
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
individually or in small lots, and the complete engine would be assembled at this location. Fuel elements would be tested at the PIPET test stand. This stand included a complete effluent scrubber with hydrogen entering the reactor from pressurized hydrogen tanks. The hot exhaust would enter a steam ejection system where it was cooled, and any higher melting point fission products scrubbed from the exhaust. Next, the exhaust would be entrained through a liquid nitrogen scrubber where lower melting temperature fission products such as Iodine would be removed. The exhaust was routed through a set of HEPA filters, and the exhaust gas released to the environment. Any remaining radioactive material that might be released was analyzed through the NEPA process and found to be well within NESHAPs standards. Fully integrated nuclear rocket engines would be tested at the Reactor/Engine Test stand, where a scrubber system equivalent to the PIPET system, only up-scaled would be installed to remove any radioactive materials from full-scale testing. Maximum fuel element testing was slated for up to several hours continuously, while full-scale nuclear rocket engine testing was limited to 1000 s on continuous operation due to NESHAPs limits at the site boundary. SMTS did not progress beyond the Title II design phase due to program cancellation.
413
with a complete scrubber system failure, would result in the source term at the site boundary exceeding NESHAP limitations for radioactive emissions. We determined that we could match our test requirements with 6 to 8–1000 s runs separated by 2 weeks each. In no case did a core disassembly coupled with a scrubber failure exceed the NESHAPS limits. The program, had it proceeded, would have saved tens of millions of dollars by adopting the revised testing strategy, simply by revising a somewhat arbitrarily instituted set of initial test requirements. Re-structuring the test program not only reduced potential costs, but also enhanced public sentiment and ease of EIS approval. We also found ways to quite cost-effectively perform the EIS process. When the SNTP program started, we were informed that the EIS compliance program would cost $70M and require 7 years of effort. We completed a splendid EIS in 2 years for $4M, well under cost and schedule.
6. Summary and conclusion
5.2. Environmental impact statement
Lessons-learned from the SNTP program are many and diverse. While some of the recommendations are PBR specific, the apply equally well to most programs in the space nuclear arena. The most important lessons-learned include the following in no particular order.
Perhaps the watershed event for the SNTP program in an administrative sense was the completion and public hearings on the Ground Test Environmental Impact Statement. The completion of this process with a very favorable consensus for the program was extremely important. While the EIS covered only the ground test, because of the requirement to test beyond operational requirements, of the actual flight unit, the ground test conditions were more severe than normal flight scenarios in terms of run times, hence greater fission product inventory. The EIS process was exceptionally informative, from it we learned that test requirements can, at times, drive the overall program to very costly decisions without substantively adding technical content. In our case, we had originally set the required ground run time as 4–2000 s runs separated by 1 week each run. We discovered that in the case of a catastrophic core disassembly on the fourth run coupled
1. A well-developed nuclear fuel composition and form must exist before committing to a given design. If the fuel is not qualified to meet conditions it must be qualified before major design efforts are initiated. If you do not have fuel, you do not have bupkis. 2. A nuclear rocket requires a good fuel element. Get working on this quickly. 3. Keep the team small and hard-hitting at all times. 4. The program manager must be in charge—if the program works, everyone gets the credit—if it fails the program manager wears the failure around his neck for everyone to see. If he is fair and directed, but you do not like him, TOUGH! 5. The program manager should not try to directly solve technical problems, the technical team is actually much better at that—segregate the problems into areas that they can solve, and problems that
414
6.
7.
8.
9.
