ML20082L845

Mit Research Reactor Annual Rept to NRC for Jul 1990 to June 1991

ML20082L845
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	06/30/1991
From:	Bernard J, Kwok K MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9109040401
Download: ML20082L845 (32)

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MASSACHUSETTS INSTITUTE OF TECHNOLOGY O K HARLING - 138 A!baay Shmt Cambncae. Mass 02133 J A BE RNARD JH Diractar Teief as N3 (617 253 7330 D 'edm v Recto Opwadons Teles Na 92 1473 MiT CAM Te: No Mn 253-4211/4202 August 30,1991 U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 ATTN: Document Control Desk

Subject:

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

Forwarded herewith is the Annual Report for the MIT Research Reactor for the period July 1,1990 to June 30, 1991, in compliance with paragraph 7.13.5 of the Technical Specifications for Facility Operating U. cense R-37.

Sincerely, dL.- A Kwan S. Kwok, Ph.D.

Superintendent MIT Research reactor

.k St.

John A. Bernard, Ph. .

Director of Reactor Operations MIT Research Reactor JAB /KSK:gw

Enclosure:

As stated cc: USNRC - Region I - Olief, Reactor Projects Section No. 3A USNRC - Region I - Reactor Engineer, Reactor Projects Section No. 3A USNRC - Senior Resident Inspector, Pilgrim Nuclear Station USNRC - Project Manager, Standardization and Non-Power Reactor P.oject Directorate USNRC - Region I - Chief, Eftluents Radiation Protection Section 9109040401 910630 nn

(. POR R

ADOCK 05000020 PDR 7/

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MIT RESEARCH REACTOR NUCLEAR REACTOR LABORATORY MASSACHUSETTS INSTITUTE OF TECHNOLOGY ANNUAL REPORT to I

l United States Nuclear- Regulatory Commission- 1 for the Period July 1,1990 - June 30,1991 H by L:

l REACTOR STAFF .

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August 30,1991 l

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l l Table of Contents s

a Section M Tabl e of Con tents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i in trod u c tio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ........ . 1 A. Summary of Operating Experience ...................... .......................... 3 B. Reactor Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 C. Shutdowns and Scrams .................... ............. .....................10 D. Major Maintenance ..............................................................12 E. S ection 50.59 Changes, Tests, and Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 F. Environmental S urveys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 G. Radiation Exposures and S urv eys Within the Facility . . .. . . . . . . . . .. . . . . . . . . . . . 25 H. Radioactive Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6

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MIT RESEARCH REACTOR ANNUAL REPORT TO U.S. NUCLEAR REGULATORY COMMISSION EOR THE PERIOD JULY 1 1990 - JUNE 30. lo91 Introduttien 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 cornpliance with the recuirements 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 er 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 imtial 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-Fnday) commenced in July 1959. The authorized power level was increased tc, two megawatts in 1962 and to five megawatts (the design power level) in 1965.

Studies of an improved design were first undertaken in 1967. He concept which was finally adopted consisted of a more compact core, cooled by light water, and surrounded laterally and at the bottom by a heavy water reflector, it is undermoderated for the purpose of maximizing the peak of thennal neutrons in the heavy y ' ster 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. He com is hexagonal m shape,15 inches across, and utilizes fuel elements which are rhomboidal in cross section and which contain UALx intermetallic 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 themul shields, secondary cooling systems, containment, etc., has been retained.

After Construction Permit No. CPRR-118 was issued by the former U.S. Atomic Energy Commission in April 1973, major components for the modified reactor were procured and the MITR 1 was shut down on May 24, 1974, having logged 250,445 megawatt hours during nearly 16 years of operation.

The old core tank, associated piping, top shielding, control rods and drives, and some expenmental facilities were disassembled, removed, and subsec uently replaced with new equipment. After preoperational tests were conducted on a,1 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-ll 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 sixteenth annual report required by the Technical Specifications, and it covers the period July 1,1990 through June 30,1991. Previous reports, along with the

2-a "MITR il fi v9 Repon"(R:pon No MITNE-1981:ebruary 14 1977) have covered the startup testb. s%1 and the transition to routine reactor op: ration. This irpon covers the founeenth fuit year of routine reactor operation at the 5 MW licensed power level. It was anotner year in which the safety and reliability of tractor operation met the mluirements of reactor users.

A sumnuuy of operating experience and other activities ano related statistical data ur povided in Sections A Il of tlas repon.

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A.

