ML20114D926

Mit Research Reactor Nuclear Reactor Lab Mit,Annual Rept to NRC for 910701-920630

ML20114D926
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	06/30/1992
From:	Bernard J, Kwok K MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9209100173
Download: ML20114D926 (31)

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M eb NUCLEAR REACTOR LABORATORY D AN INTERDEPARTMENTAL CENTER OF MASSACHUSLTTS INSTITUTE OF TECHNOLOGY k '

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- O K HARL'NG 138 Albany Street. Cambridge, Mass,02139 -JA. BERN ARD, JR-l_

Daector Telefax No- (617)253-7300 ' Director of Reactor Operations

, Telen No. 92 1473 MiT C AM Tel. No (617) 253 4211/4202 i

j August 30,1992 l U.S. Nuclear Regulatory Commission, l Washington, D.C. 20555 l- A'ITN: Document Control Desk i

Subject:

- Annual Report, Docket No. 50-20, License R-37,

. Technical Specification 7.13.5

! Gentlemen:-

i Forwarded herewith is the Annual Report for the MIT Research Reactor for the period July 1,1991 to June 30, 1992, in compliance with paragraph 7.13.5 of the l Technical Specifications for Facility Operating License R-37.

i Sincerely, J

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j Kwan S. Kwok, Ph.D.

Superintendent -

MIT Research reactor

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!' ohn A.' Bernard, Ph.D.

l Director of Reactor Operations -

4 MIT Research Reactor; l JAB /KSK:gw

Enclosure:

As stated l

h cc: USNRC - Project Manager, .

. NRR/PDNP t.

USNRC - Region I - Chief, i

' Effluents Radiation Protection Section (ERPS) .

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. 1 MIT RESEARCH REACTOR NUCLEAR REACTOR LABORATORY '

MASSACMUSETFS INSTITUTE OF TECHNOLOGY ANNUAL REPORT to United States Nuclear -Regulatory. Commission f or the Period July 1,1991 - June 30,1992.

r by REACTOR STAFF l

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Table of ConteD11 Section Eagt Table of Contents .............. .. ... ........... .. . ........ .... ................ ...... i ,

Intmduction .. . .. ... . ...... . .. .. . ...... .. .. . .... ................. ...... 1 A. Summarf of Operating Experience ............ .................................... 3 B. Reactor Operation ............ ..... ... . ......... ................... ... . . ..... 9 C. Shutdowns and Scrams .........,.................................................10 D. Major Maintenance ...........................................................12 E. Section 50.59 Changes, Tests, and Experiments . ... . .....................14 F. Environmental S urveys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 G. Radiation Exposures and Surveys Within the Facility .... ...................... 24 H. Radioactive Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 1

I

-1 MIT RESEARCil REAC'IOR ANNUAL REPORT TO U. S. NUCLEAR REGULATORY COMMISSION FOR TIIE PERIOD JULY 1.1991 - JUNE 30.1992 Introduction This report has been prepared by the staff of the Massachusetts Institute of Technology Research Reactor for submission to the Administrator of Region 1. United States Nuclear Regulatory Commission, in compliance with the requirements of the Technical Specifications to Facility Operating License No. R-37 (Docket No. 50-20),

Paragraph 7.13.5, which requires an annual report following the 30th of June of each year.

The MIT Research Reactor (MITR), as originally constructed, consisted of a core of MTR type fuel, 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 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 increased to two megawatts in 1962 and to five megawatts (the design power level) in 1965.

Studies of an improved design were first undertaken in 1967. ne 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 ,he 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-il8 was issued by the former U.S. Atomic Energy Commission in April 1973, major components for the modified reactor were procured and the MlTR-1 was shut down on May 24, 1974, having logged 250,445 megawatt hours during nearly 16 years of operation.

He old core tank, associated piping, top shielding, control rods and drives, and some experimental facilities were disassembled, runoved, 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 startup testing, power was raised to 2.5-MW in December. Routine -

5-MW operation was achieved in December 1976.

This is the seventeenth annual report required by the Technical Specifications, and it l covers the period July 1,1991 through June 30,1992. Previous reports, along with the I "MITR-Il Startup Report" (Report No. MITNE-198, February 14,1977) have covered the l

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. 2-i startup testing period and the transition to routine rextor operation. 'Ihis report covers the 4 fifteenth full year of routine reactor operation at the 5-MW licensed power level. It was ,

another year in which the safety and reliability of reactor operation nrt the requirenrnts of reactor users.

4 A summary of operating experience and other activities and rel;ted statistical data

are provided in Sections A-li of this report.