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
only you, the program manager can solve. Then solve them. At program reviews do not ask for what is going well, ask about what is broken—you do not need to fix what is not a problem. Unless there is negligence or carelessness, do not blame the technical team for technical difficulties—these are refractory and anfractuous issues. It is your job to get the team focused on solving the problem. Make that decision! As a program manager, you are being paid to make wise engineering judgments with exiguous data available. A monkey can make a decision with all the required data. Generating more paper will not generate more data. Only three things will happen: (1) You make the right decision; everyone wins; (2) the wrong decision—you find out soon enough; (3) no decision everyone and the program loses and money is wasted. Most paper is not a real product, it is a means to a product. EISs and SARs are among the few exceptions. Test data, hardware, experiments and operating systems—these are products. Get the program focused on producing relevant hardware as soon as feasible. Environmental Impact Statements and Safety Analysis Reports are crucial items get them started as soon as feasible.
References [1] R.X. Lenard, A short course on nuclear propulsion systems, Course presentation at IAC Conference in Versailles, Fr 13 May 2002. [2] E.J. Parma, A critical assembly designed to measure neutronic benchmarks in support of the Space Nuclear Thermal Propulsion Program, 10th Symposium on Space Nuclear Thermal Propulsion, Albuquerque, NM, 13 January 1993. [3] G.S. Hoover, R.M. Ball, As-built description and excess reactivity of reference CX core 94WS100, B&W Report ASE 51-3001 774-00, March 1994. [4] R.M. Ball, G.S. Hoover, CX Peek-A-Boo reactivity worth around center stalk, B&W Report ASE-58-3001803-00, November 1992. [5] R.M. Ball, G.S. Hoover, Cold moderator reactivity effects and neutron temperature measurements in a cryogenic hydrogenous moderator, B&W Report ASE-58-3001811-00, January 1994. [6] R.M. Ball, G.S. Hoover, Reactivity measurements of polyethylene moderator rods, bistem control rods, and central hole tantalum and hydrogen in the CX-SNTP, B&W Report ASE-58-3001804-00, October 1994. [7] R.M. Ball, G.S. Hoover, Isothermal temperature coefficient of reactivity in the CX-SNTP, B&W Report ASE-58-3001803-00, November 1994. [8] K.R. Boldt, To cover NET 1.2 Post-Irradiation Examination Experiment Plan. RCSC Meeting Minutes 11 November 1993, Internal Sandia National Laboratories Memorandum.
The advisability of prototypic testing for space nuclear systems Roger X. Lenard Little Prairie Services, USA Available online 23 May 2005
Abstract From October 1987 until 1993, the US Department of Defense conducted the Space Nuclear Thermal Propulsion program. This program’s objective was to design and develop a high specific impulse, high thrust-to-weight nuclear thermal rocket engine for upper stage applications. The author was the program manager for this program until 1992. Numerous analytical, programmatic and experimental results were generated during this period of time. This paper reviews the accomplishments of the program and highlights the importance of prototypic testing for all aspects of a space nuclear program so that a reliable and safe system compliant with all regulatory requirements can be effectively engineered. Specifically, the paper will recount how many non-prototypic tests we performed only to have more representative tests consistently generate different results. This was particularly true in area of direct nuclear heat generation. As nuclear tests are generally much more expensive than non-nuclear tests, programs attempt to avoid such tests in favor of less expensive non-nuclear tests. Each time this approach was followed, the SNTP program found these tests to not be verified by nuclear heated testing. Hence the author recommends that wherever possible, a spiral development approach that includes exploratory and confirmatory experimental testing be employed to ensure a viable design. © 2005 Published by Elsevier Ltd.