SUMMARY

OF OPERATING EXPERIENCE

1. O s ml The MIT Research Reactor, MITR ll, has traditionally been operated on a routine, four days per week schedule, modified as necessary to facilitate the preoperational testing "

and installation of several in core experiments. When operating, the reactor is nonnally at a nominal 5 MW. Ilowever, as was the case last year, substantial departures were made from this schedule during the period covered by this report (July 1,1990 June 30,1991).

Specifically, a five-day per week operating schedule was followed during much of the past twelve months and for several months the reactor was run at full power almost continuousiv (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 pmgram conceming the development of methods to raluce the activation and transport of corrosion products in pressurized water reactor coolant The period covered by thls report was the fourteenth fuJ l year of normal operation for MITR il.

The reactor averaged 61.9 houn per week at full power compared to 48,1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per week for the previous year and 40.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per week two years ago. As was the case in FY90 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 multifonne) and possibly skin cencer (melanma). When neither the conosion reduction experiments nor the medical beam design was in propen, the reactor was usually operated from late Monday afternoon unti; late Friday attemoon, with maintenance scheduled for Monday mornings ar d, as necessary, for Saturdays.

The reactor was operated throughout the year with either 24 or 25 clements in the core. The 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 alumintun dununy. Compensation for reactivity lost due to burnup was pmvided by five refuelings. nese followed standard MITR practice which is to introduce fresh fuel te the inner portion of the core (the A and B Rings) wh:re peaking is least and and to place panially spent fuelin the outer portion of the core (the C-Ring). In addition, elements were inverted and rotated so as to achieve more uniform burnup gradients in those elements. Twenty two other refuelings were perfomied for the purpose of making accurate reactivity measurements and runs of the various Coolant Chemistry Loop experimental fwilities.-

The MITR Il fuel management yogram remains quite successful. All of the original MITR ll elements (445 grams b235) have been ynnanently discharged. The average overall burnu for the discharged elements was 42% (Hom: One element was removed prematurel because of excess outgassing.) The maximum overall burnup achieved was 48%. ity one of the newer, h gher loaded elements (506 grams U 235) have been introduced to the core. Of them, nineteen have attained the maximum allowed fission density. liowever, these may be reused if that limit is increased as would seem warranted based on metallurgical studies by DOE. Another five 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 twenty seven 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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4 Protective system surveillance tests are conducted on Friday evenings after shutdown (about IMO), 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 reDector beneath the core. These had been removed in November 1976 in order to gain the reactivity necessary to support more in-core facilities.

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

Experiments and irradiations of the following types were conducted:

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

The analysis is to support our neutmn capture therapy pmgram.

b) Experimental neasurements to determine the suitability of various nuterials to sen>e as a neutron filter in a medical therapy beam. Rese measurernents are used to benchmark theoretical predictions.

c) Studies of the material composition of supreonducting phases of various alloys were performed by activating samples and tien 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 registance to high doses of fast neutron',.

h) Use of the facility for reactor operator training.

1) Irradiation of geokmical materials to determine quantities and distribution of fissile materials using solk state nuclear track detectors, j) Closed-loop direct digital control of reactor power using a shim blade as well as the regulating red dtuing some steady state and transient conditions. .

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

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

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m) hicasurements of the energy spectrum of leakage neutrons using t mechanical chopper in a mdial beam pon (4 Dill). Measurements of the neutron wavelength by Dragg reflection then permits demonstration of the Dellroglie relationship for l physics courses at MIT and other universities. l n) Gamma irradiation of seeds for demonstration of radiation damage effects for high school students, o) Study of radiation damage effects on magnets that are intended for use in the superrollider.

p) Re slication of the radiation environment in space for the study of possible methods !

of ow temperature annealing of electronic devices that would te used in spacecraft. i q) Neutron activation analysis of serum samples in an effon to correlate mineral deficiencies with cenain diseases.

Dose reduction studies for the light water reactor industry began reactor use on a regular basis in 1989. (Planning and out of core evaluatiens had been in progress for several years.) Rese studies entall installing loops in the reactor core to investigate the

{ chemistry of corrosion and the transpon of radioactive crud. loops that replicate both pressurized and boiling water reactors have been built. The PWR loop has been operational and in-core since August 1989. The BWR loop became operational in October 1990, in addition, an experiment involving irmdiation 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.

His is a collaborative effon with the Tufts New England Medical Center. Patient trials are scheduled to begin late in 1991.

