ML20056G801
| ML20056G801 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 06/30/1993 |
| From: | Bernard J, Lau E, Newton T MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9309070184 | |
| Download: ML20056G801 (32) | |
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7 NUCLEAR REACTf0R LABORATORY -
,_l AN INTERDE PARTMENTAL CENTER OF MASSACHUSETTS INSTITUTE OF TECHNOLOGY O. K. HARLING 138 Albany Street. Cambndge, Mass. 02139-4296 J. A. BERNARD, JR. !
Director Telef ax No. (617) 253-7300 Dicector of Reactor Operanons Te'ex No. 92-1473-MIT-CAM
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Tel. No. (617) 253-4211/4202 {
August 30,1993 j l
1 U.S. Nuclear Regulatory Commission, j Washington, D.C. 20555 ,
ATIN: Document Control Desk i
Subject:
Annual Report, Docket No. 50-20, License R-37, ,
Technical Specification 7.13.5 !
Gentlemen: f i
Forwarded herewith is the Annual Report for the MIT Research Reactor for the ,
period July 1,1992 to June 30, 1993, in compliance with paragraph 7.13.5 of the ;
, Technical Specifications for Facility Operating License R-37. !
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Sincerely, / /- .
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Edward S. u, NE omas H. Newton, Jr., PE Asst. Superintendent for Engineering Asst. Superintendent for Operations MIT Research Reactor MIT Research Reactor-t f .:
ohn ~ A. Bernard, Ph.D. :
Director of Reactor Operations -!
MIT Research Reactor !
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Enclosure:
As stated f i
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cc: USNRC - Project Manager, !
NRR/ONDD USNRC - Region I - Chief, )
i Effluents Radiation Protection Section (ERPS)
FRSSB/DRSS 030071 /
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, s MIT RESEARCH REACTOR '
NUCLEAR REACTOR LABORATORY MASSACHUSETTS INSTITUTE OF TECHNOLOGY :
ANNUAL REPORT l to i
i United States Nuclear Regulatory Commission !
f or the Period July 1,1992 - June 30,1993 1 i
by l i
i REACTOR STAFF ,
August 30,1993 !
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- Table of Contents Section Eage f
Tabl e of Content s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i !
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In trodu ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 3 :
A. Summary of Operating Experience ........ .............................. ......... ,
B. Reactor Operation . . . . . . . . . . . . . . . . . . . .... .... . ...................... .... ..... 9 l C. S hutdowns and S crams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . 10 D. Major Maintenance ................................................................12 ;
E. S ection 50.59 Changes, Tests, and Experiments . .... . . . . . . . .. .. .. .. . . . . . . .. . . . .. 15 l F. Emironmental S urveys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 G. Radiation Exposures and Surveys Within the Facility . .. . . .. . . . . . . . . . . 24 !
H. Radioactive Effluents .. . . . . . ... .. . .. . . ... . . . . .. . . . . . . . . . 25 e t
- 1. Summary of Use of the Medical Facility for Human Therapy ... ... .......... 29 ;
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i MIT RESEARCH REACTOR ANNUAL REPORT TO U. S. NUCLEAR REGULATORY COMMISSION i t
FOR THE PERIOD JULY 1.1992 - JUNE 30.1993 Introduction ,
t This repon 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 i Technical Specifications to Facility Operating License No. R-37 (Docket No. 50-20),
Paragraph 7.13.5, which requires an annual repon following the 30th of June of each year.
He 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 ,
in a four-foot diameter core tank, surrounded by a graphite reflector. After initial criticality on July 21, 1958, the first year was devoted to startup experiments, calibration, and a l 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. l 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 ,
surmunded laterally and at the bottom by a heavy water reflector. '. 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 origir.al facility, e.g. graphite reflector, biological and thermal ;
shields, secondary cooling systems, containment, etc., has been retained.
After Construction Permit No. CPRR-ll8 was issued by the former U.S. Atomic Energy Comtnission 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. ,
ne 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-Il on August 14,1975, and several months of stanup testing power was raised to 2.5-MW in December. Routine 5-MW operation was achieved in December 1976.
This is the eighteenth annual report required by the Technical Specifications, and it covers the period July 1,1992 through June 30,1993. Previous reports, along with the "MITR-Il Stanup Repon" (Repon No. MITNE-198, February 14, 1977) have covered the
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startup testing period and the transition to routine rertor operation. This report covers the -I sixteenth full year of routine reactor operation at the 5-MW licensed power level. It was another yearin which the safety and reliability of reactor operation met the requirements of reactor users. I 1
A summary of operating experience and other activities and related statistical data are provided in Sections A-H of this report. j I
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SUMMARY
OF OPERATING EXPERIENCE :
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- 1. General The MIT Research Reactor, MITR-II, has in recent 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. When operating, the reactor is normally at a nominal 5-MW. However, as was the case for the last three years, substantial departures were made from this schedule during the period covered by this repon (July 1,1992 - June 30,1993). Specifically, for several months, the reactor was '
run at full power almost mntinuously (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 suppon a major experimental program concerning the development of methods to reduce the activation and transpon of corrosion products in pressurized water mactor coolant. The period covered by this report was tne sixteenth full year of normal operation for MITR-II.
