ML20198P807

Annual Operating Rept for Jul 1996 - June 1997

ML20198P807
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	06/30/1997
From:	Bernard J, Lau E, Newton T NUCLEAR REACTOR LABORATORY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9711120016
Download: ML20198P807 (41)

Text

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NUCLEAR REACTOR LABORATORY AN INTERDEPARTMENTAL CENTER OF

,, .* MASSACHUSETTS INSTITUTE OF TECHNOLOGY JOHN A BrRNARD 138 Albany Street, Cambndge. VA 02139-4296 Activation Analysis Duector Telef ax No (617) 253 7300 Coolant Chemotry Dwector of Reador Operations Tet No. (617) 253 421I/4202 Nuclear Medicine Pnncipal Reso4rch rngineer Reactor Engineenng October 31,1997 U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 ATIN: Document Control Desk

Subject:

Annual Report, Docket No. 50-20, License R-37, Technical Specification 7.13.5 Gentlemen:

Forwarded herewith is the Annual Report for the MIT Research Reactor for the period July 1,1996 to June 30, 1997, in compliance with paragraph 7.13.5 of the Tcchnical Specifications for Facility Operating License R-37.

Sincerlly,

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Thomas H. Newton, Jr., PE C-Qk Edward S. Lau, NE Asst. Superintendent for Engineering Asst. Superintendent for Operations MIT Research Reactor MIT Research Reactor k.#

J hn A. Bernard, P i.D.

Director MIT Research Reactor JAB /gw

Enclosure:

As stated i

ec: USNRC - Senior Project Manager, Y

l NRR/ONDD USNRC - Region I- Project Scientist, Effluents Radiation Protection Section (ERPS)

FRSSB/DRSS 9711120016 970630 PDR R ADOCK 05000020 PDR

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MIT RESEARCH REACTOR l NUCLEAR REACTOR LABORATORY l MASSACHUSETTS INSTITUTE OF TECHNOLOGY l

ANNUAL REPORT to United States l Nuclear Regulatory Commission for

the Period July 1,1996 - June 30,1997 l

by l'

l REACTOR STAFF .

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.-n-Table of Contents -

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Table of conten ts f. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

i n trod uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 1 A; l Summary of Operuing Experience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 B.. Reactor Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ; . . . . . . . . . . . . . . . . . : 12 -

C.- S hutdown s and Scrams ? . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 .

D. MajorMaintenance ...... ... ...................................;....................15 .

E. Section 50.59 Changes. Tests, and Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 8

~ F; Environmen tal Surveys ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 G. Radiation Expos nes and Surveys Within the Facility ........................... 29 H .- Radioactive Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 g

"1. Sumnwy of Use of Medical Facility for Human Therapy ..... ................ 34 9

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MIT RESEARCH REACTOR ANNUAL REPORT TO U. S. NUCLEAR REGULATORY COMMISSION FOR THE PERIOD JULY 1.1996 - JUNE 30.1997 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 7,13.5, which requires an annual report following the 30th of June of each year.

The MIT Research Reactor (MITR), as originally constructed, consisted of a core of MTR-type fuel, ful' enriched in uranium-235 and cooled and moderated by heavy water in a four-foot diamerc ~* tmk, surrounded by a graphite reflector. After initial criticality on July 21,1953, t 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 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 en! ancement of the neutron flux, particularly the fast component, at in-core irradiatLn facilities. The core is hexagonal in shape,15 inches across, and utilizes fuel elements which are rhomboidal in cross section and which contain UALxintermetallic fuel in the form of plates clad in aluminum and fully enriched in uranium-235. Much of the original facility, e.g., graphite reflector, biological and thermal shields, secondary cooling systems, containment, etc., has been retained.

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-1 was shut down on May 24, 1974, having logged 250,445 megawatt hours during nearly 16 years of operation.

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 preoperational tests were conducted on all systems, the U.S.

Nuclear Regulatory Conunission 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. Routine 5-MW operation was achieved in December 1976.

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T This is the twenty-second annual report required by the Technical Specifications. -i and it covers the period July 1,1996 through June 30,1997. Previous reports, along with the "MITR-il Startup Repon" (Report No. MITNE-198, February 14,1977) have covered the startup testing period and the transition to routine reactor operation.-- This report covers -

- the twentieth full year of routine reador operation at the 5-MW. licensed power level. It was-another year in which the safety and reliability of reactor -operation met-.the- ,~

. requirements of reactor users.

A summary of operating experience and other activities and related statistical data

are provided in Sections A I of this report.

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A.

SUMMARY

OF OPERATING EXPERIENCE

'). General The MIT Research Reactor, MITR-II, has in many years been operated on a routine, five days per week schedule, modified as necessary to facilitate the preoperational testing and installation of several in-core experiments and to accommodate the medical program on boron-neutron capture therapy for cancer-treatment studies. When operating, the reactor is normally at a nominal 5 MW. However, as was the case for the last several years, substantial departures were made from this schedule during the period covered by this report (July 1,1996 - June 30,1997). Specifically, for much of this reporting period, the reactor was maintained at full power almost continuously (160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> / week) for four to five weeks. It was then shut down for a week for maintenance and other necuary outage activities. It was then started up to full power and was maintained there for another four to five weeks. This schedule was followed in order to support a major experimental program testing the corrosion cracking properties of various materials. The period covered by this report is the twentieth full year of normal operation for MITR-II.

The reactor averaged 104.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> per week at full power compared to 116.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> wr week for the previous year and 107.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per week two years ago. The apparent ower number of average full power hours per week for the FY97 is dr to the increased requirement for low power operation due to an eight-fold increase m human subject irradiations for the BNCT clinical trials and their preparation. It is also due to an increase in training activities for a doubling in the nomial number of reactor operator trainees.

The reactor was operated throughout the year with 24 elements in the core. The remaining three positions were used as follows: The first non-fueled position, B3, was used by the Irradiation-Assisted Stress Corrosion Cracking Facility (IASCC) since December 1995. The second and third core positions, Al and A3, were occupied by the D 3 and D-5 dummies. Position B-3 was occupied by the solid aluminum dummy D-4 whenever the IASCC was not installed. During FY97, compensation for reactivity lost due to burnup and the installation of the IASCC in-core experiment thimble was provided by three refuelings. These followed standard MITR practice which is to introduce fresh fuel to the inner portion of the core (the A- and B-Rings) where peaking is least and to place partially spent fuelin the outer portion of the core (the C-Ring). In addition, elements were inverted and rotated so as to achieve more uniform burnup gradients in those elements.

Two other refuelings were performed for the purpose of installation and removal of the IASCC experiment. One refueling was performed for the transfer of three spent fuel elements to the Spent Fuel Pool.

The MITR-II fuel management program remains quite successful. All of the original MITR-II element ( +45 grams U-235) have been permanently discharged. The average overall burnup for the discharged elements was 42%. (Maq: One element was removed prematurely because of excess outgassing.) The maximum overall burnup achieved was 48%. Ninety-five of the newer, higher loaded elements (506 grams U-235) have been introduced to the core. Of these, forty have attained the maximum allowed fission density. However, some of these may be reused if that limit is increased as would

4 seem warranted based on metallurgical studies by DOE. Another seven have been

, identified as showing excess outgassing and two are suspected of this. All nine have been

' removed from service. The other forty-six are either currently in the reactor core or have been partially depleted and are awaiting reuse in the C ring.

Piotective system surveillance tests are conducted whenever the reactor is scheduled to be shut down.

As in pr~vious years, the reactor was operated throughout the period without the fixed hafnium absorbers, which were designed to achieve a maximum peaking of the thermal neutron flux in the heavy water reflector beneath the core. These had been removed in November 1976 in order to gain the reactivity necessary to support more in-core experiment facili:ies.

2. Experiments The MITR-Il was used throughout the year for experiments and irradiations in support of research and training programs at MIT and elsewhere.

Experiments and irradiations of the following types were conducted:

a) Prompt ganuna activation analysis for the determination of boron-10 concentration in blood and tissue. This is being performed using one of the reactor's beam tubes.

