ML25098A159

Annual Report for the Mit Research Reactor for the Period from January 1, 2024, to December 31, 2024

ML25098A159
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	03/31/2025
From:	Foster J, Lau E MIT Nuclear Reactor Laboratory
To:	Office of Nuclear Reactor Regulation
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Download: ML25098A159 (1)

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Edward S. Lau Assistant Director Reactor Operations MIT NUCLEAR REACTOR LABORATORY AN MIT INTERDEPARTMENTAL CENTER Mail Stop: NW12-122 138 Albany Street Cambridge, MA 02139 Phone: 617 253-4211 Fax: 617 324-0042 Email: eslau@mit.edu March 31, 2025 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn.:

Document Control Desk

Subject:

Annual Report, Docket No. 50-20, License R-37, Technical Specification 7.7.1 Forwarded herewith is the Annual Report for the MIT Research Reactor for the period from January 1, 2024, to December 31, 2024, in compliance with paragraph 7.7.1 of the Technical Specifications issued November 1, 2010, for Facility Operating License R-37.

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Enclosure:

As stated C

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Edward S. Lau, NE Assistant Director of Reactor Operations MIT Research Reactor 0

Director of Reactor Operations MIT Research Reactor 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, 2024 - December 31, 2024 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....................................................... 10

7.

Status of Spent Fuel Shipment.............................................................. 10 B.

Reactor Operation............................................................................................ 11 C.

Shutdowns and Scrams................................................................................... 12 D.

Major Maintenance........................................................................................... 14 E.

Section 50.59 Changes, Tests, and Experiments.............................................. 17 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, 2024 - DECEMBER 31, 2024 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-1, 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 enhancement 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 U Alx intermetallic fuel in the form of plates clad in aluminum and enriched to 93% in uranium-235. The improved design was designated MITR-II. 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-11 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. It is currently undergoing major renovation to transform it into another experimental irradiation space called the MCube or "meter-cubed".

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) and 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 normally continuous operation just under 6 MW when needed, with a maintenance shutdown scheduled every calendar quarter.

This is the fiftieth annual report required by the Technical Specifications; it covers the period from January 1, 2024, through December 31, 2024.

Previous reports, along with the "MITR-11 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-eighth full year of safe reactor operation and maintenance activities.

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 OPERA TING EXPERIENCE

1.

General The MIT Research Reactor, MITR-II, 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.

In CY2024 the reactor was successfully made critical on January 23rd, 2024, following an extended maintenance outage to identify and repair a primary system leak.

Necessary activities prior to return to normal operation were performed over the first part of the year, including shielding replacement, instrument calibration, and personnel requalification. The reactor returned to nominal full power operation at 5.7 MW in May. Due to full time employee staffing shortages resulting from the long period without operation for licensing exams and training activities, and the reduced number of student operators available over the summer, for Q3 the reactor operated on a modified schedule of six days at full power per week, with a shutdown on Saturdays and startup on Sundays. The reactor returned to continuous operation for the final operating period of the year.

In the latter half of CY2024, the reactor operated for an average of 91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br /> per week, compared to no operation for CY2023, 77 operating hours per week for CY2022, 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 with 24 fuel elements in the core. The remaining three positions were occupied by either solid aluminum dummies or in-core experiments. During CY2024 compensation for reactivity lost due to bumup was provided by five refuelings, three of which 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 bumup gradients. Nine new fuel elements were introduced into the reactor core and six spend elements were discharged from the core into the wet storage ring. The other two refuelings were to relocate partially burned elements back to the core tank after temporary storage in the SFP during the leak repair.

The MITR-II fuel management program remains quite successful. During the period of CY2024, one partial shipment totaling 6 spent elements was returned to an offsite DOE facility.

As in previous years, the reactor did not operate with any fixed hafnium absorbers.

4

2.

Experiments and Utilization With the resumption of regular full-power reactor operations in May, CY2024 saw the return of major irradiation experiments to MITR. The in-core irradiation schedule is saturated with both pre-2022 outage projects that had remaining irradiation commitments as well as many new projects that have been awaiting the opportunity to start testing.

