ML23094A110
ML23094A110 | |
Person / Time | |
---|---|
Site: | MIT Nuclear Research Reactor |
Issue date: | 03/30/2023 |
From: | Foster J, Lau E, Wade M Massachusetts Institute of Technology (MIT) |
To: | Document Control Desk, Office of Nuclear Reactor Regulation |
References | |
Download: ML23094A110 (1) | |
Text
MIT NUCLEAR REACTOR LABORATORY AN MIT INTERDEPARTMENTAL CENTER Edward S. Lau Assistant Director Reactor Operations Mail Stop: NW12-122 138 Albany Street Cambridge, MA 02139 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn.:
Document Control Desk Phone: 617 253-4211 Fax: 617 324-0042 Email: eslau@mit.edu March 30, 2023
Subject:
Annual Report, Docket No. 50-20, License R-37, Technical Specification 7.7.1 Gentlemen:
Forwarded herewith is the Annual Report for the MIT Research Reactor for the period from January 1, 2022, to December 31, 2022, in compliance with paragraph 7.7.1 of the Technical Specifications issued November 1, 2010, for Facility Operating License R-37.
EL/st Sincerely,
.~~.L Marshall B. Wade Reactor Superintendent MIT Research Reactor Edward S. Lau, NE Assistant Director of Reactor Operations MIT Research Reactor John P. Foster Director of Reactor Operations MIT Research Reactor
Enclosure:
As stated cc:
USNRC - Senior Project Manager Research and Test Reactors Licensing Branch Division of Licensing Projects Office of Nuclear Reactor Regulation USNRC - Senior Reactor Inspector Research and Test Reactors Oversight Branch Division of Licensing Projects Office of Nuclear Reactor Regulation
MIT RESEARCH REACTOR NUCLEAR REACTOR LABORATORY MASSACHUSETTS INSTITUTE OF TECHNOLOGY ANNUAL REPORT to United States Nuclear Regulatory Commission for the Period January 1, 2022 ~ December 31, 2022 by REACTOR STAFF
Table of Contents Section Introduction................................................................................................................... 1 A.
Summary of Operating Experience................................................................... 3
- 1.
General................................................................................................. 3
- 2.
Experiments and Utilization................................................................ 4
- 3.
Changes to Facility Design................................................................... 8
- 4.
Changes in Performance Characteristics............................................... 8
- 5.
Changes in Operating Procedures.......................................................... 9
- 6.
Surveillance Tests and Inspections...................................................... 11
- 7.
Status of Spent Fuel Shipment.............................................................. 11 B.
Reactor Operation............................................................................................ 12 C.
Shutdowns and Scrams................................................................................... 13 D.
Major Maintenance ************************************************************************************'.****** 15 E.
Section 50.59 Changes, Tests, and Experiments.............................................. 18 F.
Environmental Surveys..................................................................................... 22 G.
Radiation Exposures and Surveys within the Facility........................................ 23 H.
Radioactive Effluents........................................................................................ 24 Table H-1 Table H-2 Table H-3 Argon-41 Stack Releases.......................................................... 25 Radioactive Solid Waste Shipments......................................... 26 Liquid Effluent Discharges........................................................ 27 I.
Summary of Use of Medical Facility for Human Therapy............................... 28
MIT RESEARCH REACTOR ANNUAL REPORT TO U. S. NUCLEAR REGULATORY COMMISSION FOR THE PERIOD JANUARY 1, 2022 - DECEMBER 31, 2022 INTRODUCTION This report has been prepared by the staff of the Massachusetts Institute of Technology Research Reactor for submission to the United States Nuclear Regulatory Commission, in compliance with the requirements of the Technical Specifications to Facility Operating License No. R-37 (Docket No. 50-20), Paragraph 7.7.1, which requires an annual report that summarizes licensed activities from the 1st of January to the 31st of December of each year.
The MIT Research Reactor (MITR), as originally constructed and designated as MITR-I, consisted of a core of MTR-type fuel, enriched in uranium-235, cooled and moderated by heavy water in a four-foot diameter core tank that was surrounded by a graphite reflector. After initial criticality on July 21, 1958, the first year was devoted to startup experiments, calibration, and a gradual rise to one megawatt, the initially licensed maximum power. Routine three-shift operation (Monday-Friday) commenced in July 1959. The authorized power level for MITR-I was increased to two*megawatts in 1962 and to five megawatts (the design power level) in 1965.
Studies of an improved design were first undertaken in 1967. The concept which was finally adopted consisted of a more compact core, cooled by light water, and surrounded laterally and at the bottom by a heavy water reflector. It is under-moderated for the purpose of maximizing the peak of thermal neutrons in the heavy water at the ends of the beam port re-entrant thimbles and for enhancen;ient of the neutron flux, particularly the fast component, at in-core irradiation facilities. The core is hexagonal in shape, 15 inches across, and utilizes fuel elements which are rhomboidal in cross section and which contain UAlx intermetallic fuel in the form of plates clad in aluminum and enriched to 93% in uranium-235. The improved design was designated MITR-11. However, it retained much of the original facility, e.g.,
graphite reflector, thermal shield, biological shield, secondary cooling systems, containment, etc.
After Construction Permit No. CPRR-118 was issued by the former U.S.
Atomic Energy Commission in April 1973, major components for the modified reactor were procured and the MITR-I completed its mission on May 24, 1974, having logged 250,445 megawatt-hours during nearly 16 years of operation.
2 The old core tank, associated piping, top shielding, control rods and,drives, and some experimental facilities were disassembled, removed, and subsequently replaced with new equipment. After pre-operational tests were conducted on all systems, the U.S. Nuclear Regulatory Commission issued Amendment No. 10 to Facility Operating License No. R-37 on July 23, 1975. After initial criticality for MITR-II on August 14, 1975, and several months of startup testing, power was raised to 2.5 MW in December 1975.
Routine 5-MW operation was achieved in December 1976.
Three shift operation, Monday through Friday, was continued through 1995 when a gradual transition to continuous operation (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, 7 days per week with a shutdown for maintenance every 4-5 weeks) was initiated.
In December 2000, a fission converter medical facility was commissioned.
This facility generated the highest quality epithermal beam in the world for use in the treatment of certain types of cancer, and could again be made available.