The analysis is to support our neutron capture therapy program.

b) Experimental studies of the role of metallic and organo-metallic groups in the final propenies of polymers.

c) Use of neutron activation analysis to determine the concentrations of heavy metals in sludge from sewage treatment plants, d) Irradiation of archaeological, environmental, engineering materials, biological, geological, oceanographic, and medical specimens for neutron activation analysis purposes, c) Production of dysprosium-165, holmium-166, copper-64 and gold-198 for medical research, diagnostic, and therapeutic purposes.

f) Irradiation of tissue specimens on particle track detectors for plutonium radiobiology.

g) Irradiation of semi-conductors to determine resistance to high doses of fast neutrons.

h) Use of the facility for reactor operator training.

i) Irradiation of geological materials to determMe quantities and distribution of fissile materials using solid state nuclear track detect <

j) Evaluation of various chemical additives for the suppression of nitrogen-16 activity in a boiling water reactor environment.

k) Use of trace analysis techniques to identify and monitor sources of acid deposition (rdn).

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1) . Operation of an in-core slow strain rate testing rig to evaluate irradiation-assisted

, stress corrosion cracking of metals.

m) Measurements of the energy spectrum of leakage neutrons using a mechanical chopper in a radial beam port (4DH 1). Measurements of the neution wavelength by Bragg reflection then permits demonstration of the DeBroglie relationship for physics courses at MIT and other universities.

n) Gemmairradiation of seeds for demonstration of radiation damage effects for high school students.

o) Use of beam tubes for testing of prototype neutron detectors.

p) Neutron activation analysis to determine iron oxide contamination in aluminum specimens.

q) Neutron activation analysis of serum samples in an effort to correlate mineral deficiencies with certain diseases, r) Determination of uranium concentrations in samples of mica.

s) Gamma irradiation of PLGA-coated alumina fibers for medical research and applications, t) Neutron activation of uranium samples for detector calibration at the McClellan AFB, California.

Dose reduction studies for the light water reactor industry began reactor use on a reca) tr basis in 1989. (Planning and out-of-core evaluations had been in progress for sevaal years.) These studies entail installing loops in the reactor core to investigate the chemistry of corrosion and the aansport of radioective crud. Loops that replicate both pressurized and boiling water reactors have been built. The PWR loop has been operational since August 1989. The BWR loop became operational in October 1990. A tlurd loop, one for the study of irradiation-assisted stress corrosion cracking (IASCC),

became operational in June 1994 and a fourth one, also for the study of crack growth, in April 1995. The IASCC loop has been in operation in-core since April 1996.

Another major research project that is now making and will continue to make extensive use of the reactor is a program to design a facility for the treatment of glioblastomas (brain tumors) and melanomas (skin cancer) using neutron capture therapy.

This is a collaborative effort with the Beth Israel-Deaconess Medical Center which is affiliated with the Harvard Medical School.

3. Changes to Facility Design Exce)t for minor changes reported in Section E, no changes in the facility design were made curing the year. As indicated in past reports the uranium loading of MITR-II

- fuel was increased from 29.7 grams of U-235 per plate and 445 grams per element (as made by Gulf United Nuclear Fuels, Inc., New Haven, Connecticut) to a nominal 34 and 510 grams respectively (made by the Atomics International Division of Rockwell International, Canoga Park, California). With the exception of seven elements (one Gulf, six AI) that were found to be outgassing excessively, performance 1.as been good.

(Please see Reportable Occurrence Reports Nos. 50-20S9-4, 50-20/83-2, 50-20/85-2, 50-20/86-1, 50-20/86 2, 50-20/88-1, and 50-20/91-1.) The heavier loading results in

6-41.2 w/o U in the core, based on 7% voids, and corresponds to the maximum loading in Advanced Test Reactor (ATR) fuel. - Atomics Intemational completed the production of

' forty-one of the more highly loaded elements in 1982, forty of which have been used to some degree. Thirty-one with about 40% burnup have been discharged because they have attained the Hssion density limit. Only one other Al element remains in use. Of the other eight, six were, as previously reported to the U.S. Nuclear Regulatory Commission, removed from service because of excess outgassing and two were removed because of suspected excess outgassing. Additional elements are now being fabricated by Babcock &

Wilcox, Naval Nuclear Fuel Division, Lynchburg, Virginia. Fifty-five of these have been received at MIT, forty-one of which are in use. One has been removed because of suspected excess outgassing and fourteen have been discharged because they have attained the fission density limit.

The MITR staff has been following with interest the work of the Reduced Enrichment for Research and Test Reactors (RERTR) Program at Argonne Natiomd Laboratory, particularly the development of advanced fuels that will permit uranium-loadings up to several times the recent upper limit of 1.6 grams total uranium / cubic centimeter. Consideration of the thermal-hydraulics and reactor phyr.cs of the MITR-Il core design show that conversion of MrlR-Il fuel to lower enrichnent must await the successful demonstration of the proposed advanced fuels.

4. Changes in Performance Characteristics Performance characteristics of the MITR-il were reported in the "MITR-Il Startup Report." Minor changes have been described in previous reports. There were no changes during the past year.

5. Changes in Onerating Procedures With respect to operating procedures subject only to MITR internal review and approval, a summary is given below of those changes implemented during the past year.

Those changes related to safety and subject to additional review and approval are discussed in Section E of this report.

a) PM 3.6, " Waste Storage Tank Dump Procedure," was revised to clarify the sequence of steps and to make the procedure user friendly. No substantive change was made. (SR#0-96-11) b) PM 7.4.3.2, 7.4.3.3, 7.4.3.4, " Chemical Cleaning of Ileat Exchangers," were rewritten to improve both the efficiency of the procedure and to add a caution thit the cleaning chemical be added to the flush cart at a speciDed control rate.

(S R#0-96-12) c) PM 7.3.2, "D 02 lon Column Dedeuterization/Deuterization," was revised to improve clarity. No substantive change was made. (SR#0-96-13) d) The Reactor Systems Manual was rewritten to reflect updates in the reactor flow systems, the non-nuclear instrumentation section, and system alarms and interlocks. (Nntc; All of the changes made in this update had themselves been subjects of previous safety reviews. The safety review reported here was solely concerned with the rewrite of the descriptive materials.) (SNO-96-17) e) PM 6.1.2.5, " Charcoal Filter Efficiency Test," was revised and rewritten to improve the clarity and conciseness of the previous procedure. A single procedure

7-now covers the testing of both charcoal filter banks. The new procedure also

, includes a new schematic diagram deaicting the setup and layout for the test. The change does not affect the radiologica aspect of the test. (SR#0-96-18) f) AOP 5.7.18, " Ammonia Detector," and PM 6.5.21, " Ammonia Iktection System Calibration," were revised to increase the external building ammonia detector system setpoints to reduce the frequency of false alarms which were caused primarily by atmospheric conditions. (SR#0-96-19) g) A computer program for xenon worth calculation was written for control room operator assistance in the detennination of xenon reactivity worth following shutdown from full power operations. (SR#0-96-20) h) Control Room reference information for temperature and xenon worth calculations was written and documented as an operator aid in the detennination of estimated critical positions. (SR#-0-96-21) i) PM 6.6.2.1.1 " Fire Extinguisher Monthly inspection," and PM 6.6.2.1.2, " Fire Extinguisher Semi-Annual inspection," were revised to include new types, weights and requirements for inspection of reactor fire extinguishers. There is no reduction or decreased efficiency of the fire protection equipment. Also, the revision has no impact on ALARA because the changes do not significantly affect the amount of time spent in radiation areas. (SR#0-96-22) j) PM 3.8.4, "Startup of Make-Up H2O Storage Tank System in Standby (Recirculation) Mode," was revised to increase the initial system recirculation flow rate. Safety is improved by this change by maintaining an above minimum flow rate in the Make-Up Water Recirculation System. (SR#0-97-1) k) PM 3.5, " Daily Surveillance Check," was revised to update a daily surveilhince procedure to reflect current equipment and revised operation setpoints of mstruments more accurately. Safety is improved by increasing the clarity and accuracy of the procedure for use with current equipment. The safety review has no impact on ALARA because the changes do not involve radiological work.

(SR#0-97-3)

1) PM 6.1.3.10. " Emergency Battery Discharge Test," was revised to update the surveillance and maintenance procedure and requirements on the reactor's new emergency battery bank. Safety is improved by better consistency in tracking the status of the reactor's emergency power system. (SR#0-97-4) m) PM 6.1.1, " Emergency Cooling System," was revised to update the test and calibration procedure. Cautions were added to emphasize the correct order of valve realignments at key points in the procedure. Other steps were clarified per current practices, including handling emergency core cooling spray nozzles separately, and directing discharge of primary coolant during the test back into the core tank. The change improves reactor safety by reducing the possibility of incorrect valving causing primary coolant to be inadvertently siphoned out of the primary core tank.

