ML20072T304

Annual Rept for Mit Research Reactor for 930701-940630

ML20072T304
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	06/30/1994
From:	Bernard J, Lau E, Newton T NUCLEAR REACTOR LABORATORY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9409150165
Download: ML20072T304 (33)

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g,i NUCLEAR REACTOR LABORATORY I,gASpf{ld AN INTERDEPARTMENTAL CENTER OF hig7 MASSACHUSETTS INSTITUTE OF TECHNOLOGY O. K. HARLING 138 Albany Street Cambndge Mass 02139-4296 J. A BERNARD JR.

Director Telefax No. (617) 253-7300 Director of Reactor Operations Telex No. 92-1473-MIT CAM Tet. No. (617) 253-4211/4202 August 30,1994 U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 A1TN: 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 1,1993 to June 30, 1994, in compliance with paragraph 7.13.5 of the Technical Specifications for Facility Operating License R-37.

Sincerely, /

'7 f a x f' _

Edward S. Latt(NE Asst. Superintendent for Engineering 4  % / '..

Thomas H. Newton, Jr., PE

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Asst. Superintendent for Operations MIT Research Reactor MIT Research Reactor t 4 ohn A. Bernard, Ph.D.

Director of Reactor Operations MIT Research Reactor JAB /gw

Enclosure:

As stated cc: USNRC - Project Manager, NRR/ONDD USNRC - Region I- Chief, Effluents Radiation Protection Section (ERPS)

FRSSB/DRSS 000CCS 9409150165 940630 f

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MIT RESEARCH REACTOR NUCLEAR REACTOR LABORATORY MASSACHUSETTS INSTITUTE OF TECHNOLOGY l

i ANNUAL REPORT to United States Nuclear Regulatory Commission f or the Period July 1,1993 - June 30,1994 l

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REACTOR STAFF August 30,1994 l

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i Section Page I

Tabl e of Con te n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i l

In trod u ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4

i A. Summary of Operating Experience ................................................. 3 l

j B. Reac tor Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 l C. Shutdowns and Scrams ............................................................12 1

D. Major Maintenance .................................................................14 j E. S ection 50.59 Changes, Tests, and Experiments . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 F. Environmental S urveys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 G. Radiation Exposures and Surveys Within the Facility ........................... 27  !

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H. Radioactive Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 1

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I. Summary of Use of Medical Facility for Human Therapy ...................... 32 i 1

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MIT RESEARCH REACTOR ANNUAL REPORT TO U. S. NUCLEAR REGULATORY COMMISSION FOR THE PERIOD JULY 1.1993 - JUNE 30.1994 ,

INTRODUCTION Ris report has been prepared by the staff of the Massachusetts Institute of Technology Research Reactor for submission to the Administrator of Region I, United States Nuclear Regulatory Commission, in compliance with the requirements of the Technical Spectfications 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.

We MIT Research Reactor (MITR), as originally constructed, consisted of a core

, of MTR-type fuel, fully enriched in uranium-235 and cooled and moderated by heavy water i in a four-foot diameter core tank, surrounded by a graphite reflector. After initial criticality l on July 21, 1958, the first year was devoted to startup experiments, calibration, and a I gradual rise to one megawatt, the initially licensed maximum power. Routine three-shift )

operation (Monday-Friday) commenced in July 1959. He authorized power level was I increased to two megawatts in 1962 and to five megawatts (the design power level) m ,

1965. l Studies of an improved design were first undertaken in 1967. He 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 undermoderated 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 UAL xintermetallic 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-I 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 Commission issued Amendment No.10 to Facility Operating License No. R-37 on July 23,1975. After initial criticality for MITR-II on August 14,1975, and several months of startup testing, power was raised to 2.5-MW in December. Routine 5-MW operation was achieved in December 1976.

This is the nineteenth anwal report required by the Technical Specifications, and it covers the period July 1,1993 througn June 30,1994. Previous reports, along with the "MITR-II Startup Report" (Report No. MITNE-198, Februanj 14, 1977) have covered the

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4 startup testing period and the transition to routine reactor operation. 'Ihis report covers the seventeenth full year of routine reactor operation at the 5-MW licensed power level. It was j another year in which the safety and reliability of reactor operation met the requirements of

{- reactor users.

i j' A summary of operating experience and other activities and related statistical data j are provided in Sections A-I of this report.

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A.

SUMMARY

OF OPERATING EXPERIENCE

1. General Re MIT Research Reactor, MITR-II, has in recent years been operated on a routine, Sve days per week schedule, modified as necessary to facilitate the preoperational testing and installation of several in-core expenments. When operating, the reactor is j normally at a nominal 5-MW. However, as was the case for the last four years, substantial departures were made from this schedule during the period covered by this report (July 1,1993 - June 30,1994). Specifically, for several months during 1994, the reactor i was run 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). This schedule was followed in order to support a major expenmental program concerning the development of methods to reduce the activation and transport of corrosion products in pressunzed water reactor coolant. The period covered by this report is the seventeenth full year of normal cperation for M'TR-II.

The reactor averaged 46.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per week at full power compared to 61.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per week for the previous year and 69.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per week two years ago. As was the case in FY 93 a lot of operation was conducted at low power in order to make measurements of the i medical therapy facility beam. Rese measurements are for the purpose of maintaining an

. epithermal neutron beam for the treatment of brain cancer (glioblastoma multiforme) and I possibly skin cancer (melanoma). When neither the corrosion reduction experiments nor the medical beam design was in progress, the reactor was usually operated from late Monday aftemoon until late Friday afternoon, with maintenance scheduled for Monday momings and, as necessary, for Saturdays.

The reactor was operated throughout the year with 24 elements in the core. The remaining three positions were used as follows: position Al was occupied, during the early part of FY 94 by the Pressurized Water Reactor (PWR) Coolant Chemistry Ioop (PCCL) and later on (December 93 - April 94) by the Boiling Water Reactor (BWR)

Coolant Chemistry Icop (BCCL). These two loops reproduce chemistry conditions in power reactors and are part of a major effort to identify methods for reducing radiation exposures in the nuclear industry. De second non-fueled position, B3, was occupied by the Irradiation-Assisted Stress Corrosion Cracking Facility (IASCC) during the latter half of FY 94. The third core position, A3, was occupied by a solid aluminum dummy as were

positions Al and B3 whenever the PCCIJBCCL or IASCC was not installed.