De mactor averaged 61.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per week at full power compared to 69. I hours per week for the previous year and 61.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> per week two years ago. As was the case in FY92 a lot of operation was conducted at low power in order to make measurements of the ;
medical therapy room beam. These measurements are for the purpose of designing an ;
epithermal neutron beam for the treatment of brain cancer (glioblastoma multiforme) and 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 i Monday afternoon until late Friday afternoon, with maintenance scheduled for Monday ;
mornings and, as necessary, for Saturdays.
The reactor was operated throughout the year with 24 elements in the core. He )
remaming three positions were used as follows: position A-1 was occupied, during the 1 carly part of FY 93 by the Boiling Water Reactor (BWR) Coolant Chemistry Imop (BCCL) ;
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experiment thimble T4 and in the latter part of FY 93 by the Pressurized Water Reactor (PWR) Coolant Chemistry Loop (PCCL) experiment thimble B4 These two loops i
reproduce chemistry conditions in power reactors and are part of a major effort to identify methods for reducing radiation exposures in the nuclear industry. The other two core positions, A-3 and B-4, were occupied by solid aluminum dummies. Compensation for .
reactivity lost due to burnup was provided by five 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 ponion of the i core (the C-Ring). In addition, elements were inverted and rotated so as to achieve more ;
uniform burnup gradients in those elements. Five other refuelings were performed for the !
i purpose of making accurate reactivity measurements and trial fits of the various Coolant Chemistry Loop experimental facilities.
He MITR-Il fuel management program remains quite successful. All of the f original MITR-II elements (445 grams U-235) have been permanently discharged. The i average overall burnup for the discharged elements was 427c. (Note One element was removed prematurely because of excess outgassing.) The maximum overall burnup achieved was 487c. Sixty-eight of the newer, higher loaded elements (506 grams U-235) have been introduced to the core. Of these, twenty-five have attained the maximum l 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 six have, as reported previously to the U.S. Nuclear Regulatory Commission, been identified as i showing excess outgassing and have been removed from service. As for the other thirty- :
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seven higher loaded elements, they are either currently in the reactor core or have been pantally depleted and are awaiting reuse in the C-ring. ,
Protective system surveillance tests are conducted on Friday evenings after shutdown (about 1800), on Mondays, and on Saturdays as necessary.
As in previous 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 i removed in November 1976 in order to gain the reactivity necessary to suppon more in-core facilities. !
- 2. Emeriments The MITR-Il was used throughout the year for experiments and irradiations in ;
suppon of research and training programs at MIT and elsewhere. j Experiments and irradiations of the following types were conducted:
a) Prompt gauuna activation analysis for the determination of boron-10 concentration l in blood and tirsue. 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 measurements to determine the suitability of various materials to serve i as a neutron filter in a medical therapy beam. These measurements are used to benchmark theoretical predictions.
c) Studies of the material composition of superconducting phases of various alloys were performed by activating samples and then identifying characteristic radiations. .
d) Irradiation of archaeological, environmental, engineering materials, biological, i geological, oceanographic, and medical specimens for neutron activation analysis ;
purposes.
e) Production of gold-198 and dysprosium-165 for medical research, diagnostic, and '
therapeutic purposes.
f) Inadiation 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 determine quantities and distribution of fissile i 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. l k) Use of trace analysis techniques to identify and monitor sources of acid deposition j (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 j chopper in a radial beam pon (4DH1). Measurements of the neutron wavelength by Bragg reflection then permits demonstration of the DeBroglie relationship for physics courses at MIT and other universities.
n) Gamma irradiation of seeds for demonstration of radiation damage effects for high school students.
o) Experimental evaluation of flux synthesis methods as a means of estimating reactivity.
p) Replication of the radiation environment in space for the study of possible methods of low temperature annealing of electronic devices that would be used in spacecraft. [
q) Neutron activation analysis of serum samples in an effort to correlate mineral deficiencies with certain diseases.
r) Activation of superconducting material to extract the vibrational amplitude of the copper atoms. Measurements of the angular yield of positrons emined by Cu-64 in a ceramic superconductor, Y i Ba2C u307, provide the vibrational amplitude of the copper atoms. The temperature dependence of the angular yield of these positrons provides information relative to the mechanism behind superconductivity for these ,
. materials. l t
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 pmgress for '
several years.) These studies entail installing loops in the reactor core to investigate the chemistry of cormsion and the transport of radioactive crud. Imops 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. In addition, an experiment involving irradiation-assisted stress corrosion cracking is planned. >
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 effon with the Tufts New England Medical Center. Patient trials are scheduled to begin later this year. j
- 3. Chances to Facility Design Except for minor changcs reponed in Section E, no changes in the facility design ,
were made during 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 j 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. J (Please see Reportable (kcurrence Reports Nos. 50-20/79-4, 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 7'70 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
t some degree. Twenty-seven with about 40% burnup have been discharged because they 1 have attained the fission density limit. Additional elements are now being fabricated by !