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SUMMARY

OF OPERATING EXPERIENCE

1. General

' he MIT Research Reactor, MITR-II, has traditionally been operated on a routine.

l four days per week schedule, modified as necessary to facilitate the preoperational testing and installation of severalin core experiments. When operating, the reactor is nomully at a j nominal 5 MW. Ilowever, as was the case for the last two years, substantial departures were made from this schedule during the period covered by this report (July 1,1991 - June 30, 1992), Specifically, a five-day per week operating schedule was followed during much of the past twelve months but for several months the reacte 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 experimental program concerning the development of methods to reduce the acuvation and transport of corrosion products in pressurized water reactor coolant. The period covered by this report was the fifteenth full year of normal operation for MITR-II.

He reactor averaged 69.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per week at full power compared to 61.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> per week for the previous year and 48.1 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 FY91 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 moltiforme) and possibly skin cancer (melanoma). When neither the corrosion reduction experiments nor the medical beam design was in prognss, the reactor was usually operated from late

, 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

remaining positions were occupied by an irradiation facility used for the coolant chemistry loop which is designed to reproduce conditions in power reactors, by a facility for irradiating metal specimens to be used later for a study on irradiation-assisted stress corrosion cracking, and by a solid aluminum dununy. Compensation for reactivity lost due 1

to burnup was provided by four refuelings. These followed standard MITR practice which is to introduce fresh fuel to the inner portion of the core (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 as to achieve more unifomi burnup

(- gradients in those elements. Eleven other refuelings were performed for the ptprpose of making accurate reactivity measurements and runs of the various Coolant Chemistry Loop experimental facilities.

! ne MITR-II fuel management program remains quite successful. All of the original MITR-II clements (445 grams U-235) have been permanently discharged. The average overall bumup for the discharged elements was 42% (Note: One element was i

removed prematurely because of excess outgassing.) ne maximum overall burnup achieved was 48E Fifty-nine of the newer, higher loaded elements (506 grams U-235) have been introduced to the core. Of these, twenty-two have attained the max, imum allowed fission density. However, 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 showing excess outgassing and have been removed from service. As for the other thirty-one higher loaded elements, they are either currently in the reactor core or have been partially depleted and are awaiting reuse in the C-ring.

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1 i 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

, thernul 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 support more in-core facilities.

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,' 2. Eweriments t

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

Experiments and irradiations of the following types were conducted:

a) Prompt 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.

He analysis is to support our neutron capture therapy program.

b) Experimental measurements to detennine the suitability of various materiais to serve as a neutron filter in a medical therapy beam. These measurementr are used to benchmark theoretical predictions.

c) Studies of the material composition of superconducting pnases of various alloys -

were performed by activating samples and then identifying characteristic radiations.

d) Irradiation of archaeological, environmental, engineering materials, biological, geological, oceanographic, and medical specimens for neutron activation analysis purposes, e) Production of gold-198, dysprosium-165, and holmium-166 for medical research, diagnostic, and therapeutic purposes, f) Irradiation of tissue specimens on particle track detectors for plutonium radiobiology.

g) Irradiation of semi-conductors to determine resistance to high doses of fast neutrons, h) Use of the facility for reactor operator training.

i) Irradiation of geological materials to determine quantities and distribution of fissile materials using solid state nuclear track detectors.

j) Evaluation of various chemical additives for the suppression of nitrogen-16 activity 1 in a boiling water reactor environment. l l

k) Experimental studies of various closed-loop control techniques with emphasis on j methods for trajectory tracking.

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

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5-m) Measurements of the energy spectrum of leakage neutrons using a mechanical i chopper in a radial beam port (4DHl). 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) Preparatory work for the 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.

4 q) Neutron activation analysis of serum samples in an effort to correlate mincral deficiencies with certain diseases.

, r) Activation of superconducting material to extract the vibrational am31itude of the copper atoms. Measurements of the angular yield of positrons emittec by Cu 64 in a ceramic superconductor, Y iBa2Cu3 0,7 provide the vibrational amplitude of the copper atoms. He temperature dependence of the angular yield of these positrons i provides information relative to the mechanism behind superconductivity for these materials.

Dose reduction studies for the light water reacto. 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 cote to investigate the chemistry of corrosion and the transport of radioactive crud. Loops that replicate bo*h pressurized and boiling water reactors have been built. The PWR loop has been operational rince 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 us. of the reactor is a program to design a facility for the treatment of glioblastouns (brain tumors) and melanomas (skin cancer) using neutron capture therapy.

His is a collaborative effott with the Tufts New England Medical Center. Patient trials are scheduled to begin late in 1992 or early in 1993.-

3. Changes to Facility Design

. Except for minor changes reported in Section E, no changes in the facility design were made during the year. As indicated in past reports the uranium loading of MITR-Il fuel was increased from 29.7 grams of U-235 per plate and 445 grams per element (as made by Gulf United Nuclear Fuels, Inc., New Haven, Connecticut) to a nominal 34 and 510 grams respectively (made by the Atomics International Division of Rockwell

. International,- Canoga Park, California). With the exception of seven elements (one Gulf, six AI) that were found to be outgassing excessively, performance has been good.