Work at the reactor also included experiment preparation, post-irradiation examination, and in support of training and education programs.

Experiment work conducted in CY2024 includes:

a) The XEN02 inert gas vehicle began irradiation in May and completed irradiation at the end of December after three cycles.

This experiment follows on from the XEN0 1 irradiation completed in 2022 and consists of three capsules, each containing three sub-capsules.

Each capsule was designed for a set irradiation temperature of 300, 500, or 800°C and contained a variety of ceramic and metallic samples.

The specimens represent reactor structural and neutron moderator materials of interest from a variety of sponsors including X-energy and the ONWARDS and GAMOW programs. After irradiation the sub-capsules will be distributed to other labs for post-irradiation examination of the samples.

b) The WHEP thermosiphon irradiation began in May in an in-core dry position and also completed at the end of December after three cycles.

This irradiation vehicle contained a single sodium-filled thermosiphon built at Idaho National Lab and designed to operate with a hot-side (evaporator) temperature of 800°C.

The thermosiphon moves heat from the in-core evaporator region to the above-core condenser region through passive circulation of the sodium.

Two zones of electric heaters provided supplementary heating as needed, for instance to prevent sodium freezing at low reactor power. Due to uncertainty about the dynamic startup behavior of the thermosiphon, there was initially concern about the integrity of the thermosiphon seal. However, subsequent testing and analysis enabled by support from the MITRSC In-Core Subcommittee has shown that the experiment was functioning normally, but with higher heat losses above core than expected, and sodium circulation was occurring as expected. At the end of irradiation the thermosiphon will be shipped for post-irradiation examination.

c) The ATFB-1 and ATFB-2 irradiations occurred in the May-June, July-September, and October-December MITR cycles in the High Temperature Water Loop (HTWL) facility.

These irradiations are sponsored by the "Understanding of ATF Cladding Performance under Radiation using MITR" NSUF IRP project, which includes approximately ten cycles of in-core irradiations in the water loop and dry positions through 2026 occupying at least 50% of the in-core space each. These include Accident Tolerant Fuel (ATF) cladding samples from vendors such as Framatome, Westinghouse, and GE-Hitachi, as well as international university-produced cladding samples. The October cycle also contained dry, passively-loaded spring specimens for X-energy. The current tests have been conducted under BWR

5 NWC conditions, and samples will undergo PIE at MIT to assess corrosion behavior versus exposure.

There are samples placed in three general locations in these experiments: in-core, above core (predominantly gamma flux), and out-of-core (water chemistry only).

d) The JAEA-sponsored project "Light-Water Reactor Fuels Cladding Testing in the MIT Research Reactor" (JSiC) is also utilizing the HTWL facility under BWR conditions. Similar to the ATF IRP project, advanced ceramic cladding samples from vendors Toshiba and Hitachi-GE are being studied for corrosion behavior in BWR NWC conditions (with careful attention to dissolved oxygen levels) over a series of four cycles of irradiation through 2026, and with periodic intermediate PIE and sample change-outs. The start of this irradiation campaign was delayed due to the long reactor maintenance outage, and in compromise on schedule, some irradiations of the JAEA samples and ATF IRP samples were combined in both the May-June and July-September cycles.

Samples are undergoing PIE at MIT to assess corrosion and to inform future sample test matrices.

e) The CALOR-1 in-core irradiation was conducted over a one-week period of dedicated reactor use in the HTWL. The water loop was operated at very low temperature and pressure to provide a consistent thermal environment for a two-cell calorimeter and a gamma thermometer provided and operated by researchers from the Aix-Marseilles University.

The calorimeter and thermometer assembly was actuated axially within the water loop via a tubular support, similar to past actively-loaded water loop experiments. The experiment collected data on the nuclear heating rate at various axial positions and at varying reactor power levels within the B3 core position.