Babcock & Wilcox, Navy Nuclear Fuel Division, Lynchburg, Virginia. Twenty-eight of i these have been received at MIT, of which twenty-seven am m use. ;
De MITR staff has been following with interest the work of the Reduced l Enrichment for Research and Test Reactors (RERTR) Program at Argonne National ;
' 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 physics of the MrTR-Il core design show that conversion of MITR-II fuel to lower ennchment must await the ,
successful demonstration of the proposed advanced fuels.
- 4. Chances in Performance Characteristics j Performance characteristics of the MITR-II were reported in the "MrTR-ll Startup Report." Minor changes have been described in previous reports. There were no changes during the past year.
- 5. Chances in Oneratine Procedures Related to Safety -
l One amendment to the facility operating license was issued during the past year. It addresses the use of the MITR-lf s Medical Herapy Facility Beam for human therapy.
Also, an amendment conceming a revision to the fission density limit for the reactor fuel has been requested. It is discussed in Section E of this report.
With respect to operating procedures subject only to MITR internal review and i 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 repon. {i a) The procedures for performance of the cooling tower drain / refill and chemical j cleaning of heat exchangers were updated. (SR#0-92-10) l b) A procedure was developed to permit the in-air measurement of dose rate from spent fuel elements. Previously, all such measurements had been done under water. The new procedure makes it easier to demonstrate the self-protection feature of the spent fuel. (SR#0-92-13 and #0-92-18) c) The Administrative Procedures, Chapter One of the Procedure Manual, were revised to update the lists of names and committee memberships. (SR#0-92-19) d) The procedure for verification of the operability of all fan interlocks was revised to j incorporate suggested improvements. (SR#0-92-20) !
I e) Procedures for the shipment of spent MITR fuel were developed and issued.
(S R#0-93-2) f) ne core purge systemis designed to isolate upon receipt of a high radiation signal.
This feature makes it difficuh to determine if a series of trips are the result of high ,
radiation or an electronic problem. An analysis has been proposed and approved to permit operation without the trip feature. (SR#0-93-4) 4 g) The procedure for performance of the efficiency check of the charcoal pressure relief fihers for the removal of iodine was updated. (SR#0-93-7) ;
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h) ne procedure for the verification of control blade speed was updated. l (SR#0-93-8) i i
- 6. Surveillance Tests and Insnections There are many written procedures in use for surveillance tests and inspections required by the Technical Specifications. These procedures provide a detailed method for '
conducting each test or inspection and specify an acceptance criterion which must be met in order for the equipment or system to comply with the requirements of the Technical Specifications. The tests and inspections are scheduled throughout the year with a
- 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 stanup of the reactor if shutdown ['
for more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, before stanup if a channel has been repaired or de-energized, and at least monthly; a few are on different schedules. Procedures for such surveillance are 3 incorporated into daily or weekly startup, shutdown or other checklists.
During the reporting period, the surveillance frequency has been at least equal to that required by the Technical Specifications, and the results of tests and mspecnons were ;
satisfactory throughout the year for Facility Operating License No. R-37. ;
- 7. Status of Soent Fuel Shinment Pursuant to Amendment No. 25 to Facility Operating License No. R-37, paragraph i 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 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-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 ;
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) De Certificate of Compliance and the Safety Analysis Report of ae 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) he 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-Il spent fuel shipment was prepared and approved under the MITR safety review program. His Q/A program was approved by NRC on July 23,1991. ;
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8-(d) The decay heat load of each spent element was determined by a member of the MITR staff and found to be within :he 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 performed using the Monte-Carlo Code KENO-V which was ob:ained 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 specification.
(c) 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.
(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.
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 in accordance with our procedure for shipment preparation. A special structure for support of the BMI-l cask was designed and fabricated.
(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 crideality 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.
(1) Procedures 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. 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. Depanment 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. :
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B. REACTOR OPERATION f Information on energy generated and on reactor operating ho its is tabulated below: f I
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1 2 3 4 Total ,
- 1. Energy Generated (hnVD):
a) MITR-II(MITFY93) 114.8 107.2 173.6 267.2 662.8 (normally at 4.9 M\V) b) MITR-II 12,029.6 (HUT FY76-92) c) MITR-1 10,435.2 (MIT FY59-74) d) Cumulative, 23,127.6 MITR-I & MITR-II
- 2. MITR-Il Operation (Hrs): i (MIT FY93) a) At Power
(>0.5-MW) for 688.1 655.9 979.8 1,381.0 3,704.8 Research ,
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. (<0.5-AnV) for 91.8 47.2 42.8 38.9 220.7 i Training 0) and Test ,
c) TotalCritical 779.9 703.1 1,022.6 1,419.9 3,925.5 1
- 0) 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. i i
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C. SHUTDOWNS AND SCRAMS During the period of this report there were 6 inadvertent scrams and 14 unscheduled power reductions. ,
'Ihe 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 suberitical by the ;
reactor operator in response to an abnonnal condition indication. Rod drops and electric l power loss without protective system action are included in shutdowns.