1. Background From October 1987 until 1992, the US Department of Defense conducted the Space Nuclear Thermal Propulsion program. This program’s objective was to design and develop a high specific impulse, high thrust-to-weight nuclear thermal rocket engine for upper-stage applications. The selected technology was the Particle Bed Reactor (PBR) as shown in Fig. 1. The PBR employs the high surface-to-volume ratio exhibited by small fuel particles to minimize the
E-mail address: [email protected]. 0094-5765/$ - see front matter © 2005 Published by Elsevier Ltd. doi:10.1016/j.actaastro.2005.03.056
temperature drop within the fuel to operating gas, in this case hydrogen. The Fuel particle is a variant of the TRISO (Triply Isotropic) fuel particle, except in the case of the PBR, the fuel particle outer coating was Zirconium Carbide instead of Silicon Carbide. The fuel particle was also substantially smaller in diameter, in the PBR case, the fuel particle is 0.4 mm in diameter. Fig. 2 shows some of the as-fabricated fuel particles. These fuel particles were placed between an outer flow control assembly and an inner porous liner called a frit. This configuration is displayed as a graphic called “Fuel Element” in Fig. 1. The inner frit was required to withstand the high-temperature exhaust −3000 K and was also required to withstand any thermal or
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
405
Fig. 1. Description of the Particle Bed Reactor.
Fig. 3. Non-irradiated SNTP particle fuel.
2. Developing the fuel particle
Fig. 2. Particle Bed Reactor concept.
mechanical shocks or stresses during startup, operation and shutdown of the rocket engine. The PBR engine maintained an epi-thermal fission neutron spectrum due to the presence of a polyethylene moderator, shown in Fig. 1 as a purple area. The polyethylene moderator is shaped as a right hexagonal parallelpiped with axial grooves along the exterior of the moderator to support hydrogen flow. The polyethylene is cooled by gaseous hydrogen from the turbo-pump assembly. The entire assembly, moderator, cold flow controller, fuel particles and hot frit are assembled into a fuel element that is later assembled in a hexagonal-closepacked assembly that forms the reactor core. A composite pressure vessel, thrust vector control and turbopump assembly are installed to form the entire engine as shown at Fig. 2. Specifications for the engine are shown in the figure as well [1].
The fuel form for the PBR was substantially new. The development of a robust fuel form for Nuclear Thermal Propulsion has been an extremely difficult task. The program decided to create a particle fuel fabrication pilot line at [then] Babcock & Wilcox in Lynchburg, Virginia. We employed a sol-gelation process to convert uranium oxide to uranium carbide (UC). This formed the kernel of the fuel particle. The nominal diameter of the internal fuel kernel was −220 m. The fuel kernels were transferred to a chemical vapor deposition furnace where a thin layer of low-density graphite was placed on top of the UC kernel. The low-density graphite layer is designed to capture fission products generated in the power generating process. On top of that layer a densified layer of pyrolitic graphite was deposited, flowed by a layer of zirconium carbide. A series of un-irradiated fuel particles is shown in Fig. 3. A magnified view of a cross-sectioned fuel particle is shown in Fig. 4. This particle has undergone a nuclear heating test
406
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
Fig. 4. Fuel particle after nuclear in-core heating tests.
and has been cycled to high temperature a number of times. You can note the separation of the lowdensity graphite layer from the fuel kernel. In this figure, one can also note that some of the UC kernel has migrated into the ZrC coating. One can note the consequences of repeated high-temperature operations and substantial numbers of thermal cycles. As may be noted, there is significant grain growth in the heated sample. This grain growth continues until the grain boundaries become columnar in the radial direction. Each grain boundary is a potential source of uranium leakage or carbon depletion into the hot hydrogen fuel. Fuel particles were subjected to a number of tests in both in-core and ex-core heating tests. Results of ex-core heating tests showed good performance of the fuel particles, however, for in-core testing the results were dramatically different. Fig. 5 shows the particle heating test (PHT) apparatus. There were three zones to the test apparatus, each experiencing a different neutron fluence, hence operating power level. Since
Fig. 5. Particle heat test (PHT) in-core test column.
cooling was constant for all particle containers, the temperature at each location was different. Temperature ranged from 2100 to 3100 K. The tests showed that after repeated nuclear heated operations, some of the fuel particles sloughed off layers of materials, including the zirconium carbide coating.