(Please see Reportable 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 50-20S1-1.) 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 41 of the more highly loaded elements in 1982, 40 of which have been used to some degree.

Twemy-two with about 40% burnup have been discharged because they have attained the fission density limit. Additional elements are now being fabricated by Babcock & Wilcox, I

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Navy Nuclear Fuel Division, Lynchburg, Virginia. Nineteen of these have been received at l MIT and are now in use.

The MITR staff has been following with interest the work of the Reduced Enrichment for Research and Test Reactors (RERTR) Program at Argonne National Laboratory, particularly the development of advanced fuels that will permit uranium 1 loadings up to several times the recent upper limit of 1.6 grams total uraniunVcubic i centimeter. Consideration of the themul-hydraulics and reactor physics of the MITR-Il core design show that conversion of MITR-II fuel to lower ennchment must await the successful demonstration of the proposed advanced fuels.

4. Changes in Performance Characteristics Performance characteristics of the MITR-II were reported in the "MITR-Il Startup

Report." Minor changes have been described in previcus reports. Dere wem no changes during the past year.

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5. Changes in Ooerating Procedures Related to Safety One amendment to the facility operating license was issued during the past year. It was for an extension of the temporary possession limit for U-235. Also, an amendment concerning a levisian 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 apprt cal, a sumnury 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 used for normal reactor startup and shutdown were updated to t

incorporate a number of minor changes that had been made as authorized temporary changes. His process is ongoing and other checklists are being presently updated.

(SR #0-91-4) b) A procedure for responding to a massive spill of tritiated heavy water was prepared and issued. This was done in support of the changeout of the MITR-II's heavy water reflector that was performed in December 1991. (SR #0-91-12) c) PM 3.7.3, "Nonnal Containment Building Entry / Exit" was issued. The steps in this procedure had previously been part of several checklists that concerned reactor operation. The decision to separate out the material related to buikling entry / exit was done to facilitate security. (SR #0-91-14) d) The Adnunistrative Procedures, Chapter One of the Procedum -Manual vere revised to update the lists of names and committee memberships. (SR #0-92-1)

6. Surveillance Tests and Insoections Here are many written procedures in use for surveillance tests and inspections required by the Technical Specifications. 'Ihese 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 TechnicM Specifications. He tests and inspections are scheduled throughout the year with a frec uency at least equal to that required by the Technical Specifications. Twenty-seven suc t 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 for more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, before startup if a channel has been repaired or de-energized, and at least monthly; a few are on different schedules. Procedures for such surveillance are incorporated into daily or weekly startup, shutdown or other checklists.

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

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

MIT began efforts for spent fuel shipment as early as 1983. At that time, the plan was to use two MH-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.

Relative to the capability of shipping spent fuel from the MIT Research Reactor using the BMI-l cask, the following has been accomplished:

(a) The Certificate of Compliance and the Safety Analysis Report of the BMI-1 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) ne University of Missouri Research Reactor (MURR) basket was reviewed and found to be suitable for use with the MIT fuel elements in the BMI-1 cask. MURR has agreed to make their basket available to MIT for the required shipments. .

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

(d) ne decay heat load of each spent element was determined by a member of the MrlR staff and found to be within the limits specified in the Cernficate 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 are near completion and are being performed using the Monte-Carlo Code KENO-V which was obtained from the Radiation Shielding Information Center of the Oak Ridge National Laboratory. Preliminary results show that the degree of subcriticality of a cask fully loaded with MIT fuel elements is within specification.

(e) In order to cross-check the cross sections used in the KENO-V code, criticality analyses are being performed using a second Monte-Carlo code, MCNP. Results obtained thus far are consistent with those obtained using KENO-V.

8 (f) Arrangements have been made with the fuel receiving organization at the Savannah River facility. Specific data on the MITR-Il spent fuel elements have been compiled. The Appendix A document has been prepared and is being reviewed by the spent fuel processing center.

l (g) Spent fuel elements in the MITR spent fuel storage pool were rearranged and grouped in accordance with our procedure for shipment preparation. A special structure for support of the BMI-l cask is being designed and fabricated.

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

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

(j) A criticality study of the BMI-l cask with fresh MITR fuel was completed and

! submitted to the U.S. Department of Energy. Approval of the study is still pending.

(k) Funding was allocated by the U.S. Department of Energy for the return to a DOE facility of spent MITR fuel. He funding is for government fiscal year 1993.