From mid-April through mid-September 2010, all major piping in the primary and secondary coolant systems was replaced and upgraded. This included a titanium heat exchanger (replacing the three previous primary heat exchangers) anc;l the major instrumentation sensors that monitor system flows, temperatures, and pressures.
On November 1, 2010, NRC approved the relicensing of the reactor for 6-MW operation through November 1, 2030.
Reactor power was increased in small increments from 5 MW for observations and data collection, and reached 5.8 MW on April 23, 2011. Routine 5.8 MW operation began on May 25, 2011.
On December 4, 2019, NRC approved the licensing of a new digital nuclear safety system.
After an NRC-approved postponement due to the *nationwide COVID-19 public health emergency, implementation was completed in September 2020. The reactor was returned to full power on September 16, 2020, with the new system in service.
The current operating mode is generally continuous operation just under 6 MW when needed, with a maintenance shutdown scheduled every calendar quarter.
This is the forty-eighth annual report required by the Technical Specifications, and it covers the period from January 1, 2022, through December 31, 2022.
Previous reports, along with the "MITR-II Startup Report" (Report No. MITNE-198, February 14, 1977) have covered the startup testing period and the transition to routine reactor operation.
This report covers the forty-sixth full year of routine reactor operation, now at the 6-MW power level. It was another year in which the safety and reliability of reactor operation met and exceeded requirements and expectations.
A summary of operating experience and other activities and related statistical data are provided in Sections A through I of this report.
3 A.
SUMMARY
OF OPERATING EXPERIENCE
- 1.
General The MIT Research Reactor, MITR-11, is operated at the MIT Nuclear Reactor Laboratory (NRL) to facilitate experiments and research including in-core irradiations and experiments, neutron activation analyses, and materials science and engineering studies such as neutron imaging. It is also used for student laboratory exercises and student operator training, and education and outreach programs. Additionally, the reactor has been used for industrial production applications and other irradiation services. When operating, the reactor is normally maintained at slightly below 6 MW.
For CY2022, the nominal full power was 5.7 MW, with an operating period of up to eleven weeks at a time, followed by a scheduled outage lasting about two weeks or more for reactor and experiment maintenance, protective system surveillance tests, and other necessary outage activities. The reactor would then be re-started to* full power and maintained there for another operating period.
Throughout CY2022, the reactor averaged 77 operating hours per week, compared to 102 hours0.00118 days <br />0.0283 hours <br />1.686508e-4 weeks <br />3.8811e-5 months <br /> per week for CY2021, 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> per week for CY2020, 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> per week for CY2019, and 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per week for CY2018. The lower average for CY2020 was the result of extended shutdowns for the nationwide COVID-19 public health emergency, and for installation of the new digital nuclear safety system in the control room.
The reactor was operated throughout the year with 24 fuel elements in the core.
The remaining three positions were occupied by either solid aluminum dummies or in-core experiments. During CY2022 compensation for reactivity lost due to burnup was provided by four refuelings. These followed standard MITR practice which is to introduce fresh fuel to the inner portion of the core where peaking is least (normally the B-Ring) and to place partially spent fuel into the other portions of the core. In addition, fuel elements were inverted and rotated so as to achieve more uniform burnup gradients in them. Twelve new fuel elements were introduced into the reactor core and ten spent fuel elements were discharged from the core tank (reactor core plus wet-storage ring) during CY2022.
- The MITR-11 fuel management program remains quite successful. During the period of CY2022, one shipment totaling 8 spent fuel elements was returned to an off-site DOE facility.
As in previous years, the reactor was operated throughout the period without the fixed hafnium absorbers.
- Again in CY 2022, a 13th element was received from BWXT but was sent back after the on-site inspection deemed it necessary to return it. This element was not counted toward the 12 fresh and 10 discharged element totals mentioned for the year.
4
- 2.
Experiments and Utilization The MITR-11 was used for experiments and irradiations in support of research, training, and education programs at MIT and elsewhere. In-core experiments operated in all four reactor cycles of CY2022 from a combination of National Scientific User Facilities (NSUF) clients, National Laboratory, and Industrial entities.
This has included use of both the high temperature pressurized water loop (HTWL) and inert gas facilities.
Extended reactor outages in Q3 and Q4 resulted in some projects experiencing reduced irradiation times. Irradiations and experiments conducted in CY2022 include:
a) An NSUF-funded project led by PI Joe Palmer at Idaho National Laboratory supported the NPI inert gas irradiation for one cycle. This project intends to further the development of in-core neutron detectors.
In particular this irradiation included new fast-response self-powered neutron detectors developed by INL with rhodium emitters (SPND-Rh).
The sensors were irradiated in a single large titanium capsule with a graphite insert, instrumented along the core length with thermocouples.
The experiment operated in different regimes between 550 and 800°C using inert gas control and some holds at intermediate reactor power levels during planned power level changes while sensors were interrogated in real time.
b) The HTWL "Boone" irradiation that was installed mid-way through CY2021
- Cycle 4 was re-installed for one cycle of additional irradiation. This is an experiment designed to simultaneously irradiate two sets of specimens, SiC fiber composite coupons from Free Form Fibers and high-purity zirconium crystals from Idaho National Laboratory. Both samples are dry (not in contact with the HTWL water flow) but take advantage of the isothermal conditions within the loop at its normal PWR operating condition of 565°F. In order to reduce the neutronic impact of the HTWL it was operated under pure water conditions (no soluble chemistry additions and with helium gas saturation).
c) The XEN0l inert gas vehicle was irradiated for one cycle and then removed to accommodate other experiments. It is planned to be re-installed in. the reactor in CY2023 for a second cycle. The vehicle consists of four capsules targeting temperatures between 300 and 725°C with a variety of metal dogbones, ceramic disks, and surrogate TRISO particle specimens, with thermocouples located at various points in each capsule to monitor temperatures.
The specimens represent structural and neutron moderator materials of interest from a variety of sponsors including X-energy, NASA, and CTP.
d) Two cycles of the SFX inert gas irradiation were completed, sponsored by NSUF and led by PI Joshua Daw at INL. This project is investigating the use of sapphire optical fibers for distributed temperature measurements.
The irradiation included fibers previously irradiated at KSU with Li-6 coatings as well as commercial silica fibers for control. The fiber leads incorporated new sealed and armored junctions to allow extensions to be fitted near the reactor
5 lid for ease of installation.