This data will be used in development of the calorimeter concept and to inform nuclear heating calculations for future MITR in-core experiments.

f) Two NSUF-sponsored dry irradiation experiments, the second cycle of the INL-led photothermometry (HPR) project and the irradiation of the Kairos Power-led FLiBe salt infiltration and crack-growth capsules (FS-6) were planned for 2024 (with HPR-2 fully constructed) but have been delayed into early 2025 to accommodate delays due to the long reactor maintenance outage and reduced reactor capacity factor due to operator shortages.

g) The neutron diffractometer/neutron imaging beam.line is operational at 4DH4 beam.line. This instrument has been used since August 2024 after the reactor shutdown. We have finished a project funded by the DOE in 2021 that utilized this beam.line to demonstrate a novel polychromatic diffractometer.

The imaging instrument has been used by MIT and outside users to measure Li batteries (a company C4V and a NIST/University of Maryland collaboration), and ZrH and YH materials for an MIT project funded by the DOE, Office of Nuclear Energy. In addition, we are conducting feasibility studies (funded by DOE in CY2024) of a compact powder diffractometer at the vertical-beam facility.

These projects involved multiple MIT UROP students.

In parallel, we prepared and submitted a proposal for a simultaneous neutron/X-ray imaging facility to be built at this bearnline.

6 h) 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 significant modifications to the facility.

i) The student spectrometer (4DH1) has been in a very limited use throughout the year during the reactor shutdown. This spectrometer normally supports remote teaching and demonstration of neutron properties. It has been used for physics demonstrations for MIT Course 16 (Aero-Astro).

j) Students in the MIT NSE class 22.01 "Introduction to Nuclear Engineering and Ionizing Radiation" were given a demonstration of gamma spectroscopy and on-site lectures about neutron activation analysis. The HPGe systems were also used in an ongoing NAA research project of a post-doctoral researcher of Professors Haruko Wainwright and Michael Short.

k) The NRL's Neutron Activation Analysis facilities were used to support MITR in-core research projects and research at Commonwealth Fusion Systems, Montclair State University, and the MIT Department of Materials Science and Engineering.

1) Additional irradiations in NRL pneumatic facilities included: activation of sapphire samples and standard reference materials in 2PH1 for NAA for the University of Alabama, activation in 2PH1 and NAA of YH and ZrH materials in support of the DOE imaging project at 4DH4 as well as student research using the irradiated materials for the MIT Nuclear Engineering Department.

m) Three major 3GV6 irradiations were conducted in 2024 utilizing the expanded 3GV6 cask and shutter capabilities previously established as part of the ORNL TCR/AMFI program. Two week-long irradiations of natural convection salt loops (one loaded with FLiNaK and one with FLiBe salts) are described here. These GEMINA-1 and GEMINA-2 tests were sponsored by the ARPA-E GEMINA program and intended to demonstrate the feasibility of flowing salt experiments in 3GV positions, as well as gather data on the transport of surrogate corrosion and fission products in flowing salt with thermal gradients. They operated with a peak temperature around 700°C via two-zone electrical heating. The GEMINA-2 test was the first to utilize the new, dedicated MITR 3GV data acquisition and control system; 3GV experiments can now be operated independently of the in-core experiment systems. This will facilitate the expected increased cadence and complexity of 3GV experiments. After irradiation these salt loops are undergoing non-destructive gamma scanning in the NRL hot cells.

7 n) The third major 3GV6 irradiation experiment was supported by two NASA-funded projects, one led by INL and the other by Little Prairie Services. This experiment is the initial demonstration phase of thermal nuclear reactor (for space propulsion) materials in 3GV. Previously moderator materials and fuel kernels and coated fuel particles have been irradiated under pure hydrogen in the 2PH1 pneumatic system. Now that testing has expanded to a capsule within the 3GV6 position.

A single highly-instrumented capsule was irradiated over several hours on two different days in a series of planned power ramp-and-hold tests with both inert gases and hydrogen.

This experiment demonstrated many new capabilities in 3GV6: concentrated nuclear heating using a boron nitride target (to 400°C), electrical actuation of an irradiation capsule within 3GV6 during full-power operation combined with a large thermal neutron shield to allow rapid power ramping, operation of the irradiation capsule under a pure hydrogen environment, and extension of the 3GV6 irradiation vehicle to allow utilization of the entire length of 3GV6 (all previous major 3GV6 irradiations have been limited to the top half of the core). This experiment is intended as a capability demonstration and prelude to irradiation tests still in the conceptual phase, including some utilizing pressurized hydrogen, higher temperatures, and fissile materials.

o) The 3GV6 vertical irradiation facility was used for activations of LEU targets for Los Alamos National Laboratory p) The Thermal Neutron Beam was utilized for irradiation of cell culture samples for Draper Laboratories.

q) The reactor simulator was used for training MIT student reactor operators with the reactor shut down for a portion of the year. It was also 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") as well as other visiting groups.