The following summary of scrams and shutdowns is provided in approximately the same format as for previous years in order to facilitate a comparison.
Nuclear Safety System Scrams :
- 1. Total a) Channel #6 trip as result of personnel ermr in that setpoint was too low. 1 b) Withdraw permit circuit malfunction as a result of ;
electrical noise induced by a high-speed drill. I c) Withdraw permit circuit malfunction as a result of :
electrical noise induced by the self-checking of a UPS. 3 Subtotal 5 -
II. Process System Scrams a) MTS-1 high temperature trip as result of personnel error in that setpoint was too low. 1 >
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1 III. Uncebeduled Shutdowns or Power Reductions a) Shield systemleak alarms. 1
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b) Shutdown as result of PCCL or BCCL experiment malfunction. 6 i c) Shutdown due to loss of offsite electricity. 5 i
d) Blade #2 drop as result of incorrect setting of magnet i current. 1 c) Shutdown to investigate abnormal levels of cesium. 1 l
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l Total 20 l Experience during recent years has been as follows for scrams and unscheduled :
4 shutdowns:
Fiscal Year Number ,
Scrams Shutdowns ' Total 89 10 8 18 :
90 11 9 20 j 91 11 9 20 92 5 12 17 l 93 6 14 20 ,
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D. MAJOR MAINTENANCE Major maintenance projects performed during FY93, including the effect, if any, on the safe and reliable operation of the MIT Research Reactor are described in this section.
One of the major FY93 maintenance items was the transfer of forty-eight spent MITR-II spent fuel elements in six shipments to the DOE facility at Savannah River, SC.
New procedures were developed and completed for the comprehensive annual inspection of the BMI-l fuel shipping cask, the refurbishment of the BMI-1 cask suppon skid, and the loading / shipping of the spent fuel elements. All six shipments were completed safely and in a timely manner.
Storage capacity for spent fuel was increased from 50 to 75 elements by constructing a third spent-fuel stomge rack. The neutron absorption capability of each cadmium-lined unit in this storage rack was verified by a neutron transmission test -
following a visual check for quality. The storage rack was then positioned in the spent fuel storage pool. Cdticality analyses had previously shown that the creation of this new storage space would be within the limit (Keg 50.9) specified in the MITR Technical Specification.
In addition to shipment of spent fuel elements, a total of 88 ft3 of spent (tritiated) heavy water, around 6000 lbs., was shipped to the DOE facility at Savannah River, SC, in ,
thirteen 55-gallon stainless steel drums. He tritiated heavy water was unloaded from the reactor heavy water reflector system in FY92, at which time fresh (unitradiated) heavy water was introduced to the system. He shipment was completed in a safe and timely ,
manner. Dat provides an additional four inches of venical clearance.
Much maintenance was performed to suppon 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 July 23, 1992 and the run was successfully completed on October 2,1992. The PCCL (B4 thimble) was installed on January 19,1993. It was operated at full power almost cetinuously for more than 22,741 '
MWH and the run was successfully completed on June 10, 1993. Much preventive maintenance was performed to suppon this continuous operation. ,
Another maintenance item was to suppon the development of a new Irradiation Assisted Stress Corrosion Cracking (IASCC) experiment. The IASCC thimble and its I-I dummy were test-fitted in the reactor core twice and the resulting design of a new core thimble support bridge was completed. This IASCC experiment requires the use of a new reactor top shield lid that provides an additional four inches of vertical clearance. The new lid consists of a stainless-steel shell that is filled with lead. Prior to its installation, the lid's smface was painted with primer and epoxy paint, and its side and bottom were spray- ,
coated with clear acrylic paint. He new finish both protects the surface of the new lid and :
facilitates decontamination, if necessary. Four tie-down mechanisms were fabricated and ,
welded on to the upper shield ring to secure the new lid. The new lid provides additional shielding atlocations where radiation streaming was higher with the original lid. Gamma radiation dose rates at the center area of the lid were reduced to an average of 3.5 mR/hr versus 15 mR/hr for the original lid.
In addition to supporting the in-core loop experiments, Reactor Floor Hot Cell #2 was refurbished and its two CRL Model #8 Master-Slave manipulators were repaired. The
refurbishment, which improved user safety, included comprehensive decontamination of the cellinterior, installation of a large shatter- and scratch-resistant mirror (stainless steel),
installation of improved lighting, re-installation of improved fire protection equipment, i rewiring of the mterior electrical cable for fire and electrical protection, application of a new ,
internal wall lining for contamination reduction, and installation of external shielding to reduce radiation streaming. The neighboring ~ Reactor Floor Hot Cell #1's external radiation ,
shielding was also improved to accommodate high level radiation work being performed by the 1ASCC project group inside Hot Cell #2.