The changes update the surveilkmcc procedure but do not change its function. The update also has a favorable impact on ALARA because the changes are intended to prevent unnecessary handling of primary coolant. (SR#0-97-5) n) PM 3.1.1.1, "Ts o Loop Mechamcal Startup Checklist." PM 3.1.1.2, "Two Loop Instrumentation Startup Checklist." PM 3.1.1.3, "Two Loop Cooling Tower

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Oxration," L PM C3,1.3, "Startup i for ' less iThan 1001- kW : Operation," -

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-less "Ihan 100 kW ration, and PM 3.2.3, " Maintenance Checklist"'were all ,

revised andLmodir ifor use with current equipment, for clarity and better-1 agreement with each other, and to incorporate all previously temporary changes.L Reactor safety-is improved by increasing the accuracy and consistency of the o

' procedures. The modification has no ' impact on ALARA because the changes do not involve the radiological aspects of the procedures. (SR#0-97-6)

L o)_ _ ' PM 6.1.3.6, " Reactor Building Overpressure Scram," PM 6.1.3.8, "D2O Reflector IxvelIndication - Scram, Alarm, and Calibration," PM 6.1.4.1, " Nuclear Safety System Response Time," PM 6.1.4.4, " Primary Coolant Flow Scram Time," PM

- 6.3.9, "Iow level Medical Shutter Tank / Pump .Off," PM 6.3.13 " Withdraw PM 6.4.8,<  !

Permit Interlock," PM 6.4.3, "High Level Equipment Room Sump," Tank Level ,

"High Pressure Reactor Inlet" and _ PM 6.5.13, " Shield - Storage

' Calibration,"_ were all revised to er. hance accuracy and clarity, arid to update the -

surveillance procedures. (SR#0-97-7);,

p) - PM 3.7,2,;" Daily Security Checklist," was revised and updated for clarity. and uniformity and to acconunodate the use of new equipment and installation of new lab / office space. - (SR#0-97-8) _

q) PM 6.1.3.2, " Period Channel Calibration," was revised and modified for clarity and to standardize wording. (SR#0-97-9) -

r) ' PM 3.8.1B, "Startup of Inlet DI System in Standby (Recirculation) Mode," was revised and updated to enhance clarity of a startup procedure for the makeup water system, p;'#0-97-10)

, s) PM 7.1,1.1, " Shim Blade Drive- Mechanism Disassembly and Reassembly Procedures," was revised and modified to update the procedures to reflect current practices in the assembly _ of the reactor shim blade drive mechanism. Safety is improved in that the updated procedure will ensure that the mechanism is properly

- assembled according to current practices. In addition, verification of proper assembly by a second qualified individual is now required by the new procedure.

(SR#M 961) -_

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6. Surveillance Tests and Inspec' ions 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 inspecuon 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. The tests and inspections are scheduled throughout the year with a frec uency at least equal to that required by the Technical Specifications. Twenty-seven suc i tests and calibrations are conducted on an annual, semi-annual, or quarterly basis.

Other surveillance tests are done each time before stanup of the reactor if shut down for more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, before startup if a channel has been repaired or de-energized, and at least monthly; a few are on different schedules. Procedures for such surveillance are incorporated into daily or weekly startup, shutdown or other checklists.

During the reporting yriod, the surveillance frequency has been at least equal to that required by the Technica Specifications, and the results of tests and inspections were satisfactory throughout the year for Facility Operating License No. R-37.

7. Status of Snent Fuel Shipment Pursuant to Amenwent No. 25 to Facility Operating License No. R-37, paragraph 2.B.(2) subparagraph (b), reported herewith is the status of the establishment of a shipping capability for spent fuel and other activities relevant to the temporary increase in the possession limit.

MIT began efforts for spent fuel shipment as early as 1983. At that time, the plan was to use two MH-l A casks that had been acquired by DOE and which were beirg prepared for use by the non-power reactor community. After an MH-1 A cask became unavailable, MIT made arrangements with General Electric to use the GE-700 cask for shipment of the MITR spent fuel. When the GE-700 cask was removed from service voluntarily by GE, the BMI l cask became the only one available that is approved for transportation of irradiated fuel elements.

The capability to ship spent MITR fuel was established by the end of 1992.

Specifically, the following was accomplished:

(a) The Certificate of Compliance and the Safety Analysis Report of the BMI-l cask were reviewed by MIT and the cask was determined to be acceptable for shipping MITR spent fuel. Arrangements have been made with DOE for MIT to use this cask.

(b) The University of Missouri Research Reactor (MURR) basket was reviewed and found to be suitable for use with the MIT fuel elements in the BMI-l cask. MURR has agreed to make their basket available to MIT for the required shipments.

(c) A quality assurance program for MITR-II spent fuel shipment was prepared and approved under the MITR safety review program. This Q'A program was approved by NRC on July 23,1991.

(d) The decay heat load of each spent element was determined by a member of the MITR staff and found to be within the limits specified in the Certificate of Compliance for the cask. Radiation shielding calculations were also performed and radiation levels associated with the loaded cask were estimated to be within allowed

limits. Criticality calculations were perf,rmed using the Monte-Cmlo Code KENO-

. V which was obtained from the Radiatun Shielding Information Center of the Oak Ridge National Laboratory. Results show the ihe degree of suberiticality of a cask fully loaded with MIT fuel elements is within specification.

(c) In order to cross-check the cros; sections used in the KENO-V code, criticality analyses were performed using a second Monte-Carlo code, MCNP. Results obtained were consistent with those obtained using KENO-V.

(f) Arrangements have been made with the fuel receiving organization at the Savannah River facility. Specific data on the MITR-Il spent fuel elements were compiled.

The Appendi .s occument and criticality study were prepared and reviewed by the spent fuel processing center.

(g) Spent fuel elements in the MITR s pent fuel storage pool were arranged and grouped in accordance with our procedure I or shipment , reparation. A special structure for support of the BMI l cask was designed and fa aricated.

(h) A third fuel storage rack, which has a capacity of twenty-five fuel elements, was built and installed in the spent fuel storage pool.

(i) License Amendment No. 25 which provided a temporary increase in the possession limit was extended to 31 December 1993.

(j) A criticality study of the BMI-l cask with f~sh MITR fuel was completed and approved by the U.S. Department of Energy.

(k) Funding was allocated by the U.S. Department of Energy for the return to a DOE facility of spent MITR fuel.

(1) Proodures for spent fuel shipment were prepared.

(m) A proposed route was reviewed and approved by NRC. All necessary State and City permits were obtained.

Six shipments of eight elements each were completed during the early part of 1993.

In each case, the spent fuel was returned to the U.S. Department of Energy's facility at Savannah River, SC. As a result of these shipments, the on-site inventory of U-235 was sufficiently reduced so that there was no longer any need for the temporary increase in the possession limit that had been obtained under License Amendment No. 25. Accordingly, that amendment expired on 31 December 1993, in early 1994, one shipment of eight elements was completed. No shipments were made in FY1997. At present, several additional shipments are needed in order to reduce the inventory of spent fuel at MIT to zero. However, it is currently unclear as to when or even if these shipments will occur. The problem is that the U.S. Department of Energy (DOE) has stopped the reprocessing of spent fuel and it has only limited storage space available. DOE is currently evaluating various options that would allow continued returns

of spent fuel and MIT will notify NRC of the DOE decision as soon as it is known. DOE

! has mdicated that a shipment may be possible in 1998.

As a result of the continued inability of the U.S. Department of Energy (DOE) to accept the return of spent MITR fuel, the inventory of that fuel grew to the point where a

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tempormy increase in the possession limit was_ again necessary. That request was granted :

y,as License Amendment No; 29. It was issued effective 8 April 1996.