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 worked with INL staff to jointly develop advanced reactor instrumentation, promote the use of U.S.

irradiation facilities for materials testing and instrumentation development, and review NSUF's user proposals for various funding opportunity announcements throughout the year.

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 seventy-five elements fabricated by BWXT have been received, forty-three of which remain in use. One has been removed because of suspected excess out-gassing, two because they were 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-seven 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 CY2024.

a) PM 6.5.6.3 "System Pressure Gauge Calibration" was updated to add a caution to verify the SP-2 gauge pressure reads zero before disconnecting any of its fittings.

This improves safety, because of the lack of isolation valves at that gauge's location. (SR #2021-22) b) PM 6.1.6 "Monthly Technical Specification Tests" was updated to record the air flow velocity indication for the pressure relief system in meters per second (velocity) rather than inches of water (pressure).

This improves safety by providing the specification in a more accurate unit of measure. (SR #2021-24) c) "MIT Nuclear Reactor Logbook Changes" covered the update to version 2.00 of the software for the digital console logbook. The new version added features that were desired after four years of experience with the original software. The new features included safety improvements, such as checks for personnel licenses and medical exams being current. Other updates modernized cybersecurity functions and included capture of items from the hardcopy Job Workbook. The system backup processes remained unchanged. All of the new software was run on an independent platform to test it prior to installation. (SR #2020-31) Under "Digital Logbook Upgrade", the Job Workbook was maintained in hardcopy and digitally in parallel for over a year, and then at the start of CY2024, the hardcopy was retired. (SR #2022-21) d) PM 3.8. lA "Startup of D.I. System in Standby (Recirculation) Mode" for the make-up water system was updated to correct two typographical errors and to remove the previously-added "Reviewed by Supervisor" line. The corrections for accuracy and streamlining were evaluated as improving safety. (SR #2023-4) e) PM 6.6.1.4 "Communications Links Test" was revised to include additional personnel lists, to ensure timely update of all titles, responsibilities, and contact information. (SR #2023-5) f) PM 6.1.3.9 "Quarterly Emergency Battery Surveillance" was updated with regards to the acceptance criteria for battery cell voltages. The update provided more accurate acceptance criterion without changing the original intent of the procedure.

(SR #2023-8)

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 CY2024, 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 CY2025 for future fuel discharges.

11 B.

REACTOR OPERATION Information on energy generated and on reactor operating hours is tabulated as follows:

Calendar Quarter 1

2 1

3 4

Total

11. Energy Generated (MWD):

a) MITR-II 26.9 208.2 205.0 289.0 729.1 (MIT CY2024)

(normally at 5.7 MW) b) MITR-II 43,715.7 (MIT FYI 976-CY2023) c) MITR-1 10,435.2 (MIT FY1959-FY1974) d) Cumulative, 54,880.0 MITR-1 & MITR-II

2. MITR-II Operation (hours):

(MIT CY2024) a) At Power (2: 0.5-MW) for 329 931 926 1285 3471 Research b) Low Power

(< 0.5-MW) for 145(2) 6 30 35 216 Training(l) and Test c) Total Critical 474 937 956 1320 3687 (I)

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).

(2)

Higher number of low power critical hours in CY2024 Q 1 in support of instrumentation recalibration following reactor defueling and repair work for primary leak.

12 C.

SHUTDOWNS AND SCRAMS During this reporting period, there were two inadvertent automatic scrams and three 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)

While Channel #4 was out of service, Channel # 1 tripped on high power, at maximum of 100 kW per other channels, during repositioning testing of its detector.

2.

Process System Scrams a)

Two groups of personnel operated both doors of the basement airlock at the same time.