Another maintenance activity in support of the water chemistry loop (CCL) reseamh was the fabrication of a 15-ft. long water-tight cylindrical container which was then submerged and secured in the spent fuel storage pool. This container provided temporary storage of irradiated CCL equipment that had been stored in five separate canisters. He container was filled with DI water to provide radiation shielding. The five canisters were ;
then lowered into the container in an orderly manner. A water recirculation system equipped with a Barnstead HN ultrapure mixed bed deionizer was built to control the water a chemistry in the container.
Another area of major maintenance activity was the shield coolant system. A water leak was identified upon inspecting the two independent cooling coil sets in the lower annular ring with an optical fiber micro-probe. De leaky coil was then isolated. ,
Measurements were then performed to verify that, as was predicted theoretically, the ;
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remaining coil was adequate to absorb the redistributed and added heat load. Upon resumption of reactor operation, the temperature change in the surrounding concrete of the !
lower annular ring was closely monitored. No significant rise in temperature was observed.
Other major preventive maintenance items perfonned in FY93 were as follows:
(i) The two main pumps of the secondary coolant system were rebuilt, their i ball bearings repacked, and shaft seals replaced. The result has been quiet, ,
reliable operation of these pumps.
(ii) The gasket on the outer door of the main personnel lock was replaced. The ,
new gasket is made of improved braided mbber material and it is more wear-resistant than the earlier gaskets.
(iii) All ventilation system filters were replaced for the chemistry hoods and -
clean-room hoods in the Back Engineering Lab, the Hot Shop (radiortive material machine shop), the Operations Counting room, and the Isotope Preparation room. The air flows for these ventilation facilities were also re-adjusted.
The Cooling Tower vibration switches were cleaned, adjusted, and !
(iv) lubricated. ,
(v) The containment building crane was inspected twice during FY93. One t inspection was performed . in-house early in the fiscal ' year and a ,
comprehensive inspection was performed by Nonheast Electric Company l later. De latter inspection included a magna-flux check on all crane hook cables for signs of cracks and fatigue. Both crane inspections certified the safe and reliable operating condition of the crane.
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(vi) He CO gasholder 2 cable that holds the counterweight was replaced with an improved cable.
(vii) Control blade drive motors were adjusted and lubricated.
(viii) All airflow sampling lines of the exhaust ventilation system were chemically Dushed and cleaned. Air sampling flowrates of 30 cfm were restored from pre-Dush Dowrates of 15 cfm.
(ix) ne Secondary Chemistry hot cell manipulator, used daily in handling of radioactive materials, was improved by replacing the cable and readjusting the cable's tension.
(x) The Reactor Top floor surface was refinished to improve contammation control.
(xi) The intake ventilation damper actuator was adjusted and replaced.
(xii) The Hot Shop (radioactive material machine shop) was decontaminated. All machine tools and drill platforms were tuned, adjusted, and lubricated. The concrete floor was sealed with three layers of epoxy coating for contamination control. Lighting in the room was refurbished.
Many other routine maintenance and preventive maintenance items were performed throughout the fiscal year.
E. SECTION 50.59 CH 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.
He review and approval of changes in the facility and in the procedunes 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 funher 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. ;
he 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."
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i Pressurized Coolant Chemistrv Imop (PCCL)
SR #0-86-9 (04/21/88), #0-38-4 (07/28/88), #0-88-5 (09/09/88), #0-88-14 (12/07/88), ;
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- 0-89 2 (01/06/89), #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 (03/2090), #0-90-7 (03/20S 0), #0-90-8 (03/20h0), #0-90-9 (03/20S0), #0-90-25 (12/10B0), #0-90-26 (12/1890), #0-90-27 (12/1860), #0-91-8 (05/2191), #0-91-21 (12/27S 1), #0-92-2 (01/27B2), #0-92-12 (08/19S2).
This project involves the design, insallation, 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). 'Ite 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 determination, 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. These 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.
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i Boi[ine Coolant Chemistrv Loon (BCCL)
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), ,
- 0-90-20 (10/15/90),#0-91-20 (01/30/92),#0-92-11 (08/15/92),#0-92-16 (09/25/92). j 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 exam' .ed 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 pmposed to demonstrate that boiling within an in-core facility would not adversely affect reactor operation. Following both a determination that ;
no unreviewed safety question was involved and approval by the MIT Reactor Safeguards Committee, this experiment was conducted. The results were as expected i
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 detemiined 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 l 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:
- Preoperational 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 BCCL began in October 1990 and have been spectacularly successful in that many theories concerning the transport of nitrogen-16 in boiling water reactors have been disproven. During the period covered by this repon, two changes were made to the BCCL experimental protocol. These were:
- Issuance of a second list of chemicals approved as additives. Each of these was to ,
be studied for its effect on the suppression of nitrogen-16 carryover. -
- Approval to operate the BCCL in a recirculating mode. Previously, it had only operated in aonce-through' manner. ;
Experiments that make use of the BCCL facility were conducted during portions of ,
this reporting period. !