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B. REACTOR OPERATION Information on energy generated and on reactor operating hours is tabulated below:

Quarter 1 2 3 4 Total

1. Energy Generated (MWD):

a) f,llTR Il(MIT FY97) 195.3 256.9 278.8 336.5 1067.5 (normally at 4.9 MW) b) MITR-Il 15,446.8 (MIT FY76-96) c) MITR-I 10,435.2 (MIT FY59-74) d) Cumulative, 36,949.5

~

MITR-1 & MITR-Il

2. MITR-Il Operation (Hrs):

(MIT FY97) a) At Power

(>0.5-MW) for 1025.5 1288.8 1414.5 1704.9 5433.8 Research b) Low Power

(<0.5-MW) for 51.2 11.4 24.4 49.1 136.0 Training (O and Test c) TotalCritical 1076.7 1754.0 5569.8 1300.2 l 1438.9 W These hours do not include reactor operator and other training conducted while the reactor is at full power for research purposes (spectrometer, etc.) or for isotope production. Such hours are included in the previous line.

4

~

C. - - SHUTDOWNS AND SCRAMS During the period of this report there were 8 inadvertent scrams and 10 unscheduled power reductions.-

The term " scram" refers to shutting down of the reactor through protective system

- action when the reactor is at power or at least critical, while the term " reduction" or

- " shutdown" refers to an unscheduled power reduction to low power or to subcritical by the reactor operator in response to an abnormal condition indication. Rod drops and electric power loss without protective system action are included in unscheduled power reductions.

The following summary of scrams and shutdowns is provided in approximately the f same format as for previous years in order to facilitate a comparison.'

l. Nuclear Safety System Scrams Ingal Channel #4 trip as result of amplifier malfunction. 2 a)~

~

Channel #1 trip as result of electronic noise. I b)

J Channel #4 trip as result oio"erly conservative setting. I c) d) Low voltage chamber power supply trip as result of 1

suspected Channel #3 power supply failure.

Subtotal -5

- !!. Process System Scrams a) - _Imw flow primary coolant trip as result of pump MM-1 A failure. I b) Low flow primary coolant trip as result of operator l

~

error while inspecting flaw recorder.

c) . Low pressure MP-6 trip as result of operator enor i while putting cooling towers to spray. -I Subtotal 3 T

, ~ _ . .. - ~ , , . .- . , . _ - - . _ _ , . _ . . - - . _ - . . _ . _ _ _ _ , - . , , . _ - = . - _ - - ., - + ,

.111. , IJnscheduled Shutdowns or Power Reductions NTD Silicon machine malfunctions. 4 a)

Shutdown due to loss of offsite electricity, 4 b)

Shutdown to repair failing Ch. 6. I c) 1 d) Shutdown to inspect IASCC.

St btotal 10 To011 18 IV. Experience during recent years has been as follows for scrams and uascheduled shutdowns:

Fiscal Year Numher Scrams Shutdowns Leial 93 6 14 20 94 13 32 45 95 17 28 45 96 18 21 39 97 8 10 18

l D. MAJOR MAINTENANCE

' Major maintenance projects performed during FY 97, including the effect, if any, on the safe and reliable operation of the MIT Research Reactor are described in this Section.

Much maintenance was performed to imorove the safe, reliable and efficient operation of the MIT Research Rc* tor and to support the ongoing research programs involving clinical trials of Boron. Neutron Capture Therapy and the identification of improved water chemistries that will result in reduced radiation exposure to workers in the nuclear industry. The latter now involves four in-core experrnents. These are the Pressurized Coolant Chemistry loop (PCCL), the Boiling Coolant Chemistry Loop (BCCL), the Irradiation Assisted Stress Corrosion Cracking i xi;ity (IASCC), and the SENSOR facility.

One of the FY 97 maintenance items was the preparation and installation of an experiment facility in the reactor core for the IASCC project. This project, which requires operation of an INSTRON machine on top of the reactor core tank lid, was imtially installed in the reactor core between July and October 1995.

The repair and maintenance of machinery and computer control and monitoring software and hardware for neutron transmutation doping of silicon (NTD Si) also required support. This machinery, installed in two of the reactor throughports, includes two twenty-foot tubes for each port, rotating and pushing mechanisms, billet handling and storage conveyors, electronics, and associated microprocessor-based controllers and computer tracking systems. Sig sal validation and fault-to 'l at routines were developed in-house and i nplemented with the existing computer control itware.

Other major maintenance items performed in FY 97 were as follows:

(1) The Reactor Containment Shell outer surface was mostly deleaded and completely coated with epoxy primer and two coats of polyurethane paint on the outside for added protection against weathering. The outer Truck Lock door and its lifting mechanism support frame and the outer side wall of the Utility Room were also stripped and paintcd with the same coating.

(2) The reactor's area radiation monite-ing system was completely replaced with new Ludlum units. This replacement includes the detector units, local and remote readout units, alarms and interlock functions of the system.

(3) New area radiation monitor units (Ludlum Digital Wall-Mount) were added to the equipment room general area and to the Reactor Floor area underneath the IASCC platform. These units provide remote readouts and independent alarms in the Control Room.

(4) Three of the effluent radiation monitoring units were replaced with new units. This includes new detectors, remote readout unit in the Control Room as well as alarms and interlock functions of the system. Three others are now being tested. The entire effluent radiation monitor m, tem will eventually be replaced with the new units.

(5) The helium gas supply bank for the NTD Si equipment was relocated to the NW12 Receiving Room. This reduces the total number of compressed gas cylinders needed inside the reactor Containment Building. It also minimizes

handling and transportation of the pressurized cylinders, and thus improves

, ,- personnel safety.

(6) All out-of-core helium gas supply polyethylene tubes for the shim blade drive mechanisms at the Reactor Top area were replaced with copper tubing.

This minimizes helium gas leakage due to aging of the polyethylane supply lines and it also simplifies the gas supply line layout in the Reactor Top area.

(7) Control Room console Channel #8 analog readout meter was replaced with a new unit.

-(8) Stack base effluent monitor system blowers were replaced with new blowers.

(9) Nuclear Safety Channels #1 through #6 potentiometers were replaced with new improved units. Many electronic components for the channels and amplifiers were replaced and renewed throughout the fiscal year as preventive maintenance.

(10) The Cooling Tower sprinkler system air compressor switching and control mechanism was replaced.

(11) Cooling Tower #1 fan motor was replaced and rewired.

(12) Heavy water reDector system auxiliary pump DM-2 was rebuilt, and the system splash guard and vicinity floor were resealed.~

(13) Many rew ;ightc and fixtures were installed around the reactor restricted area and reactor parking lot to improve lighting and security.

(14) The Containment polar crane pendant cable and crane control box were replaced. The crane system was also inspected, load test performed and its certification updated.

(15) The Reactor Irak alarm system leak tapes were extensively replaced and tested throughout the reactor systems.

(16) The main compressor air reservoir tank (120 gallon size) in the Utility Room was replaced.

(17) A special underwater vacuuming system equipped with high efficiency filters, long hose and special brush tip was designed and fabricated to remove minor deposits on the bottom and side structure of the main core tank.

'(18) The secondary system booster pumps at the Cooling Tower had pump casing drain lines and copper riser / standpipe lines replaced.

(19) The pneumatic blower for the IPHlHNW13 rabbit tube system was replaced, and its base mounting platform and weather housing on top of building NW13 was rebuilt.

(20) The entire roof of the Utility Room was replaced. The roof of the Waste

. ,- Tank Shed was also re-scaled.

(21) The heavy water reflector D2 0 Dump Tank level gauge in the Control Room were rebuilt, tested and calibrated.

(22) Self-contained emergency lights were added to new locations inside the Containment Building - the Control Room, the Equipment Room and the Primary Chemistry setup area.

(23) The SENSOR experiment thimble and specimen were prepared and shipped offsite to GE using the GE-2000 shipping cask.

(24) An additional ion column and water purification / recirculation system was installed underwater in the Spent Fuel Pool to improve cleanup efficiency.

(25) The original 4DH4 spectrometer facility was dismantled and preparatory effort started for installation of a new reflectometry facility.

(26) A new secondary ecolant water chemistry on-line monitoring system was installed at the base of the Cooling Tower.

(27) The secondary coolant system water conductivity / blowdown control monitor was replaced in the Utility Room.

(28) The city water inlet valve HV-16 in the equipment room was replaced for better makeup flow control.

Many other routine maintenance and preventive maintenance items were performed throughout the fiscal year.

E. SECTION 50.59 Cil ANGES. TESTS. AND EXPERIMENTS This section contains a description of each change to the facility or 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.

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 S AR have been or are being revised to reflect these changes, and they either have or will be forwarded to the Document Control Desk, USNRC.

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 approvalis documented by means of the Safety Review Form.

All other experiments have been done in accordance with the descriptions provided in Section 10 of the S AR, " Experimental Facilities."