The fibers were irradiated at 675°C while undergoing on-line laser-based interrogation, with thermocouples located at various axial positions for monitoring and control.
e) The first of two (non-consecutive) cycles of irradiation were completed for the NSUF-sponsored project, led by PI Zilong Hua at INL to study the deployment of optical-fiber assisted photo-thermometry sensors in reactor environments.
The HPR experiment consisted of seven gold-coated silica optical fibers -
three interrogation/sensing pairs and one continuous
- 1oop for control/transmission loss monitoring. The irradiation operated up to 450°C in inert gas, with additional planned temperature variations preempted by the early reactor Q4 outage. The fiber pairs were used to guide a laser onto three SiC and AhO3 targets within the experiment, and the surface thermal emissions were then guided to a detector. The laser is modulated and the shift in the resulting signal over time and at various temperatures is used to infer changes in the target thermal conductivity.
f)
The neutron diffractometer/neutron imaging beamline is operational.
This instrument has been used to conduct neutron imaging and tomography measurements of batteries and fuel cells to support users from the NIST Center of Neutron Research whose experiments were canceled due to the long shutdown of the NIST reactor. These measurements involved an MIT UROP student. The instrument is also used to certify control blades for MITR. An ongoing project funded by the DOE in 2021 utilizes this beamline to demonstrate a novel polychromatic diffractometer. In addition, we completed feasibility studies of a small-angle neutron scattering (SANS) facility at this beamline. The feasibility study and strong MIT faculty support for SANS helped to win MIT's internal competition for an NSF MRI proposal, which is currently in preparation. In parallel, there is a proposal for a simultaneous neutron/X-ray imaging facility to be built at this beamline.
g) The MIT graphite exponential pile (MGEP) was re-started several years ago by Professor Kord Smith with the support of NRL staff and other MIT Nuclear Science & Engineering (NSE) faculty members. It has since been used for teaching and research. A DOE-NE funded research project used the graphite pile to conduct experiments in support of demonstrating autonomous control of a subcritical system. The facility is an ideal testbed due to its inherent safety characteristics and modular construction. These allow in-pile instrumentation and pulley mechanisms to be installed without significantly modifications to the facility.
h) Additionally, reactor staff conducted a series of lectures and a remote demonstration of operation of the MGEP.
The demonstration included operational characteristics of the MGEP as a subcritical facility, a walkthrough of experimental procedures for student lab work, and an overview of an ongoing DOE project to explore first-of-a-kind autonomous control systems utilizing machine learning and neural networks.
6 i) The student spectrometer (4DH1) has been used throughout the year to support remote teaching and demonstration of neutron properties. NSE students are using it during the 22.09/22.90 course "Principles of Nuclear Radiation Measurement and Protection". In addition, it is used for demonstrations for Course 16 (Aero/Astro).
j) Elemental analyses were performed using NAA on samples of in-core experimental components prior to their use in research projects described above and several of the irradiations described below for experiment safety reviews. Analyses were performed on various samples for the MIT NSE 22.01 "Introduction to Nuclear Engineering and Ionizing Radiation" class as well as initial analyses on rice samples for an ongoing research project of NSE professors Haruko Wainwright and Michael Short.
k) Irradiations of Yb-176 targets for Lu-177 production were completed in 2PH1 for Shine Medical Technologies.
- 1) Activation of uranium foils in the 3GV6 facility for detector calibration at the Los Alamos National Laboratories and Ciambrone Laboratory at Patrick AFB.
m) Irradiations of experimental neutron detectors and target foils in the Thermal Neutron Beam for Los Alamos National Labs.
n) Activation of polymer samples and standard reference materials for further NAA studies for University of Alabama.
o) Activation and NAA of silicon oxide samples for Serva Technology.
p) Activation and NAA of various samples in support of MIT course 22.01 "Introduction to Nuclear Engineering and Ionizing Radiation".
q) Neutron and gamma sensor irradiations in the Thermal Neutron Beam for the MIT Mechanical Engineering Department in support of ingestible sensor project.
r) Researchers from CF Technologies, Inc. (Hyde Park, MA) and the NRL completed the irradiations and chemical processing tasks for a Phase II US-DOE SBIR to develop a new technology for the purification of the medical isotope 177Lu. The research included two twenty-day 2PH1 irradiations and one one-day 3GV6 irradiation. The primary irradiation target is ytterbium enriched in 176Y which produces 177Y that decays into the medically useful isotope 177Lu. The researchers used Nal and HPGe detectors to measure that the 177Lu was quantitatively recovered and separated from the unwanted radioisotope 175Y. Additional analyses are being conducted to verify that the separated 177Lu was sufficiently free of other impurities to meet the requirements for a medically useful product.
7 s) A series of irradiations have been conducted in 2PH1 and 3GV6 for the first phase of the ORNL-supported Advanced Manufactured Fuel Irradiation (AMFI) program. This project is supporting the development of the ORNL Transformational Challenge Reactor (TCR). A final irradiation of the initial batch ofTCR fuel compacts for this project was conducted in 3GV6, with three SiC matrix compacts with 7% enriched uranium nitride (UN) TRI~O particles and utilizing the new 3GV6 shutter and cask system.
After controlled temperature ramps, this irradiation reached 750°C for 24-hours using a combination of nuclear heating, gas control, and electrical heaters.
The irradiation completed successfully and multiple cover gas samples revealed no fission gas release or indication of fuel failure. The irradiated fuel compacts, which had been sent to INL for additional pre-irradiation NDE characterization, will be returned to ORNL for PIE.
t) Other use of the reactor took place for training MIT student reactor operators.
Additionally, the recently commissioned reactor simulator was used to demonstrate reactor power changes for MIT nuclear engineering classes (course 22.01 "Introduction to Nuclear Engineering and Ionizing Radiation",
- and course 22.011 "Seminar in Nuclear Science and Engineering").
An ongoing initiative is the partnership with the Department of Energy's Nuclear Science User Facilities (NSUF) for advanced materials, high temperature sensors, and fuel irradiation. The MITR became the first university research reactor to be a partner facility with the NSUF starting in 2008. MIT-NRL staff also worked with INL staff to jointly develop advanced reactor instrumentation, and reviewed NSUF's user proposals.
8
- 3.
Changes to Facility Design Except as reported in Section E, no changes in the facility design were made during this calendar year. The nominal uranium loading of MITR-11 fuel is 34 grams of U-235 per plate and 510 grams per* element (manufactured by BWXT).