Total 1

Subtotal 1

1 Subtotal 1

13

3.

Other Unscheduled Shutdowns a) b)

c)

Shutdown caused by significant fluctuation of off-site electrical power.

Shutdown upon observation of decreasing readings from Channel #3.

Shutdown to replace regulating rod drive linkage, upon observation of reactivity fluctuations when regulating rod movements were attempted.

1 1

1 Subtotal 3

Total 5

4.

Experience during recent years has been as follows:

Calendar Year 2024 2023 2022 2021 2020 2019 2018 2017 2016 2015 2014 2013 Nuclear Safety and Process System Scrams 2

0*

2 2

2 3

1 1

4 8

13 4

• Reactor remained shut down for maintenance throughout CY2023.

14 D.

MAJOR MAINTENANCE Major reactor maintenance projects performed during CY2024 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 silicon conveyor machinery, such as alignment of conveyor carriages, was performed during major outages.

Major maintenance items performed in CY2024 are summarized as follows:

Date Maintenance Descri12tion For reactor restart, background neutron counts were insufficient at the full power detector positions and as such all vertical and horizontal detectors were inserted to their closest to core positions.

These positions, along with new cabling, required multiple 1/M 1/02/2024-startup procedures to rebuild neutron population and provide 2/23/2024 adequate detector readings for future detector relocation and reactor restarts. After multiple 1/M startups, the four nuclear safety system fission chambers were installed in their full power positions and associated test and calibration procedures were performed before operating at full power.

While perform blade drop times, the blade drop timer failed.

1/02/2024 Cause determined to be a blown fuse. The fuse was replaced and blade drop timer returned to service the same day.

1/02/2024-Shim Blade #3 80% proximity switch failed. It was replaced and 1/04/2024 retested satisfactorily 1/04/2024.

1/17/2024 During performance of Scram Time Checks, MP-6A was found out of specification. Repaired and returned to service the same day.

2/01/2024 Shim Blade #2 "Drive In" light replaced.

2/08/2024 Adjustments made to Shim Blade #4 position indicator for subcritical interlock position.

15 2/08/2024 Adjustments made to Shim Blade #5 position indicator for full-in position indication.

2/26/2024 Replaced fission chamber for Channel #3 due to degraded cable.

4/01/2024-Secondary system partially drained in support ofleak repair.

4/05/2024 Secondary system refilled and retested satisfactorily 4/05/2024.

4/04/2024 Compressed air leak repaired for 1PH1/2PH1 pneumatic tube system. Leak was from a pressure regulator.

4/05/2024 Shim Blade #2 position indication at reactor top repaired and retested satisfactorily.

Shim Blade #4 showed intermittent failure while attempting to shim in. Operation returned to normal following a regular reshim.

4/10/2024 Reactor shutdown on 4/11/2024 to repair Shim Blade #4 drive.

Shear pin was found broken and replaced. Shim Blade #4 was retested satisfactorily and returned to service.

During refuel, a piece of lockwire was dropped into the core tank.

4/22/2024-4/26/2024 Through a multi-day effort, the lockwire was located and removed via a wet-vac used to vacuum the bottom of the core tank.

4/23/2024 Check valve replaced for intake ventilation damper hydraulic pump.

4/26/2024 Shim Blade #5 seal was replaced and retested satisfactorily.

4/30/2024-Core tank and wet storage ring were thoroughly cleaned utilizing 5/01/2024 an underwater wet-vac.

5/20/2024 Cable for Linear Flux Channel #7 replaced to eliminate signal n01se.

1PH1-NW13 auto transfer valve failed. Optical sensor replacement 5/30/2024 on 6/10/2024 corrected issue, however stop pin for rabbit insertion remains broken. Stop pin repaired/retested: 11/21/2024 6/07/2024 Linear Flux Channel #7 detector shifted up in vertical port.

While operating, the Regulating Rod did not respond to control 6/11/2024-signals. Reactor was shutdown, and investigation revealed a 6/12/2024 linkage from the drive mechanism was broken. Linkage replaced and retested satisfactorily.