-i Secunty Plan

, SRMbb10 .-(12/1066) .

. De Security Plan was updated to reflect administrative and other routine changes.

No substantive ~ change was made.

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20-Fmereency Plan' ~

-.- SR#0 96-141 (04S007)l- #0-96-15 _ -(04G0S7),: - #0-96-16 :(0400S7)- y

- 'Ihe MITR Emergency Plan was rewritten in its entirety and submitted to the_U.S.

- Nuclear Regulatory Commission on June 6,1997f The changes did not decrease the plan's

' effectiveness. Implementing procedures were made more " user-friendly", a more graded aporoach to emergency res ponse was adopted, and the " General Emergency" category was de eted because the MITR lacks the radionuclide inventory to create such an emergency.

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Pressurized Cwlant Chemistrv iron (PCCL)

~

. SR#0 86 9 (04/21/88), #0-88-4 (07/28/88), #0-88 5 (09A)9/88), #0-8814 (12A17/88),

' #0-89 2 (01 A)6/89). #0-89-3 (01/19/89), #0-89-6 (0I/24/89), #0-89 9 (06A)2/89), #0 8914 (06/19/89), #0-90-6 (03/20 S 0), #0-90 7 (03/20 S 0), #0-90-8 (03/20S 0), #0 90-9 (03/20S 0), #0-90 25 (12/10S 0), #0-90 26 (12/1890), #0-90 27 (12/18 S 0), #0-91-8 (05/21S1), #0-91 21 (12/27S I), #0-92 2 (01/27S 2), #0-92 12 (08/19 S 2), #0-94 6 (06/30/94), #0 94 3 (03/10S4).

This project involves the design, installation, and operation of a pressurized light-water loop in the MITR core for the purpose of studying the production, activation, and transport of corrosion to determine the optimum method for reducing the creation of activated corrosion products (crud) and thereby reducing radiation fields associated with

,ressurized water reactors (PWRs). The ultimate goal is to reduce radiation exposures to PWR maintenance personnel.

Approval for the PCCL was given by the MITR Staff and the MIT Reactor Safeguards Committee on 04/20/88. It was detennined at that time that no unreviewed safety question existed because no failure or accident associated with the PCCL could lead to an accident or failure involving reactor components. Details of that detennination, togethm with safety review #0-86-9, were submitted to the U.S. Nuclear Regulatory Commission on 04/21/88.

Subsec uent to the determination that no unreviewed safety question existed, specific procewres for PCCL operation were prepared. These included:

- Procedure for Ex-Core Testing Supplement to the Safety Evaluation Report PreoperationalTest Procedure Abnormal Operating Procedures for the PCCL

- Procedures for PCCL Startup/ Shutdown Procedures for PCCL Installation / Removal

- Procedures for Transfer of Used PCCL Components to a Separate Storage Tank in the Spent Fuel Storage Pool.

Experiments using the PCCL began in April 1989 and have been quite successful.

Ilowever, no design changes were made to the PCCL during the period covered by this report and the PCCL was not operatiomil during this reporting period.

l lloiling Coolant Chemistrv loop (BCCL)

, SR#068914 (06/19/89), #0 89-20 (12/20/89), #0 90-17 (09/17/90), #0-9018 (09/14/90),

1. 0-90 20 (10/15/90), #0-91 20 (01/30/92), #0-92-11 (08/15/92), #0-92 16 (09/25/92),

1. 0-93 IP (09/03/93), #0-94 5 (01/24/94), #0-94-6 (06/30/94).

This project involves the design, installation, and operation of a boiling light water loop in the hilTR core for the purpose of studying the production, activation, and transport of corrosion products. The effect of various water chemistries is being examined to determine the optimum method for reducing the creation of activated corrosion products (ciad) and thereby reducing radiation Gelds associated with boiling water reactors (BWRs).

The ul;imate goal is to reduce radiation exposures to BWR maintenance personnel.

In 1988 and 1989, the Reactor Staff made a determination that boiling within an in-core facility is not contrary to the technical specifications provided that reactivity limits for movable exp timents are not exceeded, it was also concluded that boiling in the proposed experiment volume v auld not significantly affect reactor operation. Accordingly, a carefully controlled experiment was proposed to demonstrate that boiling within an in-core facility would not adversely affect reactor cperation. Following both a determination that no unreGewed safety question was involved and approval by the hilt Reactor Safeguards Committee, this expenment was conducted. The results were as expected.

The final safety evaluation report for the BCCL was completed on 8 h1 arch 1989 and approveda' y the hilTR Staff. On 12/20/89, the hilt Reactor Safeguards Committee detennined that there was no unreviewed safety question involved in the conduct of the llCCL experiment and approved the BCCL SER. On 9 hiarch 1990, a copy of the BCCL SER together with the safety analysis prepared by the hilTR Staff were forwarded to the U.S. Nuclear Regulatory Commission pursuant to 10 CFR 50.59(b)'1).

Subsec uent to the determination that no unreviewed safety question existed, specific procec ures for BCCL operation wem prepared. These included:

Preoperatiomd Test Procedure.

- Abnormal Operating Procedures for the BCCL,

- Procedure for BCCL Startup.

Other necessary procedures such as BCCL shutdown and installation / removal are the same as those previously developed and approved for the PCCL. Experiments using the llCCL began in October 1990 and have been successful in that many theories concerning the traaspoit of nitrogen-16 in boiling water reactors have been disproved.

Experiments that make use of the BCCL facility were conducted during portions of this reportmg period. No changes were made to the BCCL experimental protocol during this reporting period.

Exocriments Rchted to Neutron Capture Therapy

' SRd 89-4 (01/23/89), #0-89 8 (03/01/89), #0-917 (05/06/91), #0 91 17 (03/06/92),

1. 0 92 3 (03/06/92), #0-92 4 (03/02/92), #M 92 2 (05/14/92), #0 93 5 (05/28/93),

1. 0-93 9 (07/l3/93), #0 93 20 (l1/30/93), f0 9419 (12/02/94), #0 96 5 (05/03/96),

1. 0-97 2 (02/18/97).

In conjunction with the Tufts - New England Medical Center (NEMC) and with the support of the U.S. Department of Ener :y, MIT has designed an epithermal neutron beam for the treatment of brain cancer (g loblastoma). Thermal beams have been used successfully for this treatment in Japan. The reason for designing an epithermal beam is to allow tumor treatment without having to subject the patient to surgery involving removal of a portion of the skull. Also, an epitherrnal beam gives greater penetration. In October 1991, MIT hosted an international workshop for the purpose of reviewing proposed beam designs and dosimetry. Subsequent to the receipt of advice from ti,e workshop panel members, a final design was selected for the epithermal filter for the MIT Research Reactor's Medical Therapy Facility beam. That ucsign, which was one of many that had been previously constructed and evaluated, is No. M-62, it has now been installed permanently. Approvals of the protocol for the conduct of patient trials have now been received from all requisite MIT and NEMC Com nittees as well as from the U.S. Food and Drug Administration. Also, a license amendment and quality management plan for use of the MIT Research Reactor's Medical Therapy Facility was issued by the U.S. Nuclear Regulatory Commission as License Amendment No. 27 on February 16, 1993.

Subsequent to the accei at of that license amendment and a similar one in August 1993 for our medical partner, t ic Tufts - New England Med! cal Center, both procedures for performing IlNC1 and a preoperational test package were prepared. The latter was completed durmg FY 94.

Patient trials were initiated in September 1994 as part of a Phase 1 effort that is required by the FDA. In December 1994, changes were issued to ceHain of the procedures that had been prepared for conduct of the irradiations. These changes were intended to reduce the signat.re burden on senior personnel during the trials so that their full attention could be given to de human subject.

Three subjects were irradiated in FY95. One more was done in FY96 in conjunction with NEMC. A change of medical partners then occurred. The leogr'un was a joint effort between MIT and the New England Deaconess llospital (NEDil; which was affiliated with the liarvard Medical School. This change necessitated an amendmeni to the NEDil's license for radioactive materials and their use, as well as to the various internal approvals. Subsequent to receipt of these licenses / approvals, the Phase I trial for melanoma was contmued. Also, a se 3arate Phase i protocol for glioblastoma multiforme was approved. Patient trials under t mt protocol were initiated in July 1996. In FY97, New England Deaconess llospital merged with the Ileth Israel llospital. The resulting organir.ation is lleth Israel-Deaconess Medical Center which is now also a major teaching hospital for the liarvard Medical School. Under the new partnership, a total of eigt,it human subjects were irradiated at various dose levels. A summary of these irradiations is presented in Section 1.