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

Quarter 1 2 3 4 Total

1. Energy Generated (MWD):

a) MITR-II (MIT FY92) 209.1 238.0 179.4 121.6 748.1 (nornully at 4.9 MW) b) h TR-II - 11,281.5 (MIT FY76-91) c) MITR-1 10*435~

(MIT FY59-74) d) Cumulative, 3.,

~~' 464.8 MITR-! & MITR-II

2. MITR-II Operation (Hrs):

(MIT FY92) a) At Power

(>0.5-MW) for 1080.4 1193.5 914.7 729.6 3,918.2 Research i ,

b) Low Power

(<0.5-MW) for 37.0 114.4 154.8 245.1 551.3

Training (l)and Test l

l c) TotalCritical 1117.4 1307.9 1069.5 974.7 4,469.5 1

0) These hours do not include reactor operator and other training conducted while the

reactor is at full mwer for research purposes (spectrometer, etc.) or for isotope i

production. Such 1ours are included in the previous line, i

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4 C. SilUTDOWNS AND SCRAMS 1

During the period of this report there were 5 inadvertent scrams and 12 unscheduled l l

l power reductions.

De term " scram" refers to shutting down of the reactor through protective system l 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 1

reactor operator in response to an abnomial condition indication. Rod drops and electric 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.

l I. Nuclear Safety System Scrams Icht

a) Channel #5 trip as result of personnel error. 1 b) Channel #3 power supply. 1 4

c) Period channel off-scale low during restart. 1 Subtotal 3 j II. Process System Scrams a) Pressure pulse on secondary system as

result of frozen cooling tower riser pipe. 1 b) Core outlet temperature trip during low i power dose measurements. 1 Subtotal 2 I

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l Ill. Unscheduled Shutdowns or Power Reductions a)- Primary system leak alann. 1 i

i b) Shutdown as result of PCCL or 13CCL experiment malfunction. 7 4

c) Shutdown due to loss of offsite electricity. 4 4-i- Subtotal :12 i

i Total 17 i

, Experience during recent years has been as follows for scrams and unscheduled shutdowns:

! Fiscal Year ~ Number l

88 21

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89 18 90 20 i 91 -20

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D. MAJOR MAIN'IFNANCE Major maintenance projects during FY92, including the effect, if any, on safe operation of the reactor are described in this section.

3 Major maintenance items were continued in FY92 to support the requirements of the dose reduenon projects for light water reactors. These projects are the Pressurized Water Reactor Coolant Chemistry loop (PCCL), the Boiling Water Reactor Coolant Chemistry

loop (BCCL), the Irradiation Assisted Stress Conosion Cracking (IASCC) experiment, and the Sensor project. Additional space on the reactor top to support both the IASCC experiment and the Sensor project was obtained by constructing a platfonn that is l supported fmm the reactor floor. The final design of this platform is in accordance with standard construction codes and was approved by a licensed structural engineer.

Construction of the new platfomt was completed in May 1992. The IASCC experiment also requires a reactor top shield lid that is significantly different trom the one currently in use. The new lid is all stainless-steel and filled with lead. It allows an additional four

inches of vertical clearance under the lid end provides additional shielding at locations where radiation streaming was higher than d sired on the lid that is currently in use.

Construction of the new lid is being done in accordance with the quality assurance requirements. He new lid will be installed for support of the IASCC experiment in FY93.

In addition to supporting these dose reduction p ojects for light water reactors, major evolutions were also performed for development or new experiments that use other z facilities. He prompt-gamma facility was moved from the six inch radiai beam port, 6SH4, to a four-inch radial beam port, 4DH3. A new water shutter was designed and installed in the 4DH3 port to facilitate sample changing and better beam control. The S spectrometer was removed from the 4DH5 port so as to provide the necessary floor space and clearance for the 6THi-2 through ports. He 6TH1-2 and 4Till 3 through ports are i undergoing conversions into major irradiation facilities. These conversions will be completed in FY93.

he gasket on the inner door of the main personnel lock failed. The airlock was tagged out and removed from service. An order for two new gaskets was alaced with the manufacturer. However, the manufacturer could only produce one acceptab e gasket. That gasket was installed and tested in accordance with the quality assurance requirements. The airlock was retumed to service following the satisfactory completion of a pressure test on the personnellock. De manufacturer of the airlock is no longer able to produce acceptable gaskets because of personnel turnover. He gasket uses a design that is unique to the-MITR and requtres custom fabrication. MIT is actively pursuing an ahernate supplier of these gaskets.

The outer door of the containment vehicle (truck) lock failed at one of the areas where the hydraulic rams make contact with the door. He face plate at the contact area was comded and brokeloose from the door flange causing insufficient pressure on the gasket 3 to e I the door. (Note: This failure occurred during the 1991 annual building pressure test. He outer door was functioning properly prior to that test.) Repairs related to this failure were completed in FY92. The truck lock remains tagged out-of-service because more deterioration and the need for funher improvement on its overall condition were identified while repairing the hydraulic rams. Additional repairs and improvement on the truck lock are scheduled to be implemented as part of the containment refurbishment as described in the following paragraphs.