Compensation for reactivity lost due to burnup 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 fuel in the outer portion of the core (the C-Ring). In addition, elements were inverted and rotated so 4

as to achieve more uniform burnup gradients in those elements. Birteen other refuelings were performed for the purpose of making accurate reactivity measurements and trial fits of the various Coolant Chemistry Loop experimental facilities.

The MITR-1: fuel management program remains quite successful. All of the original MITR-Il elements (445 grams U-235) have been permanently discharged. The j average overall burnup for the discharged elements was 42%. (N_ qts: One element was removed prematurely because of excess outgassing.) Le maximum overall burnup achieved was 48%. Seventy-one of the newer, higher loaded elements (506 grams U-235)

, have been introduced to the core. Of these, twenty-eight have attained the maximum

allowed fission density. However, some of these may be reused if that limit is increased as would seem warranted based on metallurgical studies by DOE. Another nine have been

1 identified as showing excess outgassing and have been removed from service. He other thirty-four are either currently in the reactor core or have been partially depleted and are 1 awaiting reuse in the C-ring.

J Protective system surveillance tests are conducted on Friday evenings after J shutdown (about 1800), on Mondays, and on Saturdays as necessary. 1 As in previous years, the reactor was operated throughout the period without the l fixed hafnium absorbers, which were designed to achieve a maximum peaking of the ,

thermal neutron flux in the heavy water reflector beneath the core. Rese had been l removed in November 1976 in order to gain the reactivity necessary to support more in- (

core facilities.

2. Exoeriments The MITR-II was used throughout the year for expenments and irradiations in support of research and training programs at MIT and elsewhere.

Experiments and irradiations of the following types were conducted:

a) nampt gamma 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. l 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 properties of polymers. l l

c) Studies of the material composition of superconducting phases of various alloys l were performed by activating samples and then idertifying characteristic radiations. l 1

d) Irradiation of archaeological, environmental, engineering materials. biological, I geological, oceanographic, and medical specimens for neutron activation analysis l purposes.

e) Production of dysprosium-165 and holmium-166 for medical mevch, diagnostic, and therapeutic purposes.

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

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

1 h) Use of the facility for reactor operator training.

i) Irradiation of geological materials to determine gur.ntit:es and distribution of fissile materials using solid state nuclear track detectors.

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 (rain).

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1) Evaluation of the efficacy of neutron capture therapy using animal (mice) models.

m) Measurements of the energy spectrum of leakage neutrons using a mechanical I chopper in a radial beam pon (4DH1). Measurements of the neutron wavelength by l Brag;; a0ection then permits demonstration of the DeBroglie relationship for

- physics corses at MIT and other universities.

n) Gamma irrullation of seeds for demonstration of radiation damage effects for high j school students.

o) Expenmental evaluation of flux synthesis methods as a means of estimating reactivity, p) Replice: ion of the radiation environment in space for the study of possible methods l oflow temperature annealing of electronic devices that would be used in spacecraft. I q) Neutron activation analysis of serum samples in an effort to correlate mineral I i denciencies with certain diseases.  !

1 r) Determination of uranium concentrations in samples of mica.

s) Operation of an in-core slow strain rate testing rig to evaluate irradiation-assisted I stress corrosion cracking of metals.  ;

t) Evaluation of scintillation fluids for use in neutrino detectors.

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Dose reduction studies for the light water reactor industry began reactor use on a regular basis in 1989. (Planning and out-of-core evaluations had been in progress for several years.) These studies entail installing loops in the reactor core to investigate the chemistry of corrosion and the transport of radioactive 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 third loop, one for the study of irradiation-assisted stress corrosion cracking, became operational in June 1994.

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. 1 4

'Ihis is a collaborative effort with the Tufts - New England Medical Center.

3. Chances to Facili!yfesyn 4

Except for minor c'ianges reported in Section E, no changes in the facility design were made during the yea". As indicated in past reports the uranium loading of MITR-Il fuel was increased from 19.7 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 has been good.

(Please see Reponable Occurrence Reports Nos. 50-20n9-4, 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 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 International completed the production of forty-one of the more highly loaded elements in 1982, forty of which have been used to

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6-some degree. Twenty-eight with about 40% burnup have been discharged because they I have attained the fission density limit. Four other AI elements remain in use. Of the otl:er 7 eight, six were, ac previously reported to the U.S. Nuclear Regulatory Comnussion.

removed from sera because of excess outgassing and two were removed because of suspected excess outgassing. Additional elements are now being fabricated by Babcock & I 1

Wilcox, Navy Nuclear Fuel Division, Lynchburg, Virginia. Riny-one of these have been 3

received at MIT, thiny of which are in use. One has been removed because of suspected excess outgassing.

He MITR staff has been following with interest the work of the Reduced Enrichment for Research and Test Reactors (RERTR) Program at Argonne National i 3-Laboratory, particularly the development of advanced fuels that will #t uranium 1

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loadings up to several times the recent upper limit of 1,6 grams total uranium / cubic i centimeter. Consideration of the thermal-hydraulics and reactor physics of the MITR-II

core design show that conversion of MITR-II fuel to lower ennchment must await the I

! successful demonstration of the proposed advanced fuels.

4. Chances in Performance Characteristics

Performance characteristics of the MITR-II were reported in the "MITR-II Startup Report." Minor changes have been described in previous reports. There were no changes

during the past year.

5. Changes in Ooerating Procedures Related to Safety f

One amendment to the facility operating license was requested during the past year.

It involves a request that the Facility Operating License No. R-37, which is due to expire in May 1996, be extended to April 2001 in order to recover time during which the reactor was either under construction, shutdown for modification, engaged in low-power testing that was a prerequisite for operation at the authorized power level, or engaged in core modification found to be necessary as a result of the low-power testing. Also, an amendment concerning a revision to the fission density limit for the reactor fuel had been previously requested. Both are discussed in Section E of this report.

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.

Rose changes related to safety are discussed in Section E of this report.

a) Procedures were developed and issued for the calibration of and the response to an ammonia detector that was installed in the ventilation intake. This was done to alert the control room operator of any possible incipient problem with the ammonia refrigerant used at a nearby non411T facility. Hat facility originally used freon.