16 PM 6.4.12 - High Temperature D2O Cleanup System sensor failed 7/08/2024 during testing. Repairs made and tested satisfactorily on 7/18/2024.

7/08/2024 PM 6.4.23 - Low Pressure Helium Supply failed during testing.

Repairs made and tested satisfactorily on 7/22/2024.

MIT Facilities confirmed that AC unit #1 for the reactor building 7/24/2024 has failed and parts are on order for repairs, unit #2 remains operational.

7/25/2024 Shim Blade #2 Selsyn wiring was entirely rebuilt.

Shim Blade #5 proximity switch for blade-in indication found to be 8/26/2024 malfunctioning. Switch replaced on 10/02/2024 and tested satisfactorily on 10/16/2024.

PM 6.3.6 -Trouble Radiation Monitors Alarm and Interlock was 9/23/2024 unsatisfactory due to failure of alarms to actuate. Alarm function was rebuilt and verified functioning on 9/24/2024.

Stack base damper wire rope broke while performing a test 9/25/2024 procedure. Wire rope repaired on 9/26/2024 and the test was then performed and completed satisfactorily.

Commenced phase 1 of utility room electrical equipment 10/07/2024 replacement. Power correction factor unit will remain secured until second phase of installation.

Two effluent radiation monitors were found out of specification 11/13/2024 during a test. They were recalibrated the same day and the test was completed satisfactorily.

11/20/2024 Waste Tank discharge valve was rebuilt to prevent waste water leak-by during discharge.

Electrical power secured to the entire NW12 building to tie in new 12/23/2024 breaker panel in the utility room. Power was restored the same day and all systems recovered to normal shutdown alignment.

Many other routine maintenance and preventive maintenance items were also scheduled and completed throughout the calendar year.

The upgrade to the fire protection alarm system in the containment building has continued through CY2024.

17 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".

18 Advanced Cladding Irradiation Facility (ACI) \\ 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 2017 through the first quarter of 2021, it was used for W ATF 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 of fast-response, self-powered neutron detectors from INL in 2022. The High Temperature Water Loop facility remained in regular use in CY2024 for in-core experiments and irradiations. - See section A.2 (Experiments and Utilization), items (c), (d), and (e).

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 principle 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 CY2024 for in-core experiments and irradiations. - See section A.2 (Experiments and Utilization),

items (a), (b), and (c).

19 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), QA #2022-25 (in progress)

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 2017, 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 10 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-chain issues, installation was postponed until mid-CY2023. The system was commissioned on June 28, 2023, and was accepted as satisfactory in an Assurance Site Visit by PNNL project managers on August 16, 2023.

In CY2024, Reactor Operations submitted quarterly reports on the system to PNNL.

20 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 #2015-30E (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 CY2024, this newly-installed system continued to operate in parallel with the existing stack effluent and water monitoring systems.

D2O Helium System Modifications SR #2022-10 (05/20/2022), SR #2022-18 (03/09/2023), SR #2022-19 (12/10/2022),

SR #2023-10 (11/30/2023)

The D2O 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, operating, testing, and calibration procedures were modified and fine-tuned for use with the new system.

21 Identification and Repair of Primary System Leak SR #2023-1 (01/27/2023), SR #2023-2 (09/08/2023), SR #2023-3 (12/18/2024),

QA #2023-9 (08/07/2023)

The MIT Reactor was shut down on December 12, 2022, in response to indications of a water leak from the primary coolant system. Subsequently the leak rate was confirmed to be steady at ~7.5 gallons/day. Lowering the core tank water level to the height of the anti-siphon valves halted the leak. After defueling the core tank and removal of several annular shield rings surrounding the reactor vessel, in February 2023, the leak was confirmed to be coming from a pipe fitting that penetrates the main core tank flange for pressure sensor MP-6A. The pipe fitting was replaced.

Other penetrations around the core were inspected, and one other similar fitting was replaced as a preventive measure. These pipe penetrations were found covered in iron rust from the annular shield rings. The pipes and all surrounding areas were cleaned.

All surfaces of the shield rings were reconditioned by grinding and sanding, and then covered with an anti-rust coating. The leak was confirmed repaired successfully.