Major efforts in 3 reserving and improving the overall condition of the containment building were continuec in FY92. He old cathodic protection system was identified as g - n - , , , p - - ~ , - nn-- -- ~,e,-- w -- - , . , , -, - ,- ,

,. becoming ineffective in FY91. A completely new cathodic protection system was designed by a licensed corrosion engineering firm and the new system was installed and tested in FY92. Finalevaluation of the new system is on-poing. Consideration is now being given to the installation of one to two more electrodes m order to provide additional coverage to the portion of the containment shell below grade, beneath the Utility Room where the electrical transformers and emergency power batteries are located. This effort will b continued in FY92.

Another effort in preserving and improving the cr.tainment building is the resurfacing and painting of the containment shell. R: urrent plan is to mechanically remove the old protective coatings and repair all deteriocation on the containment shell, vehicle (truck) lock, and main personnel lock. In addition, soil around the containment shell will be removed so that new bitumastic coating can be applied at below the grade level. New protective coatings will be applied to the exterior shell as soon as possible following the removal of the old protective coatings This action will mininuze any oxidation which may occur following removal of the old coatings.

Another major effort in preventive maintenance was the replacement of the D2 0 in the heavy water reflector sysWm in FY92. The original heavy water for MITR-II was put to use in 1975 and its tritium concentration had reached 3.25 Ci/l. New heavy water was procured from DOE and the entire inventory of the heavy water system was replaced. He tritium concentration in the D2 O following replacement was reduced to 40 mci,.. In addition, all replaceable components on valves and other instruments such as valve diaphragms and gaskets were replaced while the heavy water system was drained. This greatly reduced the risk of component failures in the heavy water system. Replacing the heavy water in the system also decreased the source term significantly if a pipe break were to occur in the D20 reflector system.

Another area of major maintenance was the secondary system. Major portions of piping for venting the three main heat-exchangers were replaced because of scale build up m these pipes. The air-conditioning piping that serves the control room and reactor floor air-conditioners was found to be corroded due to scale. A 350 ft section of the steel piping was replaced with corrosion resistant copper pipes and instrumentation manifolds so as to improve monitoring of the system status and overall control of the heat removal capacity.

In addition, another 600 ft. section of the secondary air-conditioning piping was found to be almost clogged. His section of piping is difficult to replace because of its location. A professional company was contracted to clean this section of pioing with a muriatic acid and corrosion inhibitor solution. He cleaning was effective anc the chill-water flow was restored to its originallevel.

A third spent-fuel storage rack was fabricated to increase the storage capacity from 50 to 75 spent elements in the spent-fuel storage pool. One of the three main heat-exchangers, HE-1 A, developed a leak. One of the 988 tubes was found to be leaking and was plugged. Shaft seals on one of the two primary pumps, MM-1, and on one of the two booster pumps, HM-2, developed leaks and were replaced. He bearmgs and shaft seal on the make-up water system pump, WM-1, were also replaced. Bearings on the intake fan motor were replaced and the CO2 gasholder cable that holds the counterweight was replaced. The motors for control blade #3 and the regulating rod were rebuilt. Most of these replacements and rebuilds were performed as preventive measures.

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

E.' SECTION 50.59 CHANGES. 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 canditions of Section 50.59 of 10 CFR 50, together with a sumnury of the safety evaluation in each case.

The review and approval of changes in the facility and in .the procedures as described in the SAR are documented in the MITR records by means of " Safety Review Forms". These have been paraphrased for this report and are identified on the following pages for ready reference if further information should be required with regara to any itent 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 Director, Standardization and Non-Power Reactor Project Directorate, Office of Nuclear Reactor Regulation, USNRC, he conduct of tests and experiments on the reactor are nomially 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 Fornt 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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Pressurized Coolant Chemistry Imp (PCCL1 SR #0-86-9 m4/21/88), #0-88-4 (07/28/88), #0-88-5 (09/09/88), #0-88-14 (12S7/88),

1. 0-89-2 (0100,89), #0-89-3 (01/19/89), #0-89-6 (01/24/89), #0 89-9 (06M2/89), #0-89-14 (06/19/89), #0-90-6 (03/20S 0), #0-90-7 (03/20B0), #0-90-8 103/2080), #0-90-9 (03/20S0), #0-90-25 (12/103 0), #0-90-26 (12/18S 0), #0-90-27 (12/18S 0), #0-91 8 (05/2161), #0-91-21 (12/27B1), #0-92 2 (01/27 S 2).

His project involves the design, installation, and operation of a pressunzed light-water loop in the MITR core for the purpose of studying the production, activation, and transpon of corrosion to detemune the optimum method for reducing the creation of activated corrosion products (crud) and thereby reducing radiation 'ields associated with pressurized water reactors (PWRs). The ultimate goal is to reduce radiation exposures to PWR maimenance personnel.