He switch was made to ammonia to preclude ozone layer damage. (SR#0-93-12) I b) Several minor changes were made to the daily and weekly security checklists.

(SR#0-93-13) I c) A system was installed to permit adjustment from the control room of the damper that controls the containment building's differential pressure. This eliminates the necessity of someone going to the stack base every time that an adjustment is needed. (Passing weather fronts create the need for adjustments.) His is both an ALARA and a personnel safety improvement. (SR#0-93-15)

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d) The procedures for "Startup for less than 100 kW Operation", " Shutdown from I less than 100 kW Operation," and " General Conduct of Refueling Operations" I were revised to reflect changes in instrumentation and to incorporate minor i proceduralimprovements. (SR#0-93-17) e) Procedures for the calibration of the reflector, shield, and secondary coolant system )

flows were revised following purchase and installation of a new flow recorder.

(SR#0-93-25)

I f) A procedure was written governing the transfer of irradiated components from the coolant chemistry loop disassembly area to the spent fuel storage pool. A separate tank was set up within that pool for the temporary storage of used components from the loops. (SR#0-93-28) g) De Administrative Procedures, Chapter One of the Procedure Manual, were revised to update lists of names and committee memberships. (SR#0-94-1) h) A modification was made to the circuitry of the low-range amplifiers to correct a design deficiency that was identified in Reportable Occurrence No. 50-20/1993-1.

(SR#E-94-1) i) He emergency core cooling system was modified to ensure that all fuel elements would receive the minimum required spray despite the presence of several planned in-core experiments. This was done by extending both spray nozzles so that the point of discharge would be closer to the core and by adding two auxiliary nozzles to the side of each new one. Details of the change were provided to the U.S.

Nuclear Regulatory Commission in our letter of February 18,1994. (SR#0-94-3) j) The abnormal operating procedure for a response to a "High Radiation 12 vel Core Purge" was temporarily modified to remove the automatic isolation of the system.

His was done to investigate a series of very brief trips. The automatic isolation feature will shortly be restored. (SR#0-94-9) k) A procedure was approved to measure the temperature distribution of the primary coolant in the space above the core following a loss of primary flow. The objective was to observe the establishment of natural circulation cooling and to' obtain sufficient experimental data so as to benchmark a computer-based model of the thermal-hydraulic behavior.of the MIT Research Reactor upon loss of forced circulation. (SR#0-94-11) i

1) The procedures for calibration of the primary systen '.emperature probes were revised to reflect the acquisition of new calibration equipment, a Stow Labs Model 921A digital thermometer and two precision RTD temperature probes.

(SR#0-94-13) m) A number of minor changes were made to the daily surveillance check.

(S R#0-94-14) n) The procedure for verifying operability of the emergency core cooling system was modified to allow for the physical change that was made in the system pursuant to SR#0-94-3. (SR#0-94-15) l

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o) He procedure for verifying operability of the cooling tower sprinkler system was revised to incorporate changes made in fire protection systems for MIT as a whole.

(SR#0-94-16) j p) A machine for the neutron transmutation doping (NTD) of silicon ingots was l designed, built, and installed in two of the throughports (4TH1 and 6m1) that run below and to the side of the reactor core. This machine and its operation have no effect on the reactor itself. That is, there is no feedback path from the ingots to the reactor core. The machine's design and operation are complex because ingots are I

rotated as they move through the port and their net rate of movement is carefully controlled so as to achieve the desired change in the resistivity of the silicon.

Standard and abnormal operating procedures were prepared for this machine as was '

a pre-operational test package. (SR#0-93-11, #0-93-18, #0-93-19, and #0-93-21) q) The emergency plan for the MIT Research Reactor was modified extensively during FY 94 in order to accommodate the revisions mandated by 10 CFR 20. Changes included revised definitions of radiological terms, the calculation of those EALs that were based on MPC figures in the original 10 CFR 20, the procedure for an escape of airborne radioactive material, and the instructions to the MIT Campus Police for surveys, he changes in the EALs necessitated the changes in the procedure for an airborne release and in the instmetions to the police. A new stack area monitor was purchased and installed and new (digital) survey meters were purchased for the MIT Campus Police. All of the changes made to the emergency plan either increase or have no impact on the plan's effectiveness. A complete copy of the plan will be i provided_ as a package to the U.S. Nuclear Regulatory Commission as soon as l several additional planned changes are completed. (SR#0-93-21, #0-93-26, '

l #0-93-27, #0-93-29, #0-94-9, and #0-94-10) l l .

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6. Surveillance Tests and Insoections i

nere are many written procedures in use for surveillance tests and inspections i required by the Technical Specifications. Rese procedures p,rovide a detailed method for

conducting each test or inspecuon and specify an acceptance entenon which must be met in l

- 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 i frequency at least equal to that required by the Technical Specifications. Twenty-seven such tests and calibrations are conducted on an annual, semi-annual, or quarterly basis.

Other surveillance tests are done each time before startup of the reactor if shutdown '

1 4 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.

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Dunng the reporting period, the surveillance frequency has been at least equal to

) that 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 Soent Fuel Shioment i Pursuant to Amendment No. 25 to Facility Operating License No. R-37, paragraph j 2.B.(2) subparagraph (b), reponed herewith is the status of the establishment of a shipping

capability for spent fuel and other activities relevant to the temporary merease in the l possession limit.

! MIT began efforts for spent fuel shipment as early as 1983. At that time, the plan was to use two MH-1A casks that had been acquired by DOE and which were being prepared for use by the non-power reactor community. After an MH-1A cask became

unavailable, MIT made arrangements with General Electric to use the GE-700 cask for j shipment of the MITR spent fuel. When the GE-700 cask was removed from service voluntarily by GE, the BMI-1 cask became the only one available that is approved for 4 4,

transportation of irradiated fuel elements. l i

i ne capability to ship spent MITR fuel was established by the end of 1992. )

j Specifically, the following was accomplished:

(a) The Certificate of Compliance and the Safety Analysis Repon 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 l cask.

j (b) The University of Missouri Research Reactor (MURR) basket was reviewed and

! found to be suitable for use with the MIT fud elements in the BMI-l cask. MURR has acreed to make their basket available to MIT for the required shipments.