Technical Specification #6.5, " Generation of Medical Therapy Facility licam for lluman %crapy," and its associated ilNCT Quality Management Program were updated.

The change is purely administrative in nature. No substantive changes of any type

i

. 24  !

. i resulted. - 1hc langung pdate in the two documents was to reflect transition from NRC

, regulation to State regu a ion of nulical use licensees, and thereby to prevent any possible subsequent disruption of the ongoing BNCT research program due to such admmistrative shift. The change allows MIT to conduct BNCT on human subjects from both NRC and Agreement State (the Conunonwealth of Massachusetts) nwdical use licensees whose sanses contain BNCT-specific conditions and conunitnwnts for BNCT clinical trials on human subjects conducted at the MIT reactor. The change was approved by the MIT 3 Reactor Safeguards Committee on Febiuary 18,1997 and by the NRC on April 3,1997.

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Irradiation Assisted Stress Corrosion Crackine (IASCC) Exneriment

, SR#04915 (06/19/89), #0-90 2I (10/22/90j, #0-90 22 (lU/22/90), #0-90-23 (l1/05/90),

• #0-910: (12/27/91), #0 92 5 (04/02/92), #0 92 17 (09/28/92),#0-92 21 (01/21/94)

1. 0-93 6 (05/26/93), #M 931 (05/24/93), #0-9314 (09/09/93), #0 94-4 (01/28/94),

1. 0 94-7 (06/11/94).

In the past several years a variety of austenitic stainless steel components in boiling water reactor (BWR) cores have failed by at- intergranular cracking mechanism called irradiation assisted stress corrosion cracking (in FCC). Characteristics of such failures are that the component was exposed to a fast neutron fluence under tensile stress and in an oxidizing water environment.

i A facility to study IASCC in typical BWR water and radiation environments has been designed, built, and put into in core service. This facility positions a pre irradiated test specimen in the core of the MIT Research Reactor, circulates water with controlled temperature and chemistry past the specimen, and applies a tensile load to the specimen to maintain a constant slow strain rate until specimen failure. A DC potential drop (DCPD) technique was developed to measure specimen strain during in core testing. Electrodes are incorporated to measure the s test, and for initial analysis,pecimen's the sensitivity of the electrochemical corrosion specimen's ECP to varying water potential (ECP chemistry, flowrate, in core position, and reactor power level. A chemistry control system was designed and built to measure and control the water chemistry. Remote specimen handling tools and procedures were developed to allow the fracture surface to be analyzed by scanning electron microscopy (SEM). The facility and operating procedures were desi;;ned to minimize radiation exposure of personnel during facility operation and transfer to a bot cell for specimen removal and replacement.

Initial in ccre tests, which measured the ECP of stainless steel in in-flux sections of the testing rig have been completed successfully. These tests showed that the desired oxidizing environment can be established and monitored during in core SSRT testing.

Initial in-core SSRT testing is presently underway. Results of these tests will be used to investigate the effects of neutron fluence and materials variables on IASCC.

As part of the preparations for this experiment, a new reactor top lid was designed and ins:alled in FY 93. This lid, which provides an additional four inches of vertical clearance for in-core experiments, meets or exceeds the specifications for the original lid.

Radiation levels directly above the reactor were reduced as a result of the installation of this new lid.

The IASCC experiment was operating in-core during portions of this reporting

26- l ,

k SENSOR Facility  !

, SRWO 94 I8 (l1/29/94), #0-95 8 (06/30/95) {

. 1 The CENSOR facility complements the Irradiation Assisted Stress Corrosion- ,

- Cracking (IASCC) experiment. The objective of the SENSOR expenment is to place  !

senwrs in a loop that replicates the water chemistry of a Boiling Water Reactor (BWR) and j then to place that loop in the core of the MIT Research Reactor (MITR II). The sensors are  ;

~

to measce crack growth in situ and simultaneously monitor the electrochemical potential

- (ECP) with the objective of detennining if a relation exists between crack propagation and ECP. Also, the experiment seeks to determine if hydrogen injection can arrest crack  :

g.owth. Several types of specimens will be used. Some will be thermally sensitized and  !

att expected to crack almost at once. Others will not have been presensitized. (Nete: }

Sensitization is achieved by causing chromium de  :

he donc either thermally or via neutron ) Theirradiation.pletion facility is described in detail at the in thegrain boundarieL

" Final Safety Evaluation Repon (SER) for the Sensor Irradiation Facility. [

In addition to the safety evaluation report, piocedures for the installation and -

removal of the SENSOR were prepared. These were similar to those developed earlier for the PCCL, BCCL, and IASCC in-core experiments. An ALARA plan was also prepared

' for the SENSOR experiment.

The SENSOR experiment was installed in the MITR core in March 1995 and it ran almost continuously until late June when all planned experiments were complete. Data  !

analysis remains on going. The experiment appears to have been successful in all of its major objectives. In pan cular, the capability to affect crack growth through a change in water chemistry was shown.  !

'In FY97, the SENSOR ' experiment thimble and specimen were prepared and shipped offsite to GB using the OE-2000 shipping cask.

L l' 4 9

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Revision of the Surveillance Requirement on the EmetgcDey Batterv l

. SR&954 (02/22/95)

The emergency battery for the MIT Research Reactor was replaced in its entirety during FY95. Subsequent to that replacement, a change on the surveillance req uirements i I

for measurement of the battery's voltage and specihc gravity was identified as being desirable. Accordingly, a safety analysis was prepared and, following approval by the 4 MIT Reactor Safeguards Committee, submitted to the U.S. Nuclear Regulatory Commission (NRC) on 02/22/95. A request for additional information was received on 03/13/95 and a reply submitted on 04/10/95. A further request for additional information has been received and a reply is pending.

This reque. ; was withdrawn on July 31,1997 as newer internal Administrative Procedures were set up. l d

28-F, ENVIRONMENTAL SURyliXS Environmental monitoring is performed using continuous radiation monitors and dosimetry devices. The radiation monitoring system consists of G M detectors and associated electronics at each remote site with data transmitted continuously to :he Reactor Radiation Protection Office and recorded on strip chart recorders. The remote sites are krated within a quarter mile radius of the facihty. The detectable radiation levels per sector due primarily to Ar-41 are presented below.

Site Exposure (07/01/96 06/30/97i North 0.051 mrem East 0.862 mrem South 0.082 mrem West 0.088 mrem Green (east) 0.136 mrem Fiscal Year Averages 1997 0.2 mrem 1996 0.2 mrem 1995 0.4 mrem 1994 0.4 mrem 1993 0.5 mrem 1992 0.2 mrem

G. RADIATION EXPOSURES AND SURVEYS WITillN T11E FACILITY A summary of radiation exposures received by facility personnel and experimenters is given below:

July 1.1996 - June 30.1997 i Whole Body Ermsure Rance (rems) Number of Personnel No measurable .....................................................................)13 M eas u rab le - < 0.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 0.1 - 0.25 .....................................................................I1 0.25 - 0.5 .....................................................................:0 0.5 - 0.75 .................................................................... 0 2

0.75 - 1.00 .....................................................................

Total Person Rem = 7.29 Total Number of Personnel = 165 From July 1,1996 through June 30, 1997, the Reactor Radiation Protection Office provided radiation protection services for the facility which included power and non. power operational surveillance (perfctmed on daily, weekly, monthly, quarterly, and other frequencies as required), maintenance activities, and experimental proj,ect support. Specific examples of these activities include, but are not limited to, the followmg:

1. Collection and analysis of air samples taken within the contaimnent building and in the exhaust / ventilation systems.

2. Collection and analysis of water samples taken from the secondary, D2 0, primary, shield coolant, liquid waste, and experimental systems, and fuel storage pool.

3. Performance of radiation and contamination surveys, radioactive waue collection and shipping, calibration of area radiation monitors, calibration of effluent and process radiatior monitors, calibr: . ion 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, etc.

The results of all surveys and surveillances conducted have been within the guidelines established for the facility.