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E3periments Related to Neutron Capture Therapy 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),
- 0-92-3 (03/06/92), #0-92-4 (03/02/92), #M-92-2 (05/14/92), #0-93-5 (05/28/93), :
- 0-93-9 (07/13/93).
In conjunction with the Tufts - New England Medical Center and with the support ,
of the U.S. Department of Energy, MIT is designing an epithermal neutron beam for the i treatment of brain cancer (glioblastoma). 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 patierit to surgery involving removal of a portion of the skull. Also, an epithermal beam gives greater penetration. In October 1991, MIT bosted an international workshop for the purpose of reviewing proposed beam designs and dosimetry. Subsequent to the receipt of advice from the workshop panel members, a final design was selected for the epithermal filter for the MIT Research Reactor's Medical Therapy Facility beam. That design, which was one of many that had been previously constructed and evaluated, is No. M-57. It has now been installed permanently.
Approvals of the protocol for the conduct of patient trials were received from all requisite MIT and NEMC Committees as well as from the U.S. Food and Drug Administration. ;
Changes that occurred during the period covered by this report include:
- A license amendment and quality management plan for use of the MIT Research i Reactor's Medical Therapy Facility was issued by the U.S. Nuclear Regulatory Commission.
- Procedures implementing the license amendment and its associated quality :
management program were prepared and submitted as an item of information to the ,
U.S. Nuclear Regulatory Commission.
- A preoperational test program was prepared and issued. t Patient trials are now scheduled to begin later this year.
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Digital Computer Control of Reactors Under Steadv-State and Transient Conditions 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 !
(02/08/83), #0-83-12 (04/23/83),#0-83-20 (07/20/83),#0-84-11 (06/25/84),#0-84-12 (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 (12/3 /88),#0-90-28 (12/27/90),#0-91-2 (05/14/91),#0-91-3 (06/06/91),#0-91-10 (07/15/91),#0-91-11 (06/25/91),#0-92-6 (06/09/92),#0-93-3 (03/29/93). )
The project involving computer analysis, signal validation of data from reactor 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 validation procedures, ensures that there will not be any challenge to the reactor safety system while testing closed-loop control methods. Several such methods, including decision analysis, rule-based control, and modern control theory, continue to be experimentally evaluated. The eventual goal of this program is to use fault-tolerant t 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 review procedures (Safety Review numbers listed abeve) and approved as necessary by the MIT Reactor Safeguards Committee. ,
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 (0.27o AK/K). !
Following the successful completion of these tests, facility operating license amendment ?
No. 24 was obtained from NRC (April 2,1985). It permits:
(1) Closed-loop control of one or more shim blades and/or the regulating rod ,
provided that no more than 1.87c AK/K could be inserted were all the connected control elements to be withdrawn, and ,
(2) Closed-loop control of one of the shim blades and/or the regulating rod provided that the overall controller is designed so that reactivity is constrained sufficiently to permit control of reactor power within desired or 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 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 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 >
condition. This arrangement for the conduct of these tests has been approved by the MIT Reactor Safeguards Committee.
I An extensive upgrade to the digital control system's hardware was performed m 1991. The present system consists of five interconnected computers and has been designated as the Advanced Control Computer System. The five computers are, i
(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 i implement the requirements of MITR Technical SpeciGeation #6.4, and the '
writing of data to disk. Software on this computer is 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 are first edited, compiled, and finally linked to form an executable module.
This machine is capable of supporting automated reasoning using PROLOG, LISP, or C.
(iv) IBM-XT 8088: This computer's role is to weeive validated signals from the ;
data acqaisition computer and to display model based predictive information or a safety parameter display on its screen. .
(v) LSI-11/23: This unit was the original MITR digital control computer. It is ;
now connected to the MicroVAX-Il for the purpose of providing an ,
independent machine on which a model of the reactor can be run. This ,
improves simulation studies because signals must be passed between two computers as is done for actual implementations.