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

l (d) The decay heat load of each spent element was determined by a member of the 1 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 i

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limits. Criticality calculations were performed using the Monte-Carlo Code KENO.

V which was obtained from the Radiation Shielding Information Center of the Oak )

Ridge National Laboratory. Results show that the degree of suberiticality of a cask fully loaded with MIT fuel elements is within specificanon.  ;

(e) In order to cross-check the cross 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. ,

I (f) Arrangements have been made with the fuel receiving organization at the Savannah River facility. Specific data on the MITR-II spent fuel elements were compiled. l The Appendix A document and criticality study were prepared and reviewed by the spent fuel processing center.

(g) Spent fuel elements in the MITR spent fuel storage pool were arranged and grouped l

in accordance with our procedure for shipment preparation. A special structure for i support of the BMI-l cask was designed and fabrictted.

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

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

(j) A criticality study of the BMI-l cask with fresh 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. J (1) Procedures for spent fuel shipment were prepared. l (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, i In each case, the spent fuel was returned to the U.S. Department of Energy's facility at i l Savannah River, SC. As a result of these shipments, the on-site inventory of U-235 was  !

l 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 1994, one shipment of eight elements was completed. 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 cutTently 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.

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

@ma 1 2 3 4 Total

1. Energy Generated (MWD):

a) MITR-II(MITFY94) 36.2 133.6 162.7 160.4 492.9 (normally at 4.9 MW) b) MITR-II 12'692.4 (MIT FY76-93) 1 c) MITR-I 10'435.2 (MIT FY59-74) d) Cumulative, MITR-I & MITR-II 23'620.5

2. MITR-II Operation (Hrs):

(MIT FY94) a) At Power

(>0.5-MW) for 184.6 685.4 859.8 876.2 2606.0 Research b) Low Power

(<0.5-MW) for 123.8 79.0 126.6 82.1 411.5 Training (l) and Test c) TotalCritical 308.4 764.4 986.4 958.3 3,017.5 (1) 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.

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C. SHUTDOWNS AND SCRAMS i During the period of this report there were 13 -inadvertent scrams and 32 j unscheduled power reductions.

4 j The term " scram" refers to shutting down of the reactor through protective system j 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 suberitical by the i reactor operator in response to an abnormal condition indication. Rod drops and electric

power loss without protective system action are included in shutdowns.

i j The following summary of scrams and shutdowns is provided in approximately the

same format as for previous years in order to facilitate a comparison.

l 5 1. Nuclear Safety System Scrams Tntal i

a) Channel #2 trip as result of amplifier malfunction. 1 l b) Channel #5 low range trip while inspectmg amplifier (ROR #93-1). 1 1

i c) Channel #4 trip as result of D20 shutter manipulation. 3 f

l d) Channel #1 high voltage power supply failure. 1 1

i e) Channels #1 and #6 trip as result of amplifier 4 malfunction. I u

4 Subtotal 7 II. Process System Scrams a) MTS-1 high temperature trip as result of calibration being overly conservative. 2

! b) low flow secondary coolant trip as result of frozen

cooling tower riser pipes. 1

! c) low flow secondary coolant trip as result of l electronic noise on flow recorder. 1 i

2 d) low flow shield coolant trip as result of operator

. error while inspecting flow recorder. 1 i

i- e) low flow primary coolant trip as result of failure of j pump MM-1. 1 l Subtotal 6 i

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III. Unscheduled Shutdowns or Power Reductions a) NTD Silicon machine malfunctions. 19 b) . Shutdown as result ofIASCC or BCCL experiment malfunction. 4 j c) Shutdown due to loss of offsite electricity. 1 l l

d) Blade #3 drop as result of high control room temperature. I e) Shutdown to investigate abnormallevels of argon. 1 f) Shutdown due to DM-2 pump (D2O system) failures. 3 l

g) Failures of cooling tower fans. 3 Subtotal 32 Total 45 l Experience during recent years has been as follows for scrams and unscheduled l shutdowns:

l i l Fiscal Year Number I

' l Scrams Shutdowns Inlal 90 11 9 20 l 91 11 9 -20

(. 92 5 12 17 93 6 14 20 l

94 13 32 45 l

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- 14 D. MAJOR MAINTENANCE Major maintenance projects performed during FY 94, including the effect, if any, I on the safe and reliable operation of the MIT Research Reactor are described in this l Section. i

\

One of the FY 94 maintenance items was the transfer of eight MITR-II spent fuel elements to the DOE Savannah River facility at Aiken, SC. He shipment was completed  !

safely and in a timely manner.

Much maintenance was performed to support the ongoing program to identify improved water chemistries that will result in reduced radiation exposure to workers in the nuclear industry. The projects are the Pressurized Coolant Chemistry Loop (PCCL) and the Boiling Coolant Chemistry Loop (BCCL) light-water reactor experiments. The BCCL (T4 thimble) was installed in the reactor core on November 15, 1993 and the run was successfully completed on April 5,1994. The PCCL (B4 thimble) was installed on June 6,1994. Much preventive maintenance was performed to support these operations.

Another maintenance item was the installation of the Irradiation-Assisted Stress

. Corrosion Cracking (IASCC) experiment. The IASCC was initially installed on May 16,1994. his installation included the development of procedures for rigging and handling, placement of auxiliary shielding, installation of an Instron tensile testing machine, and both plumbing and wiring of the experiment. Routine in-core testing of specimens began on June 8,1994, Installation of machinery for neutron transmutation doping of silicon required a

major maintenance effort. His machinery, installed in two of the reactor thmughports,

! includes two twenty-foot tubes for each port, rotating and pushing mechanisms, billet handling and storage conveyors, electronics, and associated microprocessor-based controllers.

i In order to support multiple in-core experiments, the reactor emergency core cooling system (ECCS) was modified. Prior to installation of the IASCC, it was decided to modify the existing ECCS nozzles so that in the event of a loss of coolant with multiple in-core experiments, the ECCS spray would still be evenly distributed across the core.

His modification was tested on a full-scale mock up of the core, approved by the MITR Safeguards Committee, and subsequently installed. 4 3

I i

Other major maintenance items performed in FY 94 were as follows:

(i) One of the two main pumps of the primary coolant system was rebuilt, the ball bearings repacked, and the shaft seals replaced. He result has been quiet, reliable operation of this pump.