Approval for the PCCI s.a given by the MITR Staff and the MIT Reactor Safeguards Committee on (M/20/88. It was determined at that time that nc .nreviewed-safety question existed because no failure or accioent associattJ with the PCCL could lead to an accident or failure involving reactor ecmponents. Details of that detemiination, together with safety review #0-86-9, were submitted to the U.S. Nuclear Regulatory Commission on (M/21/88.

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

- Procedure for Ex-Core Testing

- Supplement to the Safety Evaluation Report

- PreoperationalTest Procedure

- Abnormal Operating Procedures for the PCCL

- Procedures for PCCL Startup/ Shutdown

. - Procedures for PCCL Installatio-/ Removal ,

Expenments using the PCCL began in Apri'i 1989 and have been quite successful.

Dunng the period covered by this report, several changes were made to the FCCL Hese were:

- A procedure was prepared for transfer of used PCCL thimbles to the spent fuel storage pool. Used thimbles are generated at a rate of about one per year. They will be kept under water for shielding purposes until shipped off-site as low level waste.

- Re electrical configuration of the PCCL's heater was modified by makmg two changes. Rese were (1) integration of the heater and power cable sheaths to eliminate potential shorting at the heater sheath seal; and (2) replacement of the insulated junction box between the two heater legs by a " grounding strap," a sheath-to-sheath electricalconnection, which is welded to the heater sheaths and to the power leads at the bottom of the heater legs. The purpose of the aforementioned two changes is to preclude moisture pickup by the MgO insulation and to allow use of the heater sheath as a return current path. This latter change will allow operation of the heater system with one of the two heaters shorted. Such operation will in

_ _ _ - - __ - - --__ -_-_-_- ~

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6 l turn reduce radiation exposures because it will not be necessary to ch:aige out the j' shoned heater. His change was also made to the heaters for other in-core loop expenments.

Neither of these changes involved an unreviewed safety question. All served to increase the reliability of the experiment. .

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! Experiments that make use of the PCCL facility continued throughout most of this l reporting period.

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Boiline,Coobnt Chemistry Loop (BCCIA SR #0-80-14 (06/19/89), #0-89 20 (12/20/89), #0-90-17 (09/17/90), #0-9018 (09/14/90),

1. 0-90-20 (10/15/90),#0-9l-20 (01/30/92)

This project involves the design, installation, and operation af 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 will be examined to detennine 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 mace a determination that boiling within an in-core facility is not contrary to the technical specifications provided that reacuvity limits for movable experiments an: not exceeded. It was also concluded that boiling in the proposed experiment volume would not significantly affect reactor operation. Accordingly, a carefully controlled experiment vias proposed 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?pproval 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:

- Preoperational Test Procedure.

- Abnormal Opemting 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 beer. tpectacularly successful in that many theories conceming the transport of nitrogen-16 in boiling water reactors have been dispreven. During the period covered by this report, one change was made to the BCCL experimental protocol. It was:

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

Expe"ments that make use of the BCCL facility were conducted during portions of this reporting period.

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18 -

Exneriments Related to Neutron Cacture Thernov ~

SR' #0-89-4 (01/23/89), #0-89-8 (03/01/89),' ~#0-91-7 (05/06/91), #0-92-4 (03/02/92),

1. M-92-2 (05/14,W)

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 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 patient to surgery in rolving removal of a portion of the s':ull. Also, an epithermal beam gives greater penetration. In October 1991, MIT hosted an international workshop for the purpose of reviewing proposed beam designs and dosimetry. Subsequent to the receipt of advice from 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. Other changes that occurred during the period covered by this rt 'x>rt include:

- A license amendment and quality management plan for use of the MIT Research Reactor's Medical Therapy Facihty was prepared and submitted to the U.S. Nuclear Regulatory Commission. (SR #0-91-17 and SR #0-92-3)

- A boron-containing rgr.y paint was developed for the purpose of preventing neutron activation of the therapy facility's walls.

- Approvals of the protocol for the initiation of patient trials were received from all requisite MIT and NEMC committees as well as from the U.S. Food and Drug Administration.

Patient trials are now scheduled for late '1992 or early 1993,

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Digital Computer Contml of Reactors Under Steady-State and Transient Conditiom 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),

1. 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.I1 (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 4 (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 (06/06/91),#0-91-10 (07/15/91),#0-91-11 (06/25/91),#0-92-6 (06/09/92)

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 computers coupled with closed-loop digital control and signal validation methods to demonstate 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 above) and approved as necessary by the MIT Reactor Safeguards Committee.

Initial tests of this digital closed-loop controller were conducted in 1983-1984 usin?

the facility's regulating rod which was of relatively low reactivity worth (0.2% AK/h).

Following the successful completion of these tests, facility operating license : mendment 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 1.8% AK/K could be inserted were all the connected control elements to be withdrawn, and (2) Closed-!cm control of one of the shim blades and/or the regulating rod provided t1at the overall controller is designed so that reactivity is constrained sufficiently to permit control of reactor power within desired or authorized lirdts.