3. Changes to Facility Design Exec pt for minor changes reported in Section E, no changes in the facility design were made d uring the year. As indicated in past reports the uranium loading of MITR-il fuel was increased from 29.7 grams of U 235 xr plate and 445 grams per element (as made by Gulf United Nuclear Fuels, Inc., New alaven, Connecticut) to a nominal 34 and 510 grams respectively (made by the Atomics International Division of Rockwell Intemational, Cano ga Park, Califomia). With the exception of six elements (one Gulf, five AI) that were fourg to be outgaasing excessively, performance has been good. (Please see Reponable Occu Tence R Nos. 50 2W79 4, 50-20/83-2, 50-2(V85-2, 50-2(V86-1, 50-20/86-2, and 50 2 -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 Advanec4 Test Reactor (ATR) fuel. Atomics Internanonal completed the pmduction of 41 of the more hi;hly loaded elements in 1982,39 of which have been used to some degree. Nineteen with a mut 40% burnup have been discharged because the have attained the fission density limit.

Additional elements are now bemg fabricated b Babcock & Wilcox, Navy Nuclear Fuel Division, Lynchburg, Virginia. Twelve of these ve been received at MIT and are now in usC.

He 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 loadings up to several times the recent upper limit of 1.6 grams total uraniunVcubic l

6-l centimeter. Consideration of the thermal hydraulics and reactor physics of the hilTR Il core design show that conversion of hilTR il fuel to lower enrichment must await the successfut demonstration of the proposed advanced fuels.

4. Chances in Perfonmnce Chancteristics Performance characteristics of the hirrR 11 were reported in the "hilTR Il Startup Report." hiinor changes have been described in previous repor. There were no changes donng the past year.

5. Chances in Oncatine Procedures Related toSpG:1y No amendments to the facility operating license were issued during the past year.

Ilowever, an amendment concerning 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 approval, a summary is given below of those changes implemented during the past year.

'Ihose changes related to safety are discussed in secuon E of this report.

a) PM 3.1.1.2, " Full Power Startup Checklist - Two Loo ) Instrumentation" and PM 3.1.6, " Restart Following an Unanticipated or a arief Duration Scheduled Shutdown" were revised to include two independent calculations of the reactor's estimated critical position or ECP Previously, the ECP had been calculated by the reactor operator and checked by the supervisor. Now it is done independently by both and the results compared. This revision was part of the cor:rctive action for ROR #50 20/90-1. (SR #0-9012) b) PM 2.7.4, " Removal of Spent Fuel" was revised to allow MITR fuel to the depleted to a higher percentage of the existing technical specification limit. The basis for this revision was improved quality assurance information on the fuel fmm the vendor.

(SR #0-B13) c) PM 6.1.1, " Emergency Cooling System" was revised following relocation of the piping used to test this system. The relocation was necessitated by the installation of an experiment in the area formerly occupied by the test line draln and collection tank. (SR 49014) d) PM 6.5.10.2, " Vacuum Breaker Calibration" was modified to delete the rec uirement for an annual calibration of certain vacuum breakers. This action rec uced radiation ex posure to personnel and did not affect facility safety because the vacuum breakers being calibrated were not essential to reactor operation.

(SR #0-90-15) e) A " Quality Assurance Program for MIT Spent Fuel Shipment" was prepared, submitted to, and approved by NRC. This is further discussed in section A.7 of this report. (SR #0 9016) f) PM 4,4.4.12 " Reactor Containment Evacuation," PM 4.4.4.14," Excess Radiation at the Exclusion Area (Site) Boundary Resulting from a Contained Source," and PM 4.4.4.15, " Escape of Airborne Radicactive Material from the Containment Building" were revised to standardize the basis for calculating stay times.

Previously, different procedures had used different figures for the allowed dose.

(SR #0-9424)

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l g) The administrative procedures, Chapter One of the Procedun Manual, were revised to update the lists of names and committee memterships. (SR #0-91 1) h) PM 1.1.1, "MIT Administration and Committees" was updated to fonnally i incorporate the function of the ' Committee on Radiation Exposure to lluman Subjects.' (SR #0-915)

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 te 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 donc cach 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 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 Technical S pecifications, and the results of tests and inspections were 3

satisfactory throughout the year I or Facility Operating License No. R 37.

7. Status of Snent Fuel Shinnent i Pursuant to Amendment No. 25 to Facility Operating License No. R-37, paragraph 2.B (2) sub)aragraph (b), reponed herewith is the status of the establishment of a shippinE 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 t prepared for use by the non power stactor community. After the Mil-1 A cask became unavailable, MIT made arrangements with General Electric to use the GE 700 cask for shipment of the MITR spent fuel. When the GE-700 cask was removed from service voluntarily by GE, the BMI 1 cask became the only one available that is approved for

transponation of irradimi fuel elements.