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
407
During the course of the nuclear heating in the Annual Core Research Reactor (ACRR), some of the fuel particles simply turned into dust. This required that the program concentrate on developing a sound fuel form. Several developmental pathways were taken resulting in a mechanism to identify unacceptable particles and sound ones. These processes were used to generate fuel particle coatings with acceptable performance and lifetime for the next series of tests, called the particle in-pile element (PIPE) series of tests.
3. PIPE testing program 3.1. Background The PIPE testing program was designed to take the SNTP system beyond isolated fuel particle testing. The next step was to integrate fuel particles into a stand-alone fuel element that could be tested in the Sandia National Laboratories Annular Core Research Reactor (ACRR). The ACRR exhibits properties highly suited to stockpile stewardship verification. The ACRR is a swimming-pool reactor with a very unique (UO2 –BeO) fuel that testing. It has very high heat capacity, high thermal conductivity and enables very high power (70,000 MW)/high energy (300 MJ) pulses. It can also operate at a steady-state power level of 2 MWth or at 4 MWth for short periods of time. The ACRR has a dry central cavity that can accommodate an experimental package that is up to 9 in–225 mm in outside diameter. The PIPE experiment employed this large experiment package size to develop two different test articles. 3.2. PIPE description The PIPE experimental package featured a fullscale, half-length pre-prototypic fuel element shown at Fig. 6. The fuel element was a radial in-flow fuel element with a bed thickness of approximately 2 cm. The outer channel retainer was fabricated from sintered stainless steel to provide a low-pressure drop hydrogen flow path. The inner hot frit was made from rhenium, external source of neutrons to enable fissions to occur within the fuel bed. This heated the hydrogen flowing over the bed. The ACRR was operated in an extended pulse mode being heated in the fuel volume.
Fig. 6. PBR in-pile experiment (PIPE) test assembly.
The ACRR provided an to provide the required neutron fluence. The experimental setup was designed so that the entire capsule was self-contained. Hydrogen gas was circulated through a small pre-cooled pebble bed and into the fuel element. Heated gas was diagnosed and returned to the pebble bed. This allowed for only a short-duration test sequence.
408
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
Fig. 7. Predicted PIPE hydrogen outlet temperatures.
The program proposed that this test was simply a proof-of-principle test that would allow sufficient information to be generated and analyzed that a more realistic fuel element would be designed. Results from the PIPE-1 experiment were initially promising, in that high-temperature gas was found to flow at the predicted rate with a power density of 3 MW/l, which was at least as high as the highest power density achieved in the NERVA program, but due to limitations of the ACRR, the run was limited to approximately 10 s overall. The predicted performance, which was reasonably closely matched by actual test data is shown in Fig. 7. A second PIPE test conducted in the ACRR was not as successful due to a different material fabrication process used for the rhenium hot frit. The hot frit apparently experienced some initial failures at approximately 2800 K gas temperature and melt began to occur in the hot frit. As a result there was an increase in overall gas output temperature to approximately 3000 K. A post irradiation examination of both fuel elements showed only minor damage in the PIPE1 experiment, but major damage in the PIPE-2 experiment. However, sufficient data had been gathered to proceed with a more robust design of the cold and hot frits as well as continue development of the ZrCcoated fuel particles. In each case, the fuel element was heated thermally by an external source with few problems encountered, therefore, the differences between internal nuclear heating and external thermal heating were once again brought to the forefront of the testing program.
Fig. 8. Particle bed reactor critical experiment.
4. Critical experiment As technology began to advance for the PBR, it was necessary to understand the overall design. Many Monte Carlo neutronics calculations were performed on the SNTP core, which varied from 19 to 37 or to as high as 61 fuel elements in the core depending upon reactor power level and fuel element size. A companion effort attempted to determine pressure drop and hot channel factors associated with a PBR system. However, in order to ensure a reliable design, the MCNP neutronics code runs needed to be benchmarked [2]. To accomplish this, the PBR Critical Experiment was designed and fabricated [3]. There was widespread controversy over this decision at first, because it was believed that analytical tools had achieved a level of performance that verification tests were unnecessary. From start to first operation, the PBRCX took 18 months and cost $19M, including all safety and environmental compliance costs. The development and operation of the PBRCX was a major programmatic achievement. A photo of the PBRCX is shown at Fig. 8. After the initial critical
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
409
Fig. 9. Peek-a-Boo shutter control worth levels. Fig. 10. Platelet Technology Cold Screen.