11. 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 ssible sources of such wastes during the year include cooling tower blowdown, the li uid waste storage tanks, and various sinks. All of the liquid volumes are measured, b far the largest being the i1,479,000 liters discharged during FY96 from the cooling towers. (Larger 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 2.27 E-5 Ci for FY97. The total tritium was 1.55 E-1 Ci. The total effluent water volume was 1.40 E7 liters, giving an average tritium concentration of 1.35 E 5 pCihnt.

The above liquid waste discharges are provided on a monthly basis in the following Table 113.

All releases were in accordance with Technical Specification 3.8-1, including Part 20. Title 20, Code of Federal Regulations. All activities were substantially below the limits specified in 10 CFR 20.2003. Nevertheless, the monthly tritium releases are reponed in Table 113.

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 below the limits after 4

the authorized dilution factor of 3000 with the exception of Ar-41, which is reported in the following Table 11-1. The 1671.06 Ci of Ar-41 was released at an average concentration of 0.415x10-8 pCi/ml. This represents 41.5% of EC (Effluent Concentration (lx10-8 pCihnl)).

3. Solid Waste Only one shipment of solid waste was made during the year. The information pertaining to this shipment is provided in Table 112.

I

. TABLE H Part A ARGON-41 STACK RFI FASES FISCAL YEAR 1997 Ar-41 Average Discharged Concentration (l)

(Curies) (pCi/ml)

July 1996 154.80- 4.93 E 9.

August 154.80 ' 3.92 E-9 l September 65.98 2.08 E-9 October 34.94 0.89 E 9 November 161.36 5.14 E-9 December 94.81 3.02 E-9 ,

January 1997 174.20 4.44 E 9 -

February 148.39 4.73 E-9 ,

March 87.71 2.79 E 9 April 208.58 6.64 E 9 May 162.89 4.15 E-9 June 222.60- 7.09 E-9 Totals (12 Months) 1,671.06 4.15 E-9 EC (Table II, Column 1) 1 x 10-8 ,

% EC 41.5 %

0)After authorized dilution factor (3000). (No.te: Average concentrations do not vary linearly with curies discharged because of differing monthly dilution voltmes.)

i n

+vvy d - c-a'-v-y =e w + +-r e v - g - -r y++ grv-t1 y es

TABLE H .

SUMMARY

OF MITR 11 RADIOACTIVE SOI in WASTE SHIPMILNTS FISCAL YEAR 1997 Description Volume . 105 ft3 Weight 3216 lbs.

Activity (l) . 0.007 Ci (257 MBq)

Date of sidpment August 14,1996 Disposition to licensee for burial Barnwell Waste Management Facility, Barnwell, SC Waste broker Scientific Ecology Group, Inc.,

Oak Ridge, TN Notes: (1) Radioactive waste includes dry active waste comprised of contaminated and/or Irradiated items and dewatered resin. The pnncipal radionuclides are activation and fission products such as 60Co,58Co, SICr,65Zn.187W, 125 Sb, 95Z.*, 95Nb, 311. 46Sc,103Ru,137Cs, 55pe, 63Ni,1:91, 99Tc, IdC, liomAg,54Mn,144Ce and idlCe.

l L

i i

TABIE H.3 -

t LIOU1D EFTLUENT DISCHARGES  :

FISCAL YEAR 1997 i l

Total Total Volume Average Activity Tritium ofEffluent Tritium 1xss Tritium Activity Wate80 Concentration j (x10-6 Ci) (x10 3 Ci) (x104 liters) (x10 6 pCi/ml) f r

July 1996 3.76 4.73 106. 4.46 i Aug. 0 1.28 54.8 2.34 l Sept. 3.76 55.5 93.7 5.92-Oct. NDA 2.89- 140 2,06  ;

Nov. 5.68 .16.5 74.6 22.1 .

Dec. -

NDA 2,14 78.6 2.72 ,

Jan.- 1997 NDA -4.03 70.8 5.69 Feb.- 9.53 49.6 87.2 56.9 Mar. NDA 2.74 80.8 3.39  !

Apr. NDA 3.65 16.3 22.4 t

May NDA 2.35 71.9 3.27

~ June NDA 9.02 129 6.99 12 months 22.73 154.43 1003.7 11.52  ;

(O . Volume of effluent from cooling towers, waste tanks, and NW12-139 Engineering Lab >

sink. Does not include other diluent from MIT estimated at 2.7 million gallons / day. .

(2t No Detectable Activity (NDA); less than 1.26x10-6 pCi/ml beta for each sample. l k

, # ~ , ,, ...._.3%...,__.,_m,rm ,. w . .,...,_,,,_y. , . _ , _ . _,..,_,-.. ,7__.,,m .

,,...__.,._4_ , ..,cy.y, ., , . _ . .

34 -

1. S.UMMARY OF USE OF MEDICAL FACILITY FOR llUMAN TilERAPY

• The use of the medical therapy facility for human therapy is summarized here pursuant to Technical Specification No. 7.13.5(i):

1. Investigative Studies An important administrative change occurred in the program during this reponing I'"

" d '" '" "' '" * ' * * " * " ' ' " '""""'""'""'"*'"**"""""P"'""*d^8'""*'"'

J tate cifective March 22,"1997. Under the new arrangement, nuclear reactor facilities in Massachusetts will continue to be regulated by the NRC. Ilowever, other activities, such as those governed by medical use licenses now held by hospitals will be regulated by the State. This created an administrative issue for the MIT Keactor because the MIT program on neutron capture therapy involves participation by both MIT and affiliated hospitals. As a result, Prov sion No. 6.5 of the MIT Reactor Facility Operating License No. R-37 was modified so that the government imposed transition from NRC regulation to State regulation of medical use licensees would not cause any disruption of our on-going research 3rogram. The modification was solely to change the language in the MITR Technica Specification No. 6.5 and its associated Quality Management Program. No substantive changes of any type resulted. The new language allows human subject referrals to the MITR not only from NRC-approved but also from Agreement State-approved medical use licensees that have been authorized by either NRC or an Agreement State to utilize the MIT reactor's medical therapy facility for neutron capture clinical trials.

The request for modification was approved by the MIT Reactor Safeguards Committee on February 18,1997. It was also approved by the NRC on April 3,1997.

In addition to the administrative change to the program's documents. FY97 saw the merging of the MIT medical partner New England Deaconess llospital(NEDil) with the Beth Israel llospital. The resuhing organization is Beth Israel Deaconess Medical Center which is now also a major teaching hospital of the liarvard Medical School.

During FY97, the major DNCT effort had been on the continuation of Phase I trials for glioblastoma. Phase I studies are required by the U.S. Food and Drug Administration.

The purpose is to investigate the toxicity (or lack thereof) of a proposed therapy. No beneht is expected to those participating in these studies. There are three Phr.a 1 trials in progress. Each is sununarized in the following.

a) Original Phase I Melanoma Study with Tufts New England Medical Center (NEMC)

This study was begun in September 1994. The approach used for this protocol implementation is for the subject to be given a test dose (400 mg/kg) of the boron-containing drup (BPA). Illood and punch biopsy samples are then taken in order to determine the biodistribution of the boron in both healthy tissue and tumor over time. This is necessary because the u 3take of boron in tumor varies markedly from one person to another. The irradiations t iemselves are done in four fractions. For each, the subject is i given 400 mg/kg of BPA and a limited number of blood / biopsy samples are taken to confirm the previously measured uptake curve. The starting pomt in the Phase I ptotocol was a total dose to healthy tissue of 1000 RBE-cGy. After the third subject, this was increased to 1250 RDE cGy.

Four subjects participated during 1994 and 1995, and a summary of their responses was given in our annual reports for FY95 and FY96. During FY97, three subjects had shown significant tumor regression and one has slight shrinkage of the tumor.

This Phase I protocol is now continued under the sponsorship of the Beth Israel-

., Deac.oness Medical Center.

b) Phase 1 Mdanoma Study with New England Deaconess 11ospital(NEDID The protocol adopted here was the same as that used for the NEMC study except that: (i) the boronated drug (BPA) was introdu:ed intravenously (IV) and the total dose 1250 RBE-cGy was delivered in one fraction. The use of IV BPA greatly increases boron ,

uptake and hence dose to tumor. One subject has been irradiated thus far under this protocol, as summarized in the annual report for FY96. This subject has shown shrinkage of the tumor.

c) Phase I Glioblastoma Study with Beth Israel Deaconess Medical Center This protocol is similar to the NEDit melanoma study in that it uses IV BPA. The total dose is delivered in ene fraction in a bilateral opposed manner. During this reporting period, three subjects were irradiated at a total dose of 880 RBE-cGy, three at 970 RBE-cGy, and two at 1065 RBE-cGy. A summary of these irradiations is as follows:

Subject 96-2 52 year old female, irradiation date 07/25/96, tumor unchanged, location of irradiation: parietal lobe, thalamus.

Subject 96-3 69 year old female, irradiation date 08/01/96, tumor unchanged.

Location ofirradiation: parietallobe,temporallobe.

Subiect 96-4 56 year old male, irradiation date 11/21/96, tumor regression observed. Location ofirradiation: parietallobe, temporallobe.

Subject 97-1 65 year old female, irradiation date 01/30/97, tumor unchanged, location ofirradiation: post parietal.

Subject 97-2 53 year old male, irradiation date 02/28/97, status not yet known.

I ocation of irradiation: frontal lobe, parietal lobe.

Subject 97-3 56 year old male, irradiation date 03/26/97, complete tumor regression. Location of irradiation: occipital lobe (melanoma metastasized to brain).

Subject 97-4 64 year old female, irradiation date 04/10/97, status not yet known.

Location ofirradiation: parietallobe.

Subiect 9"i-5 58 year old female, irradiation date 04/24/97, tumor unchanged.

Location ofirradiation: post-parietal.

Subject irradiations are continuing under this Phase I protocol.

2. Iluman Theracy None.

l

.i.

Compamtive Data MITR IN 1997 Mnual Renon SECTION G: RADIATION EXPOSURE Whole Body Exposure Range (rems) No. of Personnel IW97 FY96 FY95 1W94 FY93 FY92 N o M e a s u rabl c . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . I13 116 122 149 158 126 Measurable.................................... 29 22 44 31 25 47 0,1-0.25.................................. I1 12 14 9 14 8 0.25-0.5................................... 10 8 10 16 6 9 0.5-0.75................................... 0 4 4 2 6 7 0.75-1.0................................... 2 2 0 1 1 3 1.0-1,25................................... 0 0 0 0 0 1 1.25-1.5................................... 0 0 0 0 0 2 i

Total Personnel- 165 164 194 208 210 203 Total Person - Rem 7.29 9.12 9.81 10.73 10.16 16.37 1

I

I 1

i

~

S SECTION H: RADIOACTIVE EFFLUENTS f EY 94 EY_22 FY % FY 95 fFY 93 hi22 16/94 F12/93 { j

1. GASEOUS RELEASESH2-a) Ar 41 from stack (Ci) 1671 1587 1280 399 276 923 728 t Average concentration (l x 10-s pCi/ml) 0.415 0.448 0.32 0.22 0.16 0.24 0.20 ,

6.0 4.9 I

- <l/94: Percent MPC (4 x 10-8 pCi/ml) 4.1

>lN4: Percent EC (1 x 10-8 pCi/ml) 41.5 44.8 32.0 21.7 i b) H 3 from stack (Ci) .26.6 9.76 6.57 8.50 1.79 3.1 11.5 Average concentration (1 x 10-1I pCi/ml) 6.63 2.63 1.7 5.2 1.0 0.8 2.9

<!/94: Percent MIC (2 x 10 7 pCi/ml) .

0.005 0.004 0.01

>l/94: Percent EC . (1 x 10 7 pCi/ml) 0.066 0.026 0.017 0.052 P

c) H 3 from cooling towers (Cl) . 0.112 0.027 0.029 0.007 0.002 0.01 0.03 0.8 Average concentration (1 x 10-H pCi/ml) 5.07 0.869 1.3 0.9 1.0 1.5

<lN4: Percet t MIC (2 x 10 7 pCi/ml) 0.005 0.004 0.01

>l/94: Percent EC (! x 10 7 pCi/ml) 0.051 0.009 0.013 0.009 4 Fistion products from stack 4.0 8.6 12.6  ;

Percent MPC(details on next page) .

Percent I!C (details on next page) 3.4 7.4 16.6 11.9

2. LIQUID RELEASES TO >

. SANITARY SEWERt2>

i a) 113 from cooling towers (Ci) - 0.039 0.007 0.017 0.004 0.001 0.003 0.009 .

Average concentration (I x 10-dpCi/ml) 0.035 0.005 0.009 0.007 0.003 0.005 0.014  !

<l/94: Percent MIC(3)(1 x 10-1 pCihn!) 0.0003 0.001 0.001 [

>l/94: Percent EC(3) (I x 10 2 pCi/ml) 0.035 0.005 0.009 0.007 i b) H 3 from waste tanks (Ci) 0.116 0.089 0.002 0.019 0.001 0.004 0.014 Averaye concentration (1 x 10-dpCi/ml) 34.9 0.016 0.564 0.012 0.004 0.007 0.022

<1/94: Percent MIC ~ (1 'x' 101 pCi/ml) 0.0004 0.001 0.002 .

o .>l/94: Percent EC - (1 x 10 2 Ci/ml) 34.9 0.016 0.564 0.012 c) H-3 from laboratory sinksH3 (pCi) - 0.035 0.022 0.009 0.018 10.7 1.1 0.2  ;

Average concentration (I x 10-6pCi/ml) 0.002 0.001 5.9E-4 0.019 0.196 1.9E-4 2.8E-5

<1/94: Percent MIT (I x 10-3 pCi/ml) 0.0002 1.9E 9 2.8E-8

>l/94: Percent EC (! x 10-2 pCi/ml) 0.00002 0.0001 0.0001 0.0002 4 Total activity. cxcluding H 3 (Ci). L278-5 1.08E-4 5.9E-5 $.9E-5 3,9E 5 0.3E 3 1.4E 2 Nolcs: .(l) Average concentrations of gaseous stack wastes include authorized dilution factor of 3000.

(2). Concentrations are averned 'n total liquid release volume; no credit taken for dilution by other MIT discharges estimated at 2.7 million gallons / day.

'(3) Technical Specification 3.8-1.b limits cooling tower concentration to 1 x 10-3 pCi/ml.

Multiply

• Percent MPC" by 100 and " Percent EC" by 10 to get percent of Tech Spec limit.

(4) Engineering Lab sink (approved discharge point for experimenter samples generated in the i Engineering Lab) and RRPO sink. 4

, -ed

..s.,e- ,.-.--+ . . , .< . . - .m, , .. w- *-,.,,.h r e - n.4.- m--b w..,--w - - . - ~ 5, -:.,c -

-iii-Summarv of the Radionuclides identified

- > in the MITR 11 Stack Effluent for Fiscal Year 1997.

Annual release from the stack for gaseous radionuclides other than 41 Ar and 311 based on representative samples:

FY 97 (Jul) 1.1996 -June 30,1997)

Nuchde Average Elfluent Concentration Concentration I!C Percent of (prilinl) (pCihnli liC K5 m g, 3.31;. ] 113 07 3.3E.2 1311 10 2E-08 6.6E l h7Kr 1.511 10 9E4N 1.6E 0 88Kr 1,3E.10 4E-08 INil 1 135Xe 5.4E Il 4E-08 1.3E-l 135:nxe 2E-08 5 .5 11- 1 13kXe 1.lE-10 1.4U.I 1 SE-07 2.9E-3 I33Xe 211-10 1.61!-1 1311 3.21!-13 1.71;.15 2E-08 8.6E.6 132l 8.29B 17 111-09 8 311-6 1331 1.66E 15 6E-08 2.8116 1341 3.4 percent of EC fiscal > car 1997 7.4 percent of I!C-fiscal > car 1996 16.6 percent of EC - fiscal year 1995 4 0/l1.9 percent of hilC/EC - 1st/2nd half fiscal year 1994 R.6 percent of hilt - fiscal year 1993 12.6 percent of hilt - fiscal year 1992 his sununary of annual releases from the stack is for gaseous radionuclides other than 43 At and 3fl based on representative samples wc4hted according to reactor power, nonnalised to 4.5 hlWt. These salues are averaged over the year and include dilution factors allowed within the radiological effluent section of the h11TR ll Technical Specifications.

Notet (1) EC is " Effluent Concentration" per 10 CI:R 20, App.11.

(2) hilC is *htasimum Permissible Concentration" per pre-199410 CFR 20, App.11.

YNY WLJW-gFr6derick FMeWilliams Reactor Radiation Protection Officer August 1997

__ _ _ _ _