(ii) The gasket on the inner door of the main personnel lock was replaced. De new gasket is made ofimproved braided rubber material and is more wear-resistant than the earlier gaskets. l l

Both cooling tower fans and motors were replaced with new units. The (iii) fans were replaced with fiberglass ones for iraproved efficiency.

, (iv) Infrared sensors were installed in two potential high radiation areas to provide audible alarms in accordance with the new 10 CFR 20.

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(v) A sink was installed on the reactor floor in order to facilitate easier cleaning and decontanunation.

(vi) New bus-bars were installed on the emergency batteries to replace corroded ones.

(vii) ne hydraulic system for the elevator traveling from the reactor floor to the reactor basement was refurbished.

(viii) New flow sensors and a new flow chart recorder were installed for the secondary, shield, and D20 coolant systems.

(ix) Three tubes which house proximity switches for " blade in" indication were replaced to prevent in-leakage of water.

(x) De control room air conditioner was replaced with a new unit.

(xi) De solenoid valve for the automatic operation of the medical H2O shutter was replaced.

(xii) ne magnet for shim blade #4 was replaced. l l (xiii) The D20 cleanup system pump and motor were refurbished and replaced I respectively.

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

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l E. SECTION 50.59 CHANGES. TESTS. AND EXPERIMENTS His 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 i 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 l t pages for ready reference if funher information should be required with regard to any itent Pertinent pages in the S AR have been or are being revised to reflect these changes, and they i either have or will be forwarded to the Document Control Desk, USNRC.

ne conduct of tects and experiments on the reactor are normally documented in the experiments and irradiation files. For expenments carried out under the provisions of 10 CFR 50.59, the review and approvalis documented by means of the Safety Review Form.

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

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Pressurized Coolant Chemistry 1000 (PCCL)

SR#0-86-9 (04/21/88), #0-88-4 (07/28/88), #0-88-5 (09/09/88), #0-88-14 (12M7/88),

1. 0-89-2 (014 6/89), #0-89-3 (01/19/89), #0-89-6 (01/24/89), #0-89-9 (06/02/89), #0-89-14 (06/19/89), #0-90-6 (03/20S 0), #0-90-7 (03/20S 0), #0-90-8 (03/20 S 0), #0-90-9 (03/20B0), #0-90-25 (12/10S0), #0-90-26 (12/18S0), #0-90-27 (12/18BO), #0-91-8 (05/21S 1), #0-91-21 (12/27S 1), #0-92-2 (01/27S 2), #0-92-12 (08/19S2), #0-94-6 (06/30S 4), #0-94-8 (03/10S 4).

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 pressurized water reactors (PWRs). He 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 determined 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 detemiination, together with safety review #0-86-9, were submitted to the U.S. Nuclear Regulatory Commission on 04/21/88.

Subsequent to the determination that no unreviewed safety question existed, specific procedures for PCCL operation were prepared. Rese included:

- Procedure for Ex-Core Testing

- Supplement to the Safety Evaluation Report

- Preoperational Test 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.

No design changes were made to the PCCL during the period covered by this report. For part of FY 94, this facility was used to evaluate the passivating effect of zine additives to PWR pnmary coolants.

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Boiline Coolant Chemistry Looo (BCCL)

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SR#0-89-14 (06/19/89), #0-89-20 (12/20/89), #0-90-17 (09/17/90), #0-90-18 (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. 0-93-10 (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 MITR 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 (crud) and thereby reducing radiation fields associated with boiling water reactors (BWRs).

The ultimate 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 experiments are not exceeded. It was also concluded that boiling in the proposed experiment volume would 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 operation. Following both a detennination that no unreviewed safety question was involved and approval by the MIT Reactor Safeguards Committee, this experiment was conducted. The results were as expected.

The final safety evaluation report for the BCCL was completed on 8 March 1989 and approved by the MITR Staff. On 12/20/89, the MIT Reactor Safeguards Committee determined that there was no unreviewed safety question involved in the conduct of the BCCL experiment and approved the BCCL SER. On 9 March 1990, a copy of the BCCL SER together with the safety analysis prepared by the MITR Staff were forwarded to the U.S. Nuclear Regulatory Commission pursuant to 10 CFR 50.59(b)(2).

Subsequent to the determination that no unreviewed safety question existed, specific procedures for BCCL operation were prepared. These included:

l

- Preoperational Test Procedure. i

- Abnormal Operating Procedures for the BCCL.

- Procedure for BCCL Startup. l l

Other necessary procedures such as BCCL shutdown and installation / removal are l the same as those previously developed and approved for the PCCL. Experiments using

the BCCL began in October 1990 and have been successful in that many theories concerning the transport of nitrogen-16 in boiling water reactors have been disproven.

During the period covered by this report, several changes were made to the BCCL experimental protocol. These were:

- Issuance of a third list of chemicals approved as additives. Each of these was to be l studied for its effect on the suppression of nitrogen-16 carryover.

- Approval of several minor modifications to the facility's abnomial operating procedures.

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- Approval of a design change that reduced the facility's mass and hence the magnitude of its gamma-ray heating. The original facility was contained within a lead bath. De lead attenuated many gamma rays and hence provided significant heating. For the FY 94 studies, a lower operaung temperature was desired. So, thelead bath was replaced by a solid aluminum block. His change was reviewed and approved by the MIT Reactor Safeguards Committee. The change did not involve an unreviewed safety question.

Experiments that make use of the BCCL facility were conducted during portions of this repomng period.

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Exoeriments Related to Neutron Cacture Thernov ~~

l SR'#0-89-4 (01/23/89), #0-89-8 (03/01/89), #0-91-7 (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), I

1. 0-93 9 (07/13/93),#0-93-20 (11/30/93).

In conjunction with the Tufts - New England Medical Center and with the support l of the U.S. Department of Energy, MIT has designed an epithermal neutron beam for the i treatment of brain cancer (glioblastoma). Themial beams have been used successfully for l this tacatment in Japan. The reason for designing an epithermal beam is to allow tumor l treatment without having to subject the patient to surgery involving removal of a portion of the skull. Also, an epithermal beam gives greater penetration. In October 1991, MIT i hosted an international workshop for the purpose of reviewing proposed beam designs and j dosimetry. Subsequent to the receipt of advice from the workshop panel members, a final  ;

design was selected for the epithermal filt:r for the MIT Research Reactor's Medical Therapy Facility beam. That design, whicn 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 Committees as well as from the U.S. Food and Drug l Administration. A% alicense 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 receipt of that license amendment and a similar one in August i

! 1993 for our medical partner, the Tufts - New England Medical Center, both procedures

! for performing BNCT and a preoperational test package were prepared. The latter was <

completed during FY 94.