A successful experimentation program is now continuing under the provisions of this license amendment. A protocolis 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 s,ccification #6.4 which requires that reactivity be constrained to ensure " feasibility of control." Signalimplementat on i 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.

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

(i) Rack-Mount 8038h: This is an IBM AT computer that is used for data acquisition, execution of software essential'o safety such as the code to implement the requirements of MITR Technical Specification #6.4, and the wnting 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-Compatible 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 8083: 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) LSI-l1/23: This unit was the original MITR digital control computer. It is now connected to the MicroVAX-II for the purpose of providing an independent machine on which a model of the reactor can be run. This improves f.mulation 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 mat 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 verify signal transmission, to compare software performance on both the original and upgraded systems, and to test all software r.nd hardware interlocks.

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, 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 digital control of nuclear reactors was conducted in FY 92 because of the demands placed on the reactor for steady-state operation by other experiments. It is hoped that the program will resume during the upcoming fiscal year v hen open-loop experiments are to be conducted to demonstrate the practicality of flux synthesis methods for the estimation of reactivity.

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Revision of Fission Densit';'.imit SR #0 88-12 (12/01/88)

The fission density limit for the UAl xfuel used by the hilt Research Reactor is 1.8 1021 fissions /cc. Research conducted by the ! Aho National Engineering Laboratory (Nucl. Tech.. 49,136-149, June 1980) shows that a limit of 2.3 1021 fissions /cc is technically ju,tified. Analysis of the h11TR fuel cycle showed that increasing the hilTR fission density limit to 2.3 1021 fissions/cc would eventually reduce the overall nurnber of elements in the cycle. Accordingly, a safety analysis was prepared and, following review and approval '" the hilt Reactor Safeguards Committee, submitted to the U.S. Nuclear Regulatory L.nmission (NRC) on 13 February 1989. On 27 November 1989, NRC requested additionalinformation. That material was forwarded on 6 July 19W. On 14 January 1991, the NRC requested further additional infomution. The hilTR Staffis in L e process of preparing this material for transmission to NRC.

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Irradiation Assisted Stress Corrosion Cracking GASCC) Experiment SR #0 8915 (06/19/89), #0-90-21 (10/22h0), #0-90-22 (10/22NO), #0 90-23 (11/05/90), l

1. 0-91 21 (12/27/91),#0-92 3 (04/02/92) I 1

nis pmject involves the design, installation, and operation of an in-core facility for the purpose of studying irradiation assisted stress corrosion cracking and in panicular 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 failtve.

) ne IASCC is currently under design. When finished, it will provid: the capability for placing steel specimens under a constant, measurable strain rate on samples that are contained in a 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 1 1992. The steel specimens that will be used for the IASCC tests have been inadiated as 3 art of a preconditioning process. These irradiations are donc subject to strict controis on 30th temperature and atmosphere. The facility used for this parpose was first designed in 1979 and has been used previously for sample preconditioning.

li Two safety rMews were completed during FY 92 that were specific to the IASCC.

These involved procedures for the transfer of the irradiated specimens. None involved an

unreviewed safety question.

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21 F. ENVIRONMENTAL SURVEYS Environmental monitoring is perfonned using continuous radiation monitors and dosimetry devices. The radiation monitoring system consists of G M detectors and  !

associated electronics at each remote sitc with data t ansmitted continuously to the Reactor Radiation Protection Office and recorded on strip chart recorders. The remote sites are located within a quarter taile radius of the facility. The detectable radiation levels per sector due primarily to Ar-41 are presented below.

Sitt Exoosure (07/DINI-06BOS2)

North 0.0065 mR East 0.741 mR South 0.036 mR West 0.103 mR l Green (east) 0.029 mR Fiscal Year Averages 1992 0.2 mR l')91 0.1 mR 1990 0.1 mR 1989 0.2 mR 1988 0.2 mR 1987 1.2 mR I

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i i G. RADIATION EXPOSURES AND SURVEYS WIT 111N Ti1E FACILITY a

A summary of radiation exposures received by facility personnel and experimenters is given below; i Julv 1. lo41 - June 30.1992 i

Whole 1 ody Emosure Rana (Rems) Number of Personnd i No measurable .. .................. ............................................126 a

M e a s ura b l e - < 0.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 7 0.i - 0.25 8 f ........................................ ............................

i 0.25 - 0.5 ..................................................................... 9

! 0.5 - 0.75 ....... ........... .,,.............................................. 7

0.75 - i.0 ... ................................................................. 3 1

l.0 - 1.25 ..................................................................... I 1.25 - 1.50 ..................................................................... 2-Total Man Rem = 16.37 Total Personnel = 203 From July 1,1991 through June 19.1992, the Reactor Radiation Protection Office provided radiation protection servicca 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 cooling towers, D 20 system, primary system, shield coolant system, heat exchangers, fuel storage facility, waste storage tanks and experimental systems.