Both the MITR Staff and the MIT Reactor Safeguards Committee concluded that this 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, j several preoperational test packages were prepared and performed. Included were tests to verify signal transmission, to compare software performance on both the original and upgraded systems, and to test all software and hardware interlocks. ,
~3 In addition to this upgraded hardware, an auto-ranging digital picoammeter has !
been installed to measure reactor neutronic power. This instrument provides both the level and range of the power signal. Moreover, S switches scales automatically and thus i 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. ;
i No new experimental research on the closed-loop digital control of nuclear reactors -l was conducted in FY 93 because of the demands placed on the reactor for steady-state :
operation by other experiments. Open-loop experiments were conducted as part of a l program to demonstrate the practicality of flux synthesis methods for the estimation of :
reactivity. These will continue in FY 94, 1
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Revision of Fission Density Limit SR #0-88-12 (12/01/88)
The fission density limit for the UAl xfuel used by the MIT Research Reactor is 1.8 1021 fissions /cc. Research conducted by the Idaho National Engineering Laboratory i (Nucl. Tech., 49,136-149, June 1980) shows that a limit of 2.3 1021 fissions /cc is technically justified. Analysis of the MITR fuel cycle showed that increasing the MITR fission density limit to 2.3 1021 fissions /cc would eventually reduce the overall number of elements in the cycle. Accordingly, a safety analysis was prepared and, following review and approval by the MIT Reactor Safeguards Committee, submitted to the U.S. Nuclear Regulatory Commission (NRC) on 13 February 1989. On 27 November 1989, NRC ,
requested additionalinformation. That material was forwarded on 6 July 1990. On 14 January 1991, the NRC requested further additional information. This material has been '
prepared and will be submitted to NRC shonly.
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' Irradiation-Assisted Stress Corrosion Crackine (IASCC) Emeriment ~
SR #0-89-15 (06/19/89), #0-90-21 (10/22/90), #0-90-22 (l0/22/90), #0-90-23 (11/05/90),
- 0-91-21 (12/27t?l), #0-92-5 (04/02/92), #0-92-17 (09/28/92), #0-93-6 (05/26/93),
M-93-1 (05/24/93).
This project involves the design, installation, and operation of an in-core facility for the purpose of studying irradiation-assisted stress corrosion cracking and in particular the combined effect of neutron and gamma radiation on this process. In addition to studying the mechanism of this process, steels of varying compositions will be tested with the objective of identifying materials that are resistant to this mode of failure.
I The IASCC is cunently under design. When finished, it will provide the capability for placing steel specimens under a constant, measurable strain rate on samples that are contained in an in-core facility. Review of the experiment design by both the MITR Staff :
and the MIT Reactor Safeguards Committee is scheduled to occur in November-December 1993. The steel specimens that will be used for the 1ASCC tests have been irradiated as part of a preconditioning process. These irradiations are done subject to strict controls on both temperature and atmosphere. The facility used for this purpose was first designed in 1979 and has been used previously for sample preconditiom,ng.
Two safety reviews were completed during FY 93 that were specific to the IASCC.
These involved transfer of the irradiated specimens to one of the hot cells located on the reactor floor and to a laboratory work area. Neither involved an unreviewed safety question. A comprehensive safety analysis (SR#0-92-21) has been prepared and is undergoing review.
As part of the preparations for this experiment, a new reactor top lid has been designed. This lid, which provides an additional four inches of venical clearance for in- :
core experiments, meets or exceeds the specifications for the original lid. Radiation levels i directly above the reactor have been reduced as a result of the installation of the new lid.
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F. ENVIRONMENTAL SURVEYS 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 the Reactor Radiation Protection Office and recorded on strip chart recorders. The remote sites are ;
located within a quarter mile radius of the facility. The detectable radiation levels per sector due primarily to Ar-41 are presented below.
i Silq Exnosure (07/01/92-06/30/93) !
North 0.306 mR East 1.524 mR South 0.034 mR West 0.748 mR Green (east) 0.017 mR Fiscal Year Averages -
1993 0.5 mR !
1992 0.2 mR f 1991 0.1 mR i
1990 0.1 mR 1989 0.2 mR 1988 0.2 mR i l
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i G. RADIATION EXPOSURES AND SURVEYS Wil'HIN THE FACILITY A summary of radiation exposures received by facility personnel and experimenters l is given below:
l July 1.1992 - June 30.1993 Whole Body Exoosure Rance (Rems) Number of Personnel j No measurable ...............................................................158 -l M eas urabl e - < 0.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 1 0.1 - 0.25 ............................................................14 0.25 - 0.5 ....................... ....... ..................................... 6 0.5 - 0.75 ............... ..................................................... 6 0.75 - 1.0 ... ........... ......................................... ......... . 1 Total Person Rem = 10.16 Total Number of Personnel = 210 ,
From July 1,1992 through June 30, 1993, 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 -l frequencies as required), maintenance activities, and experimental project suppon. Specific examples of these activities include, but an: not limited to, the following: ,
- 1. Collection and analysis of air samples taken within the containment building and in the exhaust / ventilation systems.
- 2. Collection and analysis of water samples taken from the secondary, D20, primary, shield coolant, liquid waste, and experimental systems, and fuel storage pool. 1,
- 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 l' protection / survey instrumentation, and establishing / posting radiological control areas.
- 4. Provision of radiation protection services during fuel movements,in core experiments, sample irradiations, beam pon 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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H. RADIOACTIVE EFFLUENTS This section summarizes the nature and amount of liquid, gaseous, and solid radioactive wastes released or discharged from the facility.