operation, a wide range of experiments were performed to verify code predictions. During the course of the program, the program manager invented a new type of internal controller called the Peek-A-Book (PABOO) shutter controller. Since it was necessary to determine the utility of such a controller, tests were performed on this and other important benchmark criticality measurements. We determined that the PABOO shutter controller provided the necessary reactivity shift to control the PBR rocket engine. Results of reactivity measurements on the PABOO controller are shown in Fig. 9 [4]. We then decided to measure the temperature coefficient of reactivity. This proved to be a watershed event. In retrospect, those who had opposed the PBRCX on grounds that analytical tools were sufficiently advanced saw that unanticipated interactive effects of the design cannot always be effectively modeled. We discovered that the PBR rocket engine, while having a goal of a negative temperature coefficient in reality exhibited a positive temperature coefficient over a modest temperature range [5,6]. This was a completely unexpected event and resulted in a short stand down for safety purposes when it was discovered that the CX could exhibit a positive temperature coefficient, when all the safety documentation showed (from analysis) that the temperature coefficient would be negative [7]. The results from the PBRCX definitely show the importance of prototypic testing in order to obtain results usable for sound engineering design.
4.1. Advanced fuel elements Based on neutronics, thermal-hydraulic and mechanical analyses, the program acknowledged the need for an advanced fuel element.
4.2. Advanced cold inlet screen The new fuel element would replace the sintered stainless-steel entry screen with an aluminum flow engineered screen employing the Honeywell stacked plate process. This process is commonly known as platelet technology, because thin films of material are photo-etched, cleaned and then stacked upon each other. The final stack is then diffusion-bonded together to form an integrated whole. The SNTP program’s advanced cold screen is shown in Fig. 10. The engineered screen allowed the flow to be enabled in relation to the way power was generated in the fuel bed. Since the reactor exhibited both radial and axial power profiles, in order to maintain a uniform outlet temperature, flow needed to be managed over the entire packed bed. The flow engineered screen allowed this to occur. The cold screen also contained a compliant layer that, at least in theory, would accommodate bed expansion and contraction as the fuel particles were heated or cooled with nuclear heat.
410
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
Fig. 11. Niobium carbide-coated graphite hot frit.
or NET-0. Performing a complete dry-run enabled our team to determine problems with equipment, procedures, safety considerations and in general ensure that the test would proceed smoothly. As will be noted below, some important findings were discovered during the pre-test.