Relative to the capability of shipping spent feel from the MIT Research Reactor using the BMI-l cask, the following has been accomp!!shed:

(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 te acceptable for shipping MITR spent fuel. Arrangements have been made with both the cask owner, Cinti Chem, Inc., and DOE for MIT to use this cask.

(b) The University of Missouri Research Reactor (MURR) basket was reviewed and found to be smtable for use with the MIT fuel elements in the BMI-1 cask. MURR has agreed to make their basket available to MIT forihe required shipments, f

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(c) A quality assurance program for hilTR-Il spent fuel shipment was prepared and approved under the hilTR safety revi " program. This Q/A program was approved by NRC on July 23,1991.

(d) The decay heat load of each spent element was determined by a member of the hilTR staff and found to be within the limits specified in the Certificate of Compliance for the cask. Radiation shielding calculations were also perfonned and radiation levels associated with the loaded cask were estimated to be within allowed limits. Criticality calculations are near completion and are being Nrformed using the hionte-Carlo Code KENO-V which was obtained fmm the Radiation Shielding Infornution Center of the Oak Ridge National Laboratory. Preliminary results show that the degree of sukriticality of a cask fully loaded with hilt fuel elements is within specification.

(e) In order to cmss-check the cross sections used in the KENO V code, criticality analyses are being perfomied using a second hionte Carlo code, h1CNP. Results obtamed thus far are 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 hilTR il spent fuel elements are being compiled for submission to the spent fuel processing center.

(g) All spent elements in the hilTR spent fuel storage pool were rearranged and grouped in accordance with our procedure for shipment preparation. A special structure for support of the Bh11 1 cask is being designed and fabricated.

(h) Additionally, an order for a thhd fuel storage rack, which has a capacity of twenty-five fuel elements, was placed with our own machine shop.

As is evident from the above, MIT is in position to shia its spent fuel.

Unfortunately, further delay appears hievitable. Specincally, the Bhil , cask is currently unavailable for use by the non power reactor community because the owner of the cask, Cinti Chem, is in the process of shipping its own spent fuel elements off site. Itis anticipated that the cask wi!! not be released for use by other research reactors until Cinti-Chem's shipments are complete. The earliesa completion date of Cinti-Chem's shipments is the beginning of 1992. MITR IIis one of several research reactors awaiting use of the cask. The current projection shows that the cask will not be available for MirR's use until well into 1992 or possibly even much later. Unless Cinti Chem's position on the BMI l cask changes, it will be necessary for MIT to request a further extension of the tempwy possession limit contained in amendment No. 25.

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.) 7. , 7 9-11 . REACTOR OPERATION Information on energy generatu! and on reactor operating hours is tabulated below:

Quaner 1 2 3 4 Total

1. Energy Generated (MWD):

a) MITR II(MIT FY91) 70.2 86.8 228.5 271.9 657.4 (normally at 4.9 MW) b) MirR-11 10'624'1 (Mir FY76-90) c) M M 1 10'435.2 (MIT FY59-74) d) Cumulative. 21'716'7 MTTR 1 tc MITR Il

2. MITR-II Operation (firs):

(MIT FY91) __

a) At Power

(>0.5 MW) for 345.8 464.5 1154.4 1384.5 3,349.2 Research b) Low Power

(<0.5 MW) for 457.3 229.2 133.1 63.9 883.5 Training 0) and Test c) TotalCritical 803.1 693.7 1287.5 1448.4 4232.7

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.

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C. SHUTDOWNS AND SCRAhiS During the period of this report there were 11 inadvenent scrams and 9 unscheduled power reductions.

The tenn " scram" refers to shutting down of the reactor through protective system action when the reactor is at power or at least critical, while the tenn " 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 power loss without pro.ective system action are included in shutdowns.

He following summary of scrams and shutdowns is provided in appmximately the same format as for previous years in order to facilitate a comparison.