- 1. Liould Waste Liquid radioactive wastes generated at the facility are discharged only to the sanitary sewer serving the facility. There were two sources of such wastes during the year: the cooling tower blowdown and the liquid waste storage tanks. All of the liquid volumes are measured, by far the largest being the 6,007,000 liters discharged during FY93 from the ,
cooling towers. (Larger quantities of non-radioactive waste water are discharged to the t 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 0.00028 Ci for ,
FY93. The total tritium was 0.0075 Ci. The total effluent water volume was :
6.05x106 liters, giving an average tritium concentration of 1.24x 10-6 pCi/ml. -[
The above liquid waste discharges are provided on a monthly basis in the following Table H-3.
All releases were in accordance with Technical Specification 3.8-1, including Part 20, Title 20, Code of Federal Regulations. All activities were substantially below the limits specified in 10 CFR 20.303, but the monthly tritium releases are reported in :
Table H-3 in accordance with Technical Specification 3.8-1 because its concentration exceeded 3x10-6 pCi/ml.
- 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 Part 20, and all nuclides were below the limits of 10 CFR '
20.106 after the authorized dilution factor of 3000. Also, all were substantially below the limits of 10 CFR 20, Appendix B, Note 5, with the exception of Ar-41, which is reported ;
in the following Table H-1. The 923.3 Ci of Ar-41 were released at an average ;
concentration of 0.24x10-8 pCi/ml for the year. This represents 6.0% of MPC (4x10-8 pCi/ml) and is consistent with the previous year's release of 728.1 Ci.
- 3. Solid Waste !
Only one shipment of solid waste was made during the year,information on which t is provided in the following Table H-2. i g , _ _ , . - - - _ - . - ,.
k TABLE H-1 ARGON-41 STACK RELEASES FISCAL YEAR 1993 ;
Aral Average '
Discharged ConcentrationW (Curies, (pCi/ml)
July 1992 24.3 0.07 x 10-8 August 70.6 0.24 September 64.4 0.18 October 51.6 0.18 l
November 34.8 0.12 December 41.5 0.11 January 1993 41.9 0.14 February 95.0 0.33 Mamh 98.5 0.34 April 195.8 0.54 May 136.6 0.47 June 68.3 0.19 Totals (12 Months) 923.3 0.24 x 10-8 MPC (Table II, Column I) 4 x 10-8 4
% MPC 6.0%
0)After authorized dilution factor (3000). (Note: Average concentrations do not vary l linearly with curies discharged because of differing monthly dilution volumes.)
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SUMMARY
OF MITR-II RADIOACTIVE SOLID WASTE SHIPMENTS FISCAL YEAR 1993 Description Units Shipment #1 Total P
Volume Cubic Feet 210 210 Weight Pounds 5266 5266 Activity 0) Curies 0.218 0.218 Date of shipment December 2,1992 l Disposition to licensee for burial U.S. Ecology, Inc. l
- 0) Radioactive waste includes dry active waste comprised of irradiated items and/or contaminated items. The principal radionuclides are activation and fission products such es 60Co, 51 Cr, 65Zn,125Sb,187W, 95Zr, 95Nb, 3H, 463e,103Ru,137Cs, 55Fe, 129J,99Tc,90Sr,14C,110 mag, 54Mn,182Ta,144Ce,and 141Ce. ,
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TABLE H-3 LIOUID EFFLUENT DISCHARGES FISCAL YEAR 1943 Total Total Volume Average Activity Tritium ofEffluent Tritium less Tritium Activity Water (I) Concentration (x10-6 Ci) (x10-3 Ci) (x104 liters) (x104 pCi/ml)
July 1992 NDA 0.05 15.2 0.34 Aug. 108 0.54 63.9 0.85 Sept. NDA 0.12 46.2 0.25 Oct. 17 9 0.68 55.2 1.24 Nov. 9.1 0.49 34.8 1.42 Dec. NDA 0.32 37.2 0.87 Jan.1993 NDA 0.07 29.6 0.24 Feb. 7.0 0.47 60.6 0.78 Mar. 0.2 0.19 45.7 0.41 Apr. NDA 0.35 80.5 0.43 May 5.2 2.19 71.0 3.08 June 134 1.99 63.8 3.12 12 months 281.4 7.46 603.7 1.24
- 0) Volume of ef0uent 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 1.26x10-6 pCi/ml beta for each sample.
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SUMMARY
OF USE OF MEDICAL FACILITY FOR HUM AN THERAPY The use of the medical therapy facility for human therapy is summarized here pursuant to Technical Specification No. 7.13.5(i): l l
1 Investicative Studies i None were performed during this fiscal year. A ' Phase One' study to investigate ;
the toxicity (or lack thereof) of neutron capture therapy is required by the U.S. Food and ;
Drug Administration and it is cheduled to begin in October 1993. It is anticipated that three !
patients will be involved per quarter. ;
- 2. Human Therany None. ,
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