3. Performance of radiation and contamination surveys, radioactive waste collection, calibration of area radiation monitors, calibration of effluent moniters, calibiation of radiation survey instruments, and establishing /

posting radiological control areas.

4. Provide radiation protection services during fuel movements, in-core experiments, sample irradiations, beam port use, ion column removal, etc.

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

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Il. RADIOACTIVE EFTLUENTS This section summarizes the nature and amount of liquid, gaseous, and solid radioactive wastes released or discharged from the facility.

1. I lauld 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 'oer ewdown and the liquid waste storage tanks. All of the liquid volumes are measured. . y w ;he largest being the 6,241,000 liters discharged during FY92 from tbr cooling t-  : Larger quantities of non radioactive waste water are discharged to the sanitary seu system by other pans of MIT, but no credit for such dilution is taken since the volume is not routinely measured.)

Total activity less tritium in the liquid efDuents (cooling tower blowdown, waste i storage tank discharges, and engineering lab sink discharges) amounted to 0.014 Ci for l FY92. The total tritium was 0.023 Ci. The total effluent water volume was 6.26x1(b liters, giving an average tritium concentration of 3.61x10 6 pCi/ml.

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

All releases were in acc ;rdance wit h Technical Specification 3.81, including Part 20, Title 20, Code of Federal Regulatiorn. All activities were substantially below the limits specified in 10 CFR 20.303, but the monthly tritium releases are reported in Table 113 in accordance with Technical Specification 3.81 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 -41, which is reported in the following Table Il 1. The 728.1 Ci of Ar 41 were meased at an average concentration of 0.20x10 8 pCi/ml for the year. This represents 4.9% of MPC (4x10 8 pCi/ml) and, given the 12% increase in reactor operating hours from FY91 to FY92, is consistent with the previous year's release of 684.4 Ci. This reflects the continued success of our efforts to identify and eliminate sources of Ar-41.

3. Solid Waste Only one shipment of solid waste was made during the year, information on which is provided in the following Table 112.

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, 26-TABLE 11-1 ARGON-41 STACK Rf! LEASES lilSCAL YEAR 1992 Ar-41 Average Discharged Concentrationm (Curies) (pCi/ml)

July 1991 42.2 0.14 x 10 8 August 58.4 0.16 September 66.3 0.23 October 85.4 0.23 November 97.8 0.33 December 63.6 0.22 January 1992 20,0 0.06 February 79.8 0.27 March 71.5 0.24 April 33.9 0.12 May 56.7 0.16 June 52.5 0.18 Totals (12 Months) 728.1 0.20 x 10 8 MPC (Table 11, Column 1) 4 x 10-8

% MPC 4.9%

0)After authorized dilution factor (3000). (N_n11: Average concentrations do not vary linearly with curies discharged because of differing monthly dilution volumes.)

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TABI.F 112 .

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SUMMARY

OF MITR-Il RADIOACrlVE SOI.lD WASTE SillPMENTS  !

FISCAL YEAR 1942

)

Description Units Shipment #1 Total l Volume Cuble Feet 127.5 127.5 Weight Pounds 3791 3791 Activity 0) Curies 0,0114 0.0114 Date of shipment > ,vember 1,1991 Disposition to licensee for burial U.S. Ecology, Inc.

0) Radioactive waste includes dry active waste comprised of irradiated items and/or contaminated items. The principal radionuclides are activation and fission products such as 60Co, slCr,65Zn,125Sb,187W,95Zr,95Nb, 311,46Sc,103Ru,137Cs,55Fe, 129), 997c, 90S r, I d C, 1 10m A g, 54 M n, 182Ta, 14dCe, and 141Ce.

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!. 28 TABLE 11-3 LIOUID EFFLUENT DISCllARGES FISCAL YEAR IW2 Total Total Volume Avenige Activity Tritium of Effluent Tritium less Tritium Activity Watedi) Concentration (x104 Ci) (x10 3 Ci) (x1(S liters) (x104 pCi/ml)

July 1991 4060 1.87 U.4 3.5 Aug. 5840 1.96 69.7 2.8 Sept. 189 6.16 73.6 8.4 Oct. NDA(2) 0.41 79.8 0.5 Nov. 21 4.75 78.4 6.1 Dec. NDA 1.51 41.0 3.7 Jan.1992 NDA(2) 0.01 10.4 1.0 Feb. 12 0.89 54.1 1.6 Mar. NDA(2) 0.07 69.8 0.1 .

Apr. 3600 4.77 21.7 22.0 May NDA(2) 0.05 35.2 0.1 June . NDA(2) 0.14 39.2 0.4 12 months 13,722 22.6 626.3 3.61

0) Volume of effluent from cooling towers, waste tanks, and NW12139 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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