1 The capability to initiate patient trials has existed since November 1993. Trials are i to be performed as part of a Phase I effort that is required by the FDA to demonstrate i toxicity or the lack thereof. Unfortunately, there have been no actual trials to date. Efforts i i are in progress to expand the patient pool. Simulated patient trials were conducted for training purposes during FY 94 and, as a result, minor revisions have been made to some of the applicable procedures.

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i Dicital Comouter Control of Reactors Under Steady-State and Transient Conditions

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l SR#M 81-3 (11/17/81), #M-81-4 (12/10/81), #E-82-2 (01/08/82), #E-82-3 (02/24/82),

! #E-82-4 (03/03/82), #E-82-5 (04/14/82), #E-82-6 (07/13/82), #0-83-5 (02/03/83), #E-83-1 i (02/08/83),#0-83-12 (04/23/83), #0-83-20 (07/20/83),#0-84-11 (06/25/84),#0 84.12 i' (07/12/84), #0-84-16 (12/06/84),#0-84-21 (11/01/84),#0-85-11 (05/09/85),#0-85-13 (06/28/85),#0-85-16 (07/12/85),#0-85 20 (08/16/85),#0 85-25 (12/01/85),#0-85-26

(12/01/85),#0-86-11 (10/17/86),#0 86-13 (11/28/86),#0-87-11 (06/01/87),#0-87-17

! (12/24/87),#0-88-10 (12/01/88),#0-90-28 (12/27/90),#0-91-2 (05/14/91),#0-91-3 l

(06/06/91),#0-91-10 (07/15/91),#0-91-11 (06/25/91),#0-92-6 (06/09/92),#0-93-3 l 1 (03/29/93),#0-93-23 (12/06/93),#0-94-12 (04/25/94). l

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The project involving computer analysis, signal validation of data from reactor i instruments, and closed-loop control of the MIT Reactor by digital computer was  ;

continued. A non-linear supervisory algorithm has been developed and demonstrated. It

! functions by restricting the net reactivity so that the reactor period can be rapidly made infinite by reversing the direction of control rod motion. It, combined with signal j validation procedures, ensures that there will not be any challenge to the reactor safety .

1 system while testing closed-loop control methods. Several such methods, including  !

decision analysis, rule-based control, and modern control theory, continue to be j experimentally evaluated. The eventual goal of this program is to use fault-tolerant computers coupled with closed-loop digital control and signal validation methods to demonstrate the improvements that can be achieved in reactor control.

Each new step in the program is evaluated for safety in accordance with standard i review procedures (Safety Review numbers listed above) and approved as necessary by the .

MIT Reactor Safeguards Committee.

l Initial tests of this digital closed-loop controller were conducted in 1983-1984 using the facility's regulating rod which was of relatively low reactivity worth (0.2% AK/K).

1 Following the successful completion of these tests, facility operating license amendment

{ No. 24 was obtained from NRC (April 2,1985). It permits:

I. (1) Closed-loop control of one or more shim blades and/or the regulating rod 5 provided that no more than 1.8% AK/K could be inserted were all the i connected control elements to be withdrawn, and (2) Closed-loop control of one of the shim blades and/or the regulating rod j provided that the overall controller is designed so that reactivity is 1 1 constrained sufficiently to permit control of reactor power within desired or  !

l authorized limits.

A successful experimentation program is now continuing under the provisions of

this license amendment. A protocol is observed in which the NRC-licensed supervisory i controller is used to monitor, and if necessary override, other novel controllers that are still

! in development. Tests of novel controllers are conducted under the provisions of technical j specification #6.4 which requires that reactivity be constrained to ensure " feasibility of

control." Signal implementation is accomplished using a variable-speed stepping motor.

This motor is installed prior to the tests and removed upon their completion. An ,

independent hard-wired circuit is used to monitor motor speed and preclude an overspeed l 3

condition. This arrangement for the conduct of these tests has been approved by the MIT i Reactor Safeguards Committee.

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1 An extensive upgrade to the digital control system's hardware was performed in l 1991. The present system consists of five interconnected computers and has been designated as the Advanced Control Computer System. The five computers are:  ;

4 (i) Rack-Mount 80386: This is an IBM-AT computer that is used for data acquisition, execution of software essential to safety such as the code to implement the requirements of MITR Technical Specification #6.4, and the writing of data to disk. Software on this computeris normally' invariant.

(ii) MicroVAX-II: This machine is dedicated for intensive floating-point computations such as are required to implement the various control concepts. This machine receives validated sensor information from the IBM-AT and returns the demanded actuator signal to that computer.

Software changes on this computer are expected to be frequent.

(iii) IBM-Comoatible 80386: This is a high-speed machine on which programs I are first edited, compiled, and finally linked to form an executable module. l This machine is capable of supporting automated reasomng using PROLOG, LISP, or C.

(iv) IBM-XT 8088: This computer's role is to receive validated signals from the data acquisition computer and to display model-based predictive information or a safety parameter display on its screen.

(v) LST-11/23: This unit was the original MITR digital control computer. It is now connected to the MicroVAX-II for the purpose of providing an i independent machine on which a model of the reactor can be run. This l improves simulation studies because signals must be passed between two l computers as is done for actual implementations.  !

Both the MITR Staff and the MIT Reactor Safeguards Committee concluded that this l upgraded digital control system was within the envelope of conditions prescribed in the 1985 license amendment issued by NRC for digital control experiments at MIT and that no unreviewed safety question was involved. As part of the installation of this new system, several preoperational test packages were prepared and performed. Included were tests to l verify signal transmission, to compare software performance on both the original and i' upgraded systems, ar.d to test all software and hardware interlocks.

In addition to this upgraded hardware, an auto-ranging digital picoammeter has i been installed to measure reactor neutronic power. This instrument provides both the level and range of the power signal. Moreover, it switches scales automatically and thus facilitates the development of control strategies for automated startups in which operation  ;

over many decades of power is required. This instrument was subjected to a preoperational test in which its accuracy was verified.