1. Nuclear Safety System Scrams Igul a) Channel #2 connectors. I b) Channel #3 power supply. I c) Channel #4 trip set inconectly (too low). I d) Channel #6 power supply. 1 e) Channel #6 noise. 4 Subtotal 8

11. Process System Scrams a) Inadvenent activation of grid latch scram test button. I b) Core outlet temperature recorder battery failure. I c) Noise on s condary flow reconfer. 1 Subtotal 3 8

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I1 111. Unscheduled Shutdowns or Power Reductions l a) Control rod drop as result of magnet failure. 2 i b) Shutdown due to PCCL or IASCC experinent malfunction. 3 c) Shutdown due to loss of offsite electricity. 4 Subtotal 9 Total 20 Experience during recent years has been as follows for scrams and unscheduled shutdowns:

Fiscal Year Numter 87 21 88 21 89 18 90 20 91 20

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D. MAJOR hiAINTENANC11 Major maintenance projects during FY91, including the effect, if any, on safe operation of the reactor are described iri this section.

Major maintenance items were continued in FY91 to support the requirements of the dose reduction projects for light water reactors. These projects are the Pressurir.ed Water Reactor Coolant Chemistry Imop (PCCL), the Boiling Water Reactor Coolant Chemistry loop (BCCL), and the Irradiation Assisted Stress Corrosion Cracking (IASCC) Loop.

Plumbing and electrical cables that had been run on a temporary basis on the reactor top platform for the PCCL and BCCL were installed permanently under the deck plate:, on the reactor top. This eliminated any protmsion on the reactor top that may have caused a tripping hazard. An area on the reactor top was cleared and designated as the location for the auxiliary e,qmpment required i y the IASCC Loop. Ilowever, additional space will be needed and tha will be obtained by building a platform that is supported from the reactor floor and proviles added Door space adjacent to the reactor top. The initial design of this extension alatform is being reviend by a licensed structural engineer. Final design will be completec in FY92. The 1ASCC Loop will also require a reactor top shield lid that is sigruficantly different from the one currently in use. Design of a new lid was completed in FY91. Fabrication and quality assurance of the new reactor top shield lid will be performed in FY92.

In addition to supponing these dose reduction projects for light water reactors, major evolutions were also performed for development of new experiments that use other .

facilities. The environmental enclosure, " blue house," for the S-1 spectrometer was removed to allow better access to the four inch through ports. Both the four and six inch through ports are being evaluated for use as irradiation facilities. Future development of these facilities will require re >ositioning of the existing spectmmeters on the main reactor floor. Modifications were a so made to the 6 CHI fast spectrum irradiation facility for testing of electronic components. A flux attenuator was designed and installed in this facility to provide radiation conditions more similar to the actual working environments of  ;

the components that are to be tested.

It was desirable to enhance the ratio of epithermal to thermal neutmns in the medical therapy room so as to increase the " advantage depth" during treatment of brain tumors. A ,

neutron filter and its associated shielding and su?pon pieces were designed and fabricated.

The old shielding pieces were removed and tie new filter and shields were installed.

Radiation measurements were made following the installation to confirm the proper design and installation of the filter.

Neutrons originating in the reactor core are admitted to the 6 CHI facility by operating a set of steel doors. A pin on the gear and chain drive of the steel doors broke off during operation. The steel doors are located in the hohlraum region which is next to the thermal column. Entry to this area was gained through the Blanket Test Facility which is at ,

the end of the hohlraum region. De inoperative drive train was repaired.

The gasket on the inner door of the main personnel airlock failed during use. Tine airlock was tagged out and removed from service. De gasket uses a design that is unique to the MITR and requires custom fabrication. The defective gasket was replaced with an unused one that was on hand and an order for two new gaskets was immediately placed with the manufacturer of the airlock. A leak test for quahty assurance was performed and i

i the result was satisfactory. De airlock was retumed to service at the end of FY91.

13-The outer door of the containment vehicle (truck) lock failed at one of the areas where *!.e hydraulic rams make contact with the door. De face plate at the contact area was corroded and broke loose from the door flange causing insufficient pressure on the gasket to seal the door. The corroded face plates were discarded and affected areas of the door flange were ground to bare metal and weld reinforced. The exposed areas were then painted. The repair and a pressure test on the vehicle lock itself are scheduled to te completed in FY92 he vehicle lock in the meantime is tagged out-of-service and its use is prohibited whenever the reactor is in a non-secured condition. (&1c: his failure occurred during the 1991 annual building pressure test. The outer door was functioning properly prior to that test.)

When the rust spots on the containment shell were removed in FY90, the " ice breaker," which is an angle iron welded to the exterior shell at mid height of the containment building, was found to have deteriorated. Repairs were perfomied in FY91 to have all corrosion removed by grinding and sandblasting. A new stainless steelice breaker was installed, primed, and painted with top coats that are the same as the