4.3. Hot frit 4.5. NET test capsule description The problems associated with the PIPE-2 experiment required that the program use materials that would not form lower temperature eutectics thereby enhancing probability of failure of the fuel element. For this reason, two developmental pathways were selected. The first pathway was a laser-drilled POCO graphite that would be CVD coated with niobium carbide. A photo of a NbC-coated graphite hot frit is shown at Fig. 11. This hot frit was extracted from the nuclear element test (NET) experiments (to be described later); this particular hot frit was fractured during initial power operations of the NET experiment. While the NbC-coated hot frit performed well from a compatibility perspective, there existed more serious design issues associated with the PBR design itself, some of which were revealed in the NET. 4.4. Nuclear element tests The advancement of component technology was sufficiently compelling that the program determined it was a fortuitous time to integrate a fuel element comprising components that had been specifically developed to counteract issues associated with the PIPE test series. To this end, a NET series fabrication development and test pan was generated. The fuel and the end fittings would be provided by Babcock & Wilcox, the hot frit by Brookhaven National Laboratories, and compliant layer cold screen by Garret Fluid System Division, and the entire unit would be integrated by Babcock and Wilcox. Since nuclear facility operations time is very expensive, the project determined that a precursor test run would be accomplished with simulated fuel particles—this would eliminate considerable security costs during shipment and at Sandia National Laboratories. The fuel element would be full-scale, but half-length. It would be fitted into the identical test device to perform a dry run on the experiment. This fuel element test was called Nuclear Element Test-0
The NET test would be a substantial increase in thermal power (1 MW versus) 3 kW for PIPE. Also, a higher power density in the bed was anticipated ∼3–5 MW/l. This required a substantial amount of cooling if the test objectives were to be achieved. The Net test capsule is shown in Fig. 12. The capsule contains a mounting location for the fuel element assembly, a fan drive motor and fan, cooling channels, a thermal heat sink instrumentation access and various safety related items. The important feature of this capsule was that the heat sink could be cooled to liquid nitrogen temperatures before the test. This would more closely replicate the actual inlet conditions of the hydrogen when the rocket engine would be in operation. The motor and fan drive were of a speciality design built by Skurka engineering. This motor was a several kilowatt motor due to the power requirements to drive the hydrogen through the circuit. The drive motor contained special magnets that, during the pre-test formed a metal hydride and turned to powder, so less capable, but hydrogen compatible magnets were required. This entire capsule was then placed inside a containment capsule that would be placed within the central cavity of the ACRR. Fig. 13 shows the NET fuel element prior to testing. 4.6. NET tests The uranium fueled NET element was shipped to Sandia National Laboratories for testing. The NET-1 fuel element is shown in Fig. 13 in its ready-to-install configuration. The NET capsule was sealed and placed into the ACRR for testing. The initial tests were simply to verify that the fuel element was in the correct position and a series of low power tests were run. After the initial modest power run was conducted, and a criticality check was performed. In this test, it was noted that the criticality configuration of the ACRR was modified slightly. Because of this, the NET
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
411
Fig. 13. Instrumented Nuclear Element Test Element.
Fig. 14. Damaged hot frit.
capsule was removed from the reactor and procedures were developed to determine if there was any fuel relocation in the test assembly or a failure of any kind [8]. Unfortunately, it was found that the fuel element had failed. One can note the failure of the hot frit in Fig. 14, and also in Fig. 11. Eventually, after a thorough postmortem, two causes were discovered for the failure. 1. The fuel element initially failed at the compositemetal interface boss due to uncompensated-for thermal expansion mismatches; and 2. The fuel particle bed compliant liner did not prevent bed resettling resulting in large thermal stresses during cyclical runs of the fuel element.
Fig. 12. Nuclear element test assembly with attached fuel element.
This first failure can be readily resolved through more detailed design and attention to the expansion joint interface, however, the second problem will be more recalcitrant. It is generally understood that packed-bed geometries can only accommodate vertical growth of a few times the particle bed thickness. This means that in some fashion or other, either the particles in the bed will need to be fused together to form a more coherent package, or the bed will have to be axially sectioned. There does not appear to be an easy mechanism to implement the first solution, and the second solution would require a major redesign of the fuel element, and it is not clear that a flow pathway that is not heated would not form, resulting in lowered exit temperatures and the potential for unanticipated hot channel factors.
412
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
Unfortunately, the SNTP program was cancelled prior to the proposed modifications to the fuel element to be implemented, although a complete after-action report and analysis was performed.
5. Prototypic fuel element ground testing and environmental impact statement The SNTP program continually looked forward to the time when prototypic testing would be required in order to validate the fuel element to its full capability. The ACRR, while acceptable for initial testing is insufficient to provide the required thermal neutron fluence of 1015 n/cm2 s. Other than in a short pulsed duration mode, the ACRR would not achieve that neutron fluence level. Therefore, a location for ground testing fuel elements, and eventually, a complete nuclear rocket system would be necessary. After a thorough site survey, the program selected the Nevada Test Site, the location of the original NERVA rocket engine test program, although the SNTP program would be tested at a completely different site at the NTS. The Saddle Mountain Test site was selected as appropriate to the program’s needs. PBR in-pile element tester (PIPET): The program identified the need for a subscale reactor, called PIPET as the ideal solution to the initial fuel element test problem. PIPET would be a smaller ∼7 fuel element reactor that would provide the necessary neutron fluence for an extended period of time essential to test the prototypic fuel element. A concept drawing of the PIPET reactor is shown in Fig. 15. The PIPET reactor design was never completed beyond the preliminary design stage, sufficient to complete Title II design of the Saddle Mountain Test Station (SMTS) depicted conceptually below. 5.1. Saddle Mountain Test Station The SMTS was a completely integrated test facility that would allow testing of fuel elements and fully integrated nuclear rocket assemblies such as the one depicted in Fig. 1. The SMTS concept drawing is shown in Fig. 16, below. The SMTS contained a large set of pressurized tanks containing nitrogen, hydrogen and helium, as well as provisions for liquid nitrogen and liquid hydrogen.
Fig. 15. PBR In-Pile Fuel Element Test Reactor (PIPET).
Fig. 16. Saddle Mountain Test Station Layout.
The entire facility was located at the Nevada Test Site, which is a controlled access site exhibiting adequate security for most of the STNP needs. However, due to program security and the quantities of Category I special Nuclear Materials, additional security was required. Therefore, a special, two-row security fence with the latest Intrusion Detection Alarm systems was installed. Fully enriched fuel elements or rocket engine assemblies arrived from the manufacturer at the shipping/receiving assembly building entrance, where QA and other inspections would be performed. If assembly was required, it was performed here. Typically, to minimize shipping requirements, fuel elements for a complete rocket engine would be sent
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
individually or in small lots, and the complete engine would be assembled at this location. Fuel elements would be tested at the PIPET test stand. This stand included a complete effluent scrubber with hydrogen entering the reactor from pressurized hydrogen tanks. The hot exhaust would enter a steam ejection system where it was cooled, and any higher melting point fission products scrubbed from the exhaust. Next, the exhaust would be entrained through a liquid nitrogen scrubber where lower melting temperature fission products such as Iodine would be removed. The exhaust was routed through a set of HEPA filters, and the exhaust gas released to the environment. Any remaining radioactive material that might be released was analyzed through the NEPA process and found to be well within NESHAPs standards. Fully integrated nuclear rocket engines would be tested at the Reactor/Engine Test stand, where a scrubber system equivalent to the PIPET system, only up-scaled would be installed to remove any radioactive materials from full-scale testing. Maximum fuel element testing was slated for up to several hours continuously, while full-scale nuclear rocket engine testing was limited to 1000 s on continuous operation due to NESHAPs limits at the site boundary. SMTS did not progress beyond the Title II design phase due to program cancellation.
413
with a complete scrubber system failure, would result in the source term at the site boundary exceeding NESHAP limitations for radioactive emissions. We determined that we could match our test requirements with 6 to 8–1000 s runs separated by 2 weeks each. In no case did a core disassembly coupled with a scrubber failure exceed the NESHAPS limits. The program, had it proceeded, would have saved tens of millions of dollars by adopting the revised testing strategy, simply by revising a somewhat arbitrarily instituted set of initial test requirements. Re-structuring the test program not only reduced potential costs, but also enhanced public sentiment and ease of EIS approval. We also found ways to quite cost-effectively perform the EIS process. When the SNTP program started, we were informed that the EIS compliance program would cost $70M and require 7 years of effort. We completed a splendid EIS in 2 years for $4M, well under cost and schedule.
6. Summary and conclusion
5.2. Environmental impact statement
Lessons-learned from the SNTP program are many and diverse. While some of the recommendations are PBR specific, the apply equally well to most programs in the space nuclear arena. The most important lessons-learned include the following in no particular order.