No new experimental research on the closed-loop digital control of nuclear reactors ,

was conducted in FY 94 because of the demands placed on the reactor for steady-state operation by other experiments. Open-loop experiments were continued as part of a program to demonstrate the practicality of flux synthesis methods for the estimation of reactivity. These will continue in FY 95.

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l-i 1 Revision of Fission Density Limit

! SR#0-8812 (12/01/88)

) The fission density limit for the UAlx fuel used by the MIT Research Reactor is

1.8 1021 fissions /cc. Research conducted by the Idaho National Engineering Lcboratory l (Nucl. Tech.. 49,136-149, June 1980) shows that a limit of 2.3 1021 fissiocs/cc is

< technically justified. Analysis of the MITR fuel cycle showed that increasing the MITR i fission density limit to 2.3 1021 fissions /cc would eventually reduce the overall number of i elements in the cycle. Accordingly, a safety analysis was prepared and, following r.: view and approval by the MIT Reactor Safeguards Committee, submitted to the U.S. Nuclear j Regulatory Commission (NRC) on 13 February 1989. On 27 November 1989, NRC j requested additional information. That material was forwarded on 6 July 1990. On 14 j January 1991, the NRC requested further additional information. The original MIT amendment request was withdrawn on 24 August 1994.

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l' Irradiation-Assisted Stress Corrosion Crackine (IASCC) Exoeriment SR#0-89-15 (06/19/89), #0-90-21 (10/22/90), #0-90-22 (10/22/90), #0-90-23 (11/G190),

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

. #0-93-6 (05/26/93), #M-93-1 (05/24/93), #0-93-14 (09/09/93), #0-94-4 (01/28/94),

1. 0-94-7 (06/13/94).

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

A facility to study 1ASCC 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 specimen's electrochemical corrosion potential (ECP) under test, and for initial analysis, the sensitivity of the specimen's ECP to varying water 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 designed to minimize radiation exposure of personnel during facility operation and transfer to a hot cell for specimen removal and replacement.

Initial in-core 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 daring in-cure 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 installed 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.

Several safety reviews were completed in FY 94 that were specific to the IASCC.

These were (1) a procedure for the pre-test treatment of specimens that had been previously irradiated in the MITR as part of the preconditioning process, (2) the safety evaluation report for the IASCC facility and its operation, (3) the abnormal operating procedures for the IASCC, and (4) procedures for the insertion / removal of the facility. None involved an unreviewed safety question.

. - _ . _ _ - . _ . - . _ _ = . . . .. .

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Adjustment of Facility Operating License No. R-37 SR#0-94 2 (03/31/94)

Facility Operating License No. R-37 expires at midnight 7 May 1996. It has been requested that this license be adjusted so that it expires at midnight 24 April 2001 in order to recover time during which the reactor was under construction, shutdown for modification, engaged in low-power testing that was a prerequisite for operation at the authorized power level, or engaged in core modification found to be necessary as a result of the low-power testing. The basis of this request for the adjustment of Facility Operating License No. R 37 is that MIT will not have received the full benefit of that license if it is allowed to expire at midnight 7 May 1996. This failure to achieve full benefit was not and is not within MIT's control. Rather it is the result of certain practices that are inherent in the licensing process. These include (1) the retroactive dating of an operating license to the date ofissuance of the associated construction permit, (2) the counting of time expended on startup testing against the operating license even though the satisfactory complenon of that startup testing is a prerequisite to routine operation at the authorized power level, (3) the counung of time expended on fuel and/or core modifications against the operating license even though those modifications were identified during the startup testing as essential to the achievement of operation at the authorized power level, and (4) the counting of time spent ,

on facility modification against the operating license even though the facility could not be legally operating during that interval (i.e., a construction permit had been issued to authorize the modification and this superseded the operating license until such time as the latter was amended).

The rationale for this request is that MIT is capable of continuing to operate the reactor facility in a manner such that the public health and safety are not endangered.

Safety Review #0-94-2 was submitted in its entirety to the U.S. Nuclear Regulatory i Commission on 31 March 1994. l l

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! l 1 F. ENVIRONMENTAL SURVEYS i

j Environmental monitoring is performed using continuous radiation monitors and l dosimetry devices. The radiation monitoring system consists of G-M detectors and 4 associated electronics at each remote site with data transmitted continuously to the Reactor  :

Radiation Protection Office and recorded on strip chart recorders. The remote sites are l located within a quarter mile radius of the facility. The detectable radiadon levels per sector j due primarily to Ar-41 are presented below, l

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l Siig Exposure (07/0lN3-06/30M4) 1 North 0.313 mrem l

East 1.313 mrem l

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' South 0.018 mrem i

j West 0.569 mrem I

Green (east) BKG i

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j Fiscal Year Averancs i i 1994 0.4 mrem  ;

1993 0.5 mrem l 1992 0.2 mrem  ;

l i 1991 0.1 mrem i

i 1990 0.1 mrem ,

i 1 l 1989 0.2 mrem i

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G. RADIATION EXPOSURES AND SURVEYS WITHIN THE FACILITY

$ A summary of radiation exposures received by facility personnel and experimenters is given below:

July 1.1993 - June 30.1994 s

Whole Body Exoosure Rance (rems) Number of Personnel 1

No measurable ................................................................149 l M eas urable - < 0.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 i 0.1 - 0.25 9 j ... ...... .. . .. ........... ... ..................................

i 0.25 - 0.5 ...............................................................16 4

O.5 - 0.75 ................. .................................................. 2 0.75 - 1.0 ........... ............. ...... ................................... I 1

1 Total Person Rem = 10.73 Total Number of Personnel = 208

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! From July 1,1993 through June 30,1994, the Reactor Radiation Protection Office 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 include, 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.

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2. Collection and analysis of water samples taken from the secondary, D20,

primary, shield coolant, liquid waste, and experimental systems, and fuel storage pool.

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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, etc.