Performance of these fuel elements has been excellent.
The loading results in 41.2 w/o U in the fuel meat, based on 7% voids, and corresponds to the maximum loading in Advanced Test Reactor (ATR) fuel.
Two hundred sixty-six elements fabricated by BWXT have been received, forty-one of which remain in use. One has been removed because of suspected excess out-gassing, another because it was dropped, and two were returned to BWXT without being placed in-core due to not meeting on-site quality assurance inspection criteria. Two hundred twenty-one have been discharged because they have attained the fission density limit.
The MITR is actively involved in studies for future use of low enrichment uranium (LEU) in the MITR, partially supported by the Reduced Enrichment for Research and Test Reactors (RERTR) Program at DOE. These studies principally focus on the use of monolithic U-Mo fuels with uranium densities in: excess of 15 g/cm3 ( compared with 1.5 g/cm3 for UAlx fuel), currently under development by the RERTR Program. Although initial studies show that the use of these fuels is feasible, conversion of the MITR-11 to lower enrichment must await the final successful qualification of these high-density fuels. In October 2018, NRC accepted a report entitled "Low Enriched Uranium (LEU) Conversion Preliminary Safety Analysis Report for the MIT Research Reactor (MITR)" supporting a future application for licensing to convert from High Enriched Uranium (HEU) to LEU fuel.
- This PSAR provides analysis determining that a power increase from 6 MW with the current HEU core to 7 MW when using the LEU core is required in order to maintain core neutronic flux performance.
- 4.
Changes in Performance Characteristics Performance characteristics of the MITR-11 were reported in the "MITR-11 Startup Report."
Minor changes have been described in previous reports.
Performance characteristics of the Fission Converter Facility were reported in the "Fission Converter Facility Startup Report", and in the FY2006 report which described a 20% improvement in the intensity of the unfiltered epithermal neutron beam. In CY2012, fuel was removed from the fission converter tank. The tank will remain unfueled pending resumption of epithermal beam research. In CY2013, the D2O coolant was removed from the fission converter system and replaced with demineralized light water. The D2O was put into on-site storage for future use.
9
- 5.
Changes in Operating Procedures With respect to operating procedures subject only to MITR internal review and approval, and not covered in Section E of this report, a summary is given below of changes implemented during CY2022.
a) PM 3.6 "Waste Tank Discharge Procedure" was updated to clarify filter pore size (0.35 µm) and improve the formatting of the procedure. (SR #2019-11) b) PM 3.1.1.1 "Full Power Startup Checklist - Two Loop Mechanical" was updated to incorporate current formatting and equipment, particularly a new cooling tower filtration system. Also, secondary system startup sections were rewritten to allow for a slow temperature equalization between the cooling towers, primary, and secondary systems. In the primary system startup section, operators are given the option of conducting either a visual or audible check of the anti-siphon and natural circulation valves, rather than being instructed to perform a visual check when possible, thus reducing crane usage and dose exposure without reducing the effectiveness of the checks. (SR #2019-14) c) AOP 5.8.16 "Spill of Heavy Water" & AOP 5.6.3 "Trouble Radiation Monitor" received language directing the operator to emergency procedure EOP 4.4.4.12 "Containment Evacuation", if personnel evacuation is needed. (SR #2019-34) d) AOP 5.5.17 "Low Pressure Instrument Air" was updated to remove follow-up actions that referenced a dedicated system air compressor, which is no longer in service. (SR #2020-18) e) PM 6.4.17 "Leak Alarm" test procedure was updated to incorporate current leak zones and current best practices, and add a diagram of the system. (SR #2020-29) f) AOP 5.8.4 "Loss of Normal Off-Site Electrical Power" was revised to more accurately reflect the behavior of the NW12 vault alarm and to incorporate several administrative designations. (SR #2020-30) g) PM 3.1.6 "Restart Following an Unanticipated Shutdown or a Brief-Duration Scheduled Shutdown" was updated to align the procedure with current equipment, formatting, and best practices. The step referencing use of a computer program for calculating estimated critical position was removed, as the program is no longer in service. (SR #2021-9) h) PM 6.1.4.7 "Shim Blade Drop Time" test procedure was established to better formalize the shim blade drop time testing process and incorporate it into the procedure manual. (SR #2021-12) i) PM 6.5.19 "Calibration of Test Equipment and Tools" was updated with a new step to perform, for any piece of test equipment that needed adjustments by the calibration lab, a spot-check of the most recently-tested associated reactor instrument. (SR #2021-13)
10 j) PM 6.5.16.lA "Regulating Rod Calibration" &
PM 6.5.16.2 "Shim Blade Calibration" were revised to reflect the relabeling of Channel 9 as Channel 7 and to improve clarity concerning what data needs to be recorded at which step. It also removed references to specific channel chart values that vary over time, instead directing operators based on corresponding reactor power levels. Use of fixed power levels is better for clarity, and further ensures compliance with the I 0kW upper power limit during the course of the procedures.
(SR#2021-15B and SR#2021-15A) k) "Special Procedure for Dummy Element Transfer from SFP to Core Tank" was revised to update references to legacy equipment. To make it suitable for all future dummy transfers between the spent fuel storage pool (SPF) and the core tank, specific references to the "AFTR" experiment dummy were removed: Also, the transfer procedure was broken into sections covering two different methods of transferring a dummy - either using the fuel transfer cask or performing a free air transfer in the basket. For a free air transfer, steps were added to require prior approval from the Radiation Protection Officer and recorded measurements to verify the dummy dose rate. (SR #2021-16)
11
- 6.
Surveillance Tests and Inspections There are many written procedures in use for surveillance tests and inspections required by the Technical Specifications. These procedures provide a detailed method for conducting each test or inspection and specify an acceptance criterion which must be met in order for the equipment or system to comply with the requirements of the Technical Specifications. Thirty such tests and inspections are scheduled throughout the year with a frequency at least equal to that required by the Technical Specifications. Together with those not required by Technical Specifications, over 100 tests and calibrations are conducted by Reactor Operations on an annual, semi-annual, or quarterly basis.
Other surveillance tests are done each time before startup of the reactor if shutdown exceeds 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, before startup if a channel has been repaired or de-energized, and at least quarterly; a few are on different schedules. Procedures for such surveillance are incorporated into daily or quarterly startup, shutdown, or other checklists.
During this reporting period, surveillance :frequencies have been at least equal to those required by the Technical Specifications, and the results of tests and inspections were satisfactory throughout the year for Facility Operating License No. R-37.
- 7.
Status of Spent Fuel Shipment In CY2022, there was one shipment made to reduce the inventory of spent fuel at MIT. These shipments are made using the BEA Research Reactor (BRR) package.
The U.S. Department of Energy has indicated that further shipments will be feasible in CY2023 for future fuel discharges.
12 B.
REACTOR OPERATION Information on energy generated and on reactor operating hours is tabulated as follows:
Calendar Quarter 1
2 1
3 4
Total J 1. Energy Generated (MWD):
a) MITR-II 281.9 (MIT CY2022) 232.2 361.4 73.6 949.1 (normally at 5. 7 MW) b) MITR-II 42,766.6 (MIT FY1976-CY2021) c) MITR-I 10,435.2 (MIT FY1959-FY1974) d) Cumulative, 54,150.9 MITR-I & MITR-II
- 2. MITR-II Operation (hours):
(MIT CY2022) a) At Power (2:: 0.5-MW) for 1207 982 1544 310 4043 Research b) Low Power
(< 0.5-MW) for 24 6
6 2
38 TrainingO) and Test c) Total Critical 1231 988 1550 312 4081 (l)
These hours do not include reactor operator and other training conducted while the reactor is at or above 0.5 MW. Such hours are included in the previous line (row 2a of the table).
13 C.
SHUTDOWNS AND SCRAMS During this reporting period, there were two inadvertent automatic scrams and four other unscheduled shutdowns.
The term "inadvertent automatic scram" in this section refers to shutting down of the reactor through protective system (nuclear safety or process system) automatic engineered action when the reactor is at power or at least critical; the reactor operator is not involved in the scram action.
The term "other unscheduled shutdown" typically refers to an unscheduled power reduction to subcritical initiated manually by the reactor operator in response to an abnormal condition indication. For such shutdowns, the reactor operator may manually use a "minor scram" (fast control blade insertion by gravity) or a "major scram" (fast control blade insertion plus reflector dump and containment building isolation), among other possible actions.
An example of another type of "other unscheduled shutdown" is a reactor shutdown due to loss of off-site electrical power, because the reactor protective system action was not the cause of the shutdown. An incidental control blade drop is likewise considered an "other unscheduled shutdown",
because such drops lower the reactor power rapidly, and require the console operator to manually bring the reactor to full shutdown condition.
The following summary of inadvertent automatic scrams and other unscheduled shutdowns is provided in approximately the same format as for previous years in order to facilitate a comparison.
- 1.
Nuclear Safety System Scrams a)
None.
- 2.
Process System Scrams a)
Trip from Low Pressure MP-6A / Low Flow Primary Coolant scrams caused by noisiness in primary flow indication.
Total 0
Subtotal 0
2 Subtotal 2
14
- 3.
Other Unscheduled Shutdowns a)
Shutdown caused by fluctuation of off-site electrical power.
1 b)
Minor scram initiated by operator upon Shim Blade #6 dropping from its magnet.
1 c)
Shutdown because of excessive cycling of D2O gasholder.
1 d)
Shutdown to investigate unexpected trend in readings of DWK 250 Channels #2 & #3.
1 Subtotal 4
Total 6
- 4.
Experience during recent years has been as follows:
Calendar Year 2022 2021 2020 2019 2018 2017 2016 2015 2014 2013 2012 Nuclear Safety and Process System Scrams 2
2 2
3 1
1 4
8 13 4
6
15 D.
MAJOR MAINTENANCE Major reactor maintenance projects performed during CY2022 are described in this Section. These were planned and performed to improve safety, reliability and efficiency of operation of the MIT Research Reactor, and hence improve the reliability of the reactor operating schedule and the availability of the reactor for experiments, research and training purposes. Additionally, Reactor Operations staff performed safety reviews for all reactor experiments and their operating procedures. The staff also provided support for installations and removals of reactor experiments, and monitored key performance data from the experiments during reactor operations.
For continuous support of neutron transmutation doping of silicon, reactor staff performed routine irradiation and shipping activities. There is an annual external audit to review the program for maintaining the ISO 9001 Certification. Preventive maintenance on conveyor machinery, such as alignment of conveyor carriages, was performed during major outages.
Major maintenance items performed in CY2022 are summarized as follows:
Date Maintenance Descri:gtion 1/3/2022 Removed HTWL-Boone and NETL in-core experiments 1/3/2022 Completed annual inspection of in-core components and fuel 1/5/2022-Replaced Shim Blade 6 blade, electromagnet, and drive 1/14/2022 1/6/2022 Completed annual ECCS test and calibration 1/8/2022 Completed 13.8 kV substation and Motor Control Center (MCC) cleaning and inspection 1/8/2022 Replaced Reactor Lighting 480 V breaker 1/18/2022 Replaced Primary Ion Column 1/19/2022-Replaced MF-lB Primary System flowrate transmitter 1/20/2022 1/24/2022 Installed NPI in-core experiment 1/25/2022 Installed HTWL-Boone in-core experiment 2/7/2022-Replaced Shim Blade 6 electromagnet 2/8/2022
16 4/4/2022-Removed HTWL-Boone and NPI in-core experiments 4/5/2022 4/6/2022 Replaced Shim Blade 5 proximity switch 4/7/2022-Completed biennial Containment Building pressure test 4/8/2022 4/11/2022-Replaced Motor Control Center 1 (MCC-1) 4/22/2022 4/21/2022 Replaced Primary Ion Column 4/26/2022-Performed drilling and structural modifications on Fission 5/10/2022 Converter Room roof 5/2/2022 Replaced Shield and D20 Reflector Ion Columns 5/4/2022 Completed annual polar crane maintenance and inspection 5/10/2022-Removed large debris and old reactor shielding/structures from 5/12/2022 Restricted Area backyard 5/16/2022 Installed XEN0 1 in-core experiment 5/18/2022 Replaced DWK #4 safety system channel fission chamber 5/20/2022 Replaced D20 Helium System gasholder with low pressure regulator 7/5/2022 Removed XEN0l in-core experiment 7/5/2022 Replaced Shim Blade 5 proximity switch 7/7/2022 Completed annual Emergency Battery Transfer Test 7/8/2022 Replaced Shim Blade 5 proximity switch and tube 7/8/2022 Completed annual Charcoal Filter Efficiency Test 7/11/2022-Replaced HV-3B and repaired Secondary System Bernoulli filters 7/12/2022 7/11/2022-Replaced Fission Converter supply transformer 7/12/2022 7/13/2022 Replaced Primary Ion Column I
7/19/2022 Installed SFX in-core experiment
17 7/20/2022 Repaired ML-3 Core Tank level float switch 8/11/2022 Installed DAK instrument for Channel 5 linear flux indication 10/3/2022 Removed SFX in-core experiment 10/5/2022-Replaced Shim Blade 1 blade, electromagnet, and drive 10/6/2022 10/12/2022 Replaced Motor Control Center 2 (MCC-2) 11/3/2022 10/12/2022 -
Replaced VFDs and cabling for Primary System pumps MM-1/lA 11/3/2022 and Secondary System pumps HM-A/B 11/4/2022 Replaced Primary Ion Column 11/8/2022-Repaired Containment Ventilation intake damper hydraulic 11/11/2022 actuator 11/14/2022-Installed SFX and HPR in-core experiments 11/15/2022 11/16/2022 Installed HTWL-FCI in-core experiment 11/23/2022 Replaced Primary Ion Column 12/12/2022 Shutdown to investigate abnormal Nuclear Safety System indications. Investigation revealed Primary System leakage into reactor shielding areas, including vertical ports housing nuclear detectors.
12/15/2022 Removed HTWL-FCI in-core experiment 12/19/2022 Removed SFX and HPR in-core experiments Many other routine maintenance and preventive maintenance items were also scheduled and completed throughout the calendar year.
18 E.
SECTION 50.59 CHANGES, TESTS, AND EXPERIMENTS This section contains a description of each change to the reactor facility and associated procedures, and of the conduct of tests and experiments carried out under the conditions of Section 50.59 of 10 CFR 50, together with a summary of the safety evaluation in each case.
Changes that affect only the operating procedures and that are subject only to MITR internal review and approval, including those that were carried out under the provisions of 10 CFR 50.59, are similarly discussed in Section A.5 of this report.
The review and approval of changes in the facility and in the procedures as described in the SAR are documented in the MITR records by means. of "Safety Review Forms". These have been paraphrased for this report and are identified on the following pages for ready reference if further information should be required with regard to any item. Pertinent pages in the SAR have been or are being revised to reflect these changes.
The conduct of tests and experiments on the reactor are normally documented in the experiments and irradiation files.
For experiments carried out under the provisions of 10 CFR 50.59, the review and approval is documented by means of the Safety Review Form. This includes all in-core experiments, which are. additionally reviewed and approved by the MIT Reactor Safeguards Committee (MITRSC) prior to installation in the reactor core. All experiments not carried out under the provisions of 10 CFR Part 50.59 have been done in accordance with the descriptions provided in Section 10 of the SAR, "Experimental Facilities".
19 Advanced Cladding Irradiation Facility (ACD \\ High Temperature Water Loop SR #0-06-4 (04/03/2006), SR #0-06-6 (05/18/2006), SR #2015-8 (05/22/2015),
SR #2015-9 (05/22/2015), SR #2017-20 (4/01/2019)
An in-core experiment loop was installed on May 22, 2006, to investigate the effects at various stages of irradiation on specimens of silicon carbide intended for use in advanced fuel cladding designs. Its envelope of operating conditions is very similar to that of previous in-core experiments such as the Zircaloy Corrosion Loop and the Electro-Chemical Potential Loop.
No new safety issues were raised.
Operation continued until October 2007. A second advanced cladding loop, designated ACI-2, operated in core from March 2009 through mid-December 2009, March to April 2010, December 2010 through June 2011, from October 2011 to July 2012, and from August through October 2013. A later version of this loop, designated the Westinghouse Accident-Tolerant Fuel (WATF) experiment, was installed in 2014 and operated until May 2015, and again from December 2015 until July 2016. The latter run featured a stepped thimble to minimize neutron streaming to the reactor top. Additionally, from May 2015 to August 2015, the facility was used to test an In-Core Crack Growth Measurement (ICCGM) system. In 2017, from January to June, the ACI facility was used for the COATI irradiation ("CTP and ORNL Accident Tolerant Irradiation") of a variety of silicon carbide composite materials. From August 201 7 through the first quarter of 2021, it was used for W ATP Phase 2 and Exelon experiments. In later 2021 and 2022, it saw dry samples - SiC fiber composite coupons from Free Form Fibers and zirconium crystals from Idaho National Laboratory-along with one cycle offast-response, self-powered neutron detectors from INL in 2022.
Heated In-Core Sample Assembly Experiment (ICSA)
SR #0-04-19 (12/01/2004), SR #M-04-2 (12/30/2004), SR #0-05-11 (07/22/2005),
SR #M-09-1 (07/30/2009), SR #M-09-2 (12/11/2009), SR #0-10-2 (03/28/2010),
SR #0-12-17 (06/04/2012), SR #0-12-19 (07/09/2012), SR #2017-6 (7/02/2019),
SR #2017-6A (05/03/2017)
High-temperature sample capsules were used with the redesigned titanium 2" ICSA tube to provide a heated irradiation environment for the specimens within.
These capsules include gamma-heating susceptors similar in principal to the High Temperature Irradiation Facility. No new safety issues were raised. An alternate 16" plug was designed and installed in the reactor top shield lid to allow simultaneous use of the ICSA and the ACI-2 in-core experiments. The ICSA operated in core from December 2009 through April 2010, from August 2010 to January 2012, from April to July 2012, and from mid-September through October 2013 for various sample irradiations using heated and unheated capsules.
The MIT Reactor Safeguards Committee (MITRSC) approved two ICSA Safety Evaluation Report amendments in early 2013 to allow the 2013 irradiation of molten fluoride salt in-core using a nickel capsule inside the ICSA. The ICSA facility remained in regular use in CY2021 for in-core experiments and irradiations. - See section A.2 (Experiments and Utilization),
items (c), (d), and (e).
20 Physical Security Plan Revisions SR #0-13-16 (05/12/2014), SR #0-13-30 (12/24/2013), SR #2014-19 (11/07/2014),
SR #2014-23 (02/18/2015), SR #2015-5 (01/23/2015), SR #2017-5 (2/14/2017),
SR #2019-7 (06/11/2019), SR #2019-9 (09/27/2019), SR #2021-2 (01/25/2021)
SR #2021-2A (04/12/2021), SR #2021-2C (04/28/2021)
MITRSC approval for the revised Plan was granted per the Security Subcommittee meeting of 6/6/2013. It was then submitted to NRC as a License Amendment Request, and approved by NRC in 2014. In 2015, a security alarm coincidence monitoring system was installed to provide local and remote notification should the weekend alarm or an intrusion alarm become deactivated during periods of unattended shutdown. Procedures were revised to incorporate use of this monitoring system. In 201 7, the Plan was revised in response to an NRC Request for Additional Information (RAI) regarding incorporation of material from NRL's responses to NRC Compensatory Action Letters. The revision and response to NRC were approved by the MITRSC Special Subcommittee for Security. In 2018, further modifications to the Plan were proposed as a followup to the RAI, and were reviewed and approved by the MITRSC in October 2018. These proposed modifications were discussed with NRC during a routine inspection in December 2018.
In May 2019, all proposed modifications to the Plan and associated security procedures were presented to the MITRSC Security Subcommittee, including proposed changes to AOP 5.8.22 "Loss or Degradation of a Security System", in accordance with new regulatory guidelines that were incorporated into the Security Plan. The Subcommittee approved the modifications, and the Plan was submitted to NRC on 6/11/2019. On 7/29/2019, NRC was satisfied with the update as being in compliance with IO CFR 73 and incorporating all of the site-specific compensatory measures to which MIT had committed. NRC then closed Confirmatory Action Letter (CAL)
No. NRR-02-005 which had been issued in 2002 in response to the 9/11 national emergency.
In CY2021, conversion from the C*CURE security management system to the Genetec system being adopted throughout the MIT campus was implemented for the reactor facility. It included, for compatibility with the new system, replacement of the iris readers with other biometric readers, with a corresponding Physical Security Plan revision sent to NRC in April 2021, shortly after the completion of the upgrade. Other security devices were either replaced or retrofitted with external interfaces to make them compatible with the new system.
A comprehensive system-wide test was performed immediately afterwards, and again in CY2022, proving the conversion successful.
In CY2022, the NRL worked with DOE-PNNL for grant funding to upgrade the security camera system for the reactor. A contract was awarded and accepted by MIT in September 2022. Due to supply issues, implementation has been postponed until mid-CY2023.
21 Stack Effluent & Water Monitor Project SR #2015-30 (pending), SR #2015-30A (12/02/2015), SR #2015-30B (07/08/2016),
SR #2015-30C (03/31/2016), SR #2015-30£ (04/21/2017)
As part of a project to install new stack effluent monitors and secondary water monitors using detectors located outside the containment building, a new 1-1/4" diameter piping penetration was installed on the south side of the containment building, about four feet below ground. It was tested as satisfactory per existing procedures for pressure-testing new penetrations. Until such time as it is connected to the main system piping, the new piping will remain blank-flanged, or isolated and tagged out, in order to ensure containment integrity is maintained. A new climate-controlled shed, the "stack monitor shed", was constructed in the reactor's back yard in CY2016, with the two new stack monitor stations fully mounted within. In CY2019 through CY2022, this newly-installed system continued to operate in parallel with the existing stack effluent and water monitoring systems.
D20 Helium System Modifications SR #2022-10 (05/20/2022), SR #2022-18 (pending), SR #2022-19 (pending)
The D20 reflector helium cover gas system was upgraded to eliminate the gasholder as the method to control routine supply and relief of the helium cover gas.
The gasholder was replaced with a high-flow, low-pressure regulator with an integral overpressure relief port. The cover gas system's overpressure and underpressure safety components were not affected by this change. The regulator's output setting of two inches of water pressure matches the previous gasholder blanket pressure. Various startup, shutdown, testing, and calibration procedures were modified for use with the new system.
22 F.
ENVIRONMENTAL SURVEYS Environmental monitoring is performed using continuous radiation monitors and passive dosimetry devices (TLD). The radiation monitoring system consists of detectors and associated electronics at each remote site with data transmitted continuously to the Reactor Radiation Protection office and recorded electronically in a database. The environmental monitoring remote sites are located within a quarter mile radius of the facility. The calendar year totals per sector, due primarily to Ar-41, are presented below. The passive TLDs were in place at all times throughout the year and are exchanged quarterly.
Site North East South West Exposure (01/01/2022 - 12/31/2022) 0.57 mrem 0.90mrem 0.46mrem 0.90mrem Calendar Year Average 2022 0.7mrem 2021 0.2mrem 2020 0.2mrem 2019 0.2 mrem 2018 0.2mrem 2017 0.4mrem 2016 0.6 mrem 2015 0.4 mrem 2014 0.8 mrem 2013 0.2 mrem 2012 0.3 mrem
23 G.
RADIATION EXPOSURES AND SURVEYS WITHIN THE FACILITY A summary of radiation exposures received by facility personnel and experimenters is given below:
January 1, 2022 - December 31, 2022 Whole Body Exposure Range (rems)
Number of Personnel No measurable..........................................................................................
Measurable - < 0.1...................................................................................
0.1 0.25 0.25 0.50 0.50 -
0.75 0.75 1.00 1.00 1.25 1.25 -
1.50 1.50 1.75 1.75 2.00 91 28 8
0 1
0 0
0 0
0 Total Person Rem= 2.19 Total Number of Personnel= 128 From January 1, 2022, through December 31, 2022, the Reactor Radiation Protection program provided radiation protection services for the facility which included power and non-power operational surveillance (performed on daily, weekly, monthly, quarterly, and other frequencies as required), maintenance activities, and experimental project support. Specific examples of these activities included, but are not limited to, the following:
- 1.
Collection and analysis of air samples taken within the containment building and in the exhaust/ventilation systems.
- 2.
Collection and analysis of water samples taken from the secondary, D2O, primary, shield coolant, liquid waste, and experimental systems, and fuel storage pool.
- 3.
Performance of radiation and contamination surveys, radioactive waste collection and shipping, calibration of area radiation monitors, calibration of effluent and process radiation monitors, calibration of radiation protection/survey instrumentation, and establishing/posting radiological control areas.
- 4.
Provision of radiation protection services during fuel movements, in-core experiments, sample irradiations, beam port use, ion column removal, diffractometer beam testing, etc.
The results of all surveys and surveillances conducted have been within the guidelines established for the facility.
24 H.
RADIOACTIVE EFFLUENTS This section summarizes the nature and amount of liquid, gaseous, and solid radioactive wastes released or discharged from the facility.
- 1.
Liquid Waste Liquid radioactive wastes generated at the facility are discharged only to the sanitary sewer serving the facility. The possible sources of such wastes during the year include cooling tower blowdown, the two on-site liquid waste storage tanks, and one controlled sink in the Restricted Area (Engineering Lab). All of the liquid volumes are measured, by far the largest being the 11,841,087 liters discharged during CY2022 from the cooling towers. (Other large quantities of non-radioactive waste water are discharged to the sanitary sewer system by other parts of MIT, but no credit for such dilution is taken because the volume is not routinely measured.)
Total activity less tritium in the liquid effluents (cooling tower blowdown, waste storage tank discharges, and engineering lab sink discharges) amounted to 5.91E-5 Ci for CY2022.
The total tritium was l.13E-1 Ci.
The total effluent water volume was 11,857,212 liters, giving an average tritium concentration of 6.865E-6 µCi/ml.
The above liquid waste discharges are provided on a monthly basis in the following Table H-3.
All releases were in accordance with Technical Specification 3.7.2.1, including Part 20, Title 10, Code of Federal Regulations. All activities were substantially below the limits specified in 10 CFR 20.2003 "Disposal by Release into Sanitary Sewerage".
Nevertheless, the monthly tritium releases are reported in Table H-3.
- 2.
Gaseous Waste Gaseous radioactivity is discharged to the atmosphere from the containment building exhaust stack. All gaseous releases likewise were in accordance with the Technical Specifications and 10 CFR 20.1302, and all nuclides were substantially below the limits, using the authorized dilution factor of 50,000 (changed from 3,000 starting with CY2011 per the renewed license's Technical Specifications). The only principal nuclide was Ar-41, which is reported in the following Table H-1. The 1977.79 Ci of Ar-41 was released at an average concentration of 2.95E-10 µCi/ml.
This represents 2.95% of EC (Effluent Concentration (lE-08 µCi/ml)).
- 3.
Solid Waste Three shipments of solid waste were made during the calendar year. The information pertaining to these shipments is provided in Table H-2.
25 TABLE H-1 ARGON-41 STACK RELEASES CALENDAR YEAR 2022 Ar-41 Average Discharged Concentration(l)
(Curies)
(µCi/ml)
January 2022 14.18 2.33E-11 February 285.64 5.86E-10 March 454.17 7.45E-10 April 0.00 0.00E+00 May 85.74 l.76E-10 June 407.35 6.68E-10 July 121.82 2.50E-10 August 220.36 4.52E-10 September 310.70 5.lOE-10 October 0.00 0.00E+00 November 0.00 0.00E+00 December 77.62 l.27E-10 Totals (12 Months)<2) 1977.79 2.95 E-10 EC (Table II, Column I) 1 X lQ-S
%EC 2.95%
(1) Average concentrations do not vary linearly with curies discharged because of differing monthly dilution volumes.
(2) Last decimal place may vary because of rounding.
26 TABLE H-2
SUMMARY
OF MITR-II RADIOACTIVE SOLID WASTE SHIPMENTS CALENDAR YEAR 2022 Descriptions Volume 22.5 ft3 Weight 461 lbs.
Activity 2.3 mCi Date of shipment May 19, 2022 Waste processor Toxco Material Management Center, Oak Ridge, TN Waste broker Ecology Services Inc., Columbia, MD Disposition to licensees for burial Energy Solutions, Clive, UT Volume 110 ft3 Weight 2793 lbs.
Activity 82mCi Date of shipment October 26, 2022 Waste processor Toxco Material Management Center, Oak Ridge, TN Waste broker Ecology Services Inc., Columbia, MD Disposition to licensees for burial Energy Solutions, Clive, UT Volume 54 ft3 Weight 288 lbs.
Activity 2mCi Date of shipment December 7, 2022 Waste processor Toxco Material Management Center, Oak Ridge, TN Waste broker Ecology Services Inc., Columbia, MD Disposition to licensees for burial Energy Solutions, Clive, UT
Jan.2022 Feb.
Mar.
Apr.
May June July Aug.
Sept.
Oct.
Nov.
Dec.
12 months 27 TABLE H-3 LIQUID EFFLUENT DISCHARGES CALENDAR YEAR 2022 Total Total Volume Activity Tritium of Effluent Less Tritium Activity Water(l)
(xl0*6 Ci)
(mCi)
(liters)
NDA<2) 0.0306 95865 5.253 23.3 1138472 NDA<2) 0.279 1906586 NDA<2) 0.0192 103765 NDA<2) 0.0564 812057 25.885 50.9 2212212 13.029 9.63 992546 7.849 2.90 2749263 NDA<2) 0.00433 771011 NDA<2) 0.00764 17075 NDA<2>
0.0128 101742 7.040 25.6 956618 59.057 113 11857212 Average Tritium Concentration (xlQ*6 µCi/ml) 0.319 20.509 0.146 0.185 0.069 23.015 9.700 1.053 0.006 0.447 0.126 26.743 6.860 (1) Volume of effluent from cooling towers, waste tanks, and NW12-139 Engineering Lab sink. Does not include other diluent from MIT estimated at l.0x107 liters/day.
(2) No Detectable Activity (NDA): less than l.26xl Q*6 µCi/ml beta for each sample.
28 I.
SUMMARY
OF USE OF MEDICAL FACILITY FOR HUMAN THERAPY The use of the medical therapy facility for human therapy is summarized here pursuant to Technical Specification No. 7.7.1.9.
- 1.
Investigative Studies Investigative studies remain as summarized in the annual report for FY2005.
- 2.
Human Therapy None.
- 3.
Status of Clinical Trials The Phase I glioblastoma and melanoma trials with BIDMC have been closed.
A beam that is superior to the original epithermal beam in the basement Medical Therapy Room in both flux and quality could again be made available from the Fission Converter Facility. No use of that beam is anticipated in the near term because of a nationwide funding hiatus for work of this type.