Perhaps the watershed event for the SNTP program in an administrative sense was the completion and public hearings on the Ground Test Environmental Impact Statement. The completion of this process with a very favorable consensus for the program was extremely important. While the EIS covered only the ground test, because of the requirement to test beyond operational requirements, of the actual flight unit, the ground test conditions were more severe than normal flight scenarios in terms of run times, hence greater fission product inventory. The EIS process was exceptionally informative, from it we learned that test requirements can, at times, drive the overall program to very costly decisions without substantively adding technical content. In our case, we had originally set the required ground run time as 4–2000 s runs separated by 1 week each run. We discovered that in the case of a catastrophic core disassembly on the fourth run coupled
1. A well-developed nuclear fuel composition and form must exist before committing to a given design. If the fuel is not qualified to meet conditions it must be qualified before major design efforts are initiated. If you do not have fuel, you do not have bupkis. 2. A nuclear rocket requires a good fuel element. Get working on this quickly. 3. Keep the team small and hard-hitting at all times. 4. The program manager must be in charge—if the program works, everyone gets the credit—if it fails the program manager wears the failure around his neck for everyone to see. If he is fair and directed, but you do not like him, TOUGH! 5. The program manager should not try to directly solve technical problems, the technical team is actually much better at that—segregate the problems into areas that they can solve, and problems that
414
6.
7.
8.
9.
R.X. Lenard / Acta Astronautica 57 (2005) 404 – 414
only you, the program manager can solve. Then solve them. At program reviews do not ask for what is going well, ask about what is broken—you do not need to fix what is not a problem. Unless there is negligence or carelessness, do not blame the technical team for technical difficulties—these are refractory and anfractuous issues. It is your job to get the team focused on solving the problem. Make that decision! As a program manager, you are being paid to make wise engineering judgments with exiguous data available. A monkey can make a decision with all the required data. Generating more paper will not generate more data. Only three things will happen: (1) You make the right decision; everyone wins; (2) the wrong decision—you find out soon enough; (3) no decision everyone and the program loses and money is wasted. Most paper is not a real product, it is a means to a product. EISs and SARs are among the few exceptions. Test data, hardware, experiments and operating systems—these are products. Get the program focused on producing relevant hardware as soon as feasible. Environmental Impact Statements and Safety Analysis Reports are crucial items get them started as soon as feasible.
References [1] R.X. Lenard, A short course on nuclear propulsion systems, Course presentation at IAC Conference in Versailles, Fr 13 May 2002. [2] E.J. Parma, A critical assembly designed to measure neutronic benchmarks in support of the Space Nuclear Thermal Propulsion Program, 10th Symposium on Space Nuclear Thermal Propulsion, Albuquerque, NM, 13 January 1993. [3] G.S. Hoover, R.M. Ball, As-built description and excess reactivity of reference CX core 94WS100, B&W Report ASE 51-3001 774-00, March 1994. [4] R.M. Ball, G.S. Hoover, CX Peek-A-Boo reactivity worth around center stalk, B&W Report ASE-58-3001803-00, November 1992. [5] R.M. Ball, G.S. Hoover, Cold moderator reactivity effects and neutron temperature measurements in a cryogenic hydrogenous moderator, B&W Report ASE-58-3001811-00, January 1994. [6] R.M. Ball, G.S. Hoover, Reactivity measurements of polyethylene moderator rods, bistem control rods, and central hole tantalum and hydrogen in the CX-SNTP, B&W Report ASE-58-3001804-00, October 1994. [7] R.M. Ball, G.S. Hoover, Isothermal temperature coefficient of reactivity in the CX-SNTP, B&W Report ASE-58-3001803-00, November 1994. [8] K.R. Boldt, To cover NET 1.2 Post-Irradiation Examination Experiment Plan. RCSC Meeting Minutes 11 November 1993, Internal Sandia National Laboratories Memorandum.