Afterward, all the shield rings were returned to their locations surrounding the reactor vessel, along with reinstallation of all the shim blade drives, the regulating rod drive, and sensors for core tank level, flow, temperature, and pressure.

After completion of all corresponding instrumentation tests and calibrations, and creation of a special procedure for the moves, the core tank was re-fueled in September 2023.

Preparations for reactor restart continued for the rest of 2023, including temporary installation of a Pu-Be neutron source, and some adjustment of fission chamber positions in order to compensate for the photo-neutron source from the reactor's D2O reflector system diminishing over time while shut down.

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 (01/01/2024 - 12/31/2024) 0.28 mrem 0.99 mrem 0.74mrem 0.99 mrem Calendar Year Average 2024 0.8 mrem 2023 0.0 mrem 2022 0.7 mrem 2021 0.2 mrem 2020 0.2 mrem 2019 0.2 mrem 2018 0.2mrem 2017 0.4 mrem 2016 0.6mrem 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, 2024 - December 31, 2024 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 71 45 8

2 0

0 0

0 0

0 Total Person Rem= 2.83 Total Number of Personnel= 126 From January 1, 2024, through December 31, 2024, 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 8,237,825 liters discharged during CY2024 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 3.92E-5 Ci for CY2024.

The total tritium was 7.69E-2 Ci.

The total effluent water volume was 8,256,624 liters, resulting in an average tritium concentration of 2.36E-5 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. The only principal nuclide was Ar-41, which is reported in the following Table H-1. The 1727.6 Ci of Ar41 was released at an average concentration of 2.74E-10 µCi/ml. This represents 2.74% of EC (Effluent Concentration (lE-08 µCi/ml)).

3.

Solid Waste Two 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 2024 Ar-41 Average Discharged Concentration(})

(Curies)

(uCi/ml)

January 2024 0.00 0.00E+00 February 2.35 3.85E-12 March 23.62 4.84E-11 April 60.90 1.25E-10 May 196.06 3.22E-10 June 326.19 6.69E-10 July 22.68 4.65E-11 August 278.88 4.58E-10 September 303.40 6.22E-10 October 156.65 2.57E-10 November 183.06 3.75E-10 December 173.81 3.56E-10 Totals (12 Months)<2) 1727.60 2.74E-10 EC (Table II, Column I) 1 x 10-8

%EC 2.74%

(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 2024 Descriptions Volume 7.5 ft3 Weight 157 lbs.

Activity 0.235 mCi Date of shipment May 5, 2024 Waste processor NI A - Direct for Burial Waste broker Ecology Services Inc., Columbia, MD Disposition to licensee for burial Energy Solutions, Clive, UT Volume 109.8 ft3 Weight 2291 lbs.

Activity 160 mCi Date of shipment August 20, 2024 Waste processor Toxco Material Management Center, Oak Ridge, TN Waste broker Ecology Services Inc., Columbia, MD Disposition for burial Waste Control Specialists, Andrews, TX

Jan.2024 Feb.

Mar.

Apr.

May June July Aug.

Sept.

Oct.

Nov.

Dec.

12 months 27 TABLE H-3 LIQUID EFFLUENT DISCHARGES CALENDAR YEAR 2024 Total Total Volume Activity Tritium of Effluent Less Tritium Activity Water<1)

(xIQ-6 Ci)

(mCi)

(liters)

NDA<2) 3.06E-03 39,089 4.840 9.22E+00 60,687 NDA<2) 2.05E-03 180,957 NDA<2) 4.96E-04 134,942 4.360 9.52E+00 907,929 NDA<2) 9.86E-0l 1,440,207 5.000 l.96E+0l 221,361 5.540 3.49E+00 1,200,596 NDA<2) 6.18E-01 1,024,431 NDA<2) 3.35E-04 176,004 12.400 7.85E+00 1,773,794 7.040 2.56E+0l 1,096,628 39.180 7.69E+0l 8,256,624 Average Tritium Concentration (xlQ-6 µCi/ml) 0.078 151.942 0.011 0.004 10.486 0.685 88.546 2.907 0.604 0.002 4.426 23.326 23.585 (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 o-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.