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

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( l H. RADIOACTIVE EFFLUENTS 1

( l 1 This section summarizes the nature and amount of liquid, gaseous, and solid i radioactive wastes released or discharged from the facility. j i

. 1. Liould Waste 1

i Liquid radioactive wastes generated at the facility are discharged only to the sanitary j sewer serving the facility. There were two sources of such wastes during the year: the J cooling tower blowdown and the liquid waste storage tanks. All of the liquid volumes are j measured, by far the largest being the 7,937,900 liters discharged during FY 94 from the i cooling towers. (Larger quantities of non-radioactive waste water are discharged to the j sanitary sewer system by other parts of MIT, but no credit for such dilution is taken i because the volume is not routinely measured.)

! Total activity less tritium in the liquid effluents (cooling tower blowdown, waste j storage tank discharges, and engineering lab sink discharges) amounted to 0.000098 Ci for 1 j FY 94. The total tritium was 0.0254 Ci. The total effluent water volume was '

j 1.85x107 liters, giving an average tritium concentration of 1.37x10-6 Ci/ml.

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

) Table H-3.

I All releases were in accordance with Technical Specification 3.8-1, including j Part 20, Title 20, Code of Federal Regulations. All activitics were substantially below the j limits specified in 10 CFR 20.2003. Nevertheless, the monthly tritium releases are i reported in Table H-3. (Technical Specification 3.8-1 requires that tritium concentration be i tallied ifit exceeds 3x10-6 Ci/ml. The concentration does not exceed this figure, but the .'

data is provided anyway.)

s j 2. Gaseous Waste i

f Gaseous radioactivity is discharged to the atmosphere from the containment

! building exhaust stack. All gaseous releases likewise were in accordance with the j Technical Specifications and Part 20, and all nuclides were below the limits of 10 CFR 20.1302 after the authorized dilution factor of 3000. Also, all were substantially below the j limits of 10 CFR 20, with the exception of Ar-41, which is reported in the following Table i H-1. The 275.79 Ci of Ar-41 for 1 July 93 to 31 December 93 and the 398.72 Ci of Ar-41 j for 1 January 94 to 30 June 94 were released at average concentrations of 0.163x10 8 j pCi/ml and 0.217x10-8 pCi/ml respectively. These represent 4.1% of MPC (4x10-8

pCi/ml) for the first half of the year and 21.7% of EC (Effluent Concentration (1x10-8 pCi/ml)) for the second half. The total of these figures is 674.5 Ci which is below the l previous year's release of 923.3 Ci.

I 1 3. Solid Waste i

Only one shipment of solid waste was made during the year,information on which is provided in the following Table H-2.

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. . TABLE H Part A l

! ARGON-41 STACK RFI EASES 1 Julv 93 -31 December 93 i

Ar-41 Average Discharged Concentration (0 (Curies) ( Ci/ml)

July 1993 3.76 1.34 E-10 August 6.91 1.% E-10 September - 44.26 1.57 E-9 October 86.32 3.07 E-9 November 56.22 2.00 E-9 December 78.32 2.78 E-9 Totals (6 Months) 275.79 1.63 E-9 MPC (Table II, Column I) 4 x 10-8 l

% MPC 4,1%

TABLE H Part B ARGON-41 STACK RELEASES 1 January 94 -- 30 June 94 Ar-41 Average l Discharged Concentration 0)

(Curies) ( Ci/ml)

January 1994 41.81 1.49 E-9 l

February 88.58 3.15 E-9 Mamh 115.96 3.30 E-9 l April 51.15 1.82 E-9 May 49.87 1.77 E-9 June 51.35 1.46 E-9 Totals (6 Months) 398.72 2.17 x 10-9 EC (Table II, Colunm !) 1 x 10-8

% EC 21.7 %

(Dafter authorized dilution factor (3000). (Mole. Average concentrations do not vary linearly with curies discharged because of differing monthly dilution volumes.)

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.. TABLE H-2

SUMMARY

OF MITR-II RADIOACTIVE SOI TD WASTE SHIPMENTS FISCAL YEAR 1994 Description Volume 457.5 ft3 Weight 16,003.5 lbs Activity (1) 0.925 Ci Date of shipment May 11,1994 Disposition to licensee foc burial Barnwell Waste Management Facility Waste broker U.S. Ecology, Inc.

(1) Radioactive waste includes dry acdve waste comprised of irradiated items and/or contaminated items. The principal radionuclides are acdvation and fission products such as 60Co, 51Cr, 65Zn,125Sb,187W, 95Zr, 95Nb, 3H, 46Sc,103Ru,137Cs,55pe, 1291, 99Tc, 90S r,14C, i 10m A g, 54M n,182Ta,144Ce, and 141Ce.

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. . TABLE H-3 LIOUID EFFLUENT DISCHARGES FISCAL YEAR 1994 Total Total Volume Average Activity Tritium ofEffluent Tritium 12ss Tritium Activity Watedl) Concentration (x10-6 Ci) (x10-3 Ci) (x104 liters) (x104 pCi/ml) j l

July 1993 NDA 2.70 E-3 2.69 0.10 1

Aug. 1.65 4.18 E-7 0.02 0.002 Sept. NDA 0.143 43.9 0.33 Oct. 21.4 0.965 83.2 1.16 Nov. 13.0 0.305 55.8 0.55 J Dec. 2.48 1.140 75.9 1.50 l

Jan.1994 NDA 5.46 E-2 42.6 0.13  !

Feb. 11.70 2.736 80.0 3.42 Mar. 20.30 0.485 104.0 0.47 i

Apr. NDA 0.215 65.5 0.33 l May 14.20 0.881 63.4 1.39 June 13.00 18.53 1230.0 1.51 12 months 97.73 25.46 1847.01 1.38 (

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(1) 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.

(2) No Detectable Activity;less than 1.26x104 Ci/ml beta for each sample.

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SUMMARY

OF USE OF MEDICAL FACILITY FOR HUMAN THERAPY The use of the medical therapy facility for human therapy is summarized here l pursuant to Technical Specification No. 7.13.5(i):

1. Inwstigative Studies None were performed during this fiscal year. A Phase I study to investigate the

toxicity (or lack thereof) of neutron capture therapy is required by the U.S. Food and Drug i Administration and it was scheduled to begin in October 1993 with three patients being involved per quarter. However, this study has not yet been initiated. Efforts are in progress to expand the size of the available patient pool and thereby move ahead with the

Phase I trials. As of this writing, the initiation of the Phase I trials is tentatively scheduled i for early September 1994.

i j 2. Human Therapy a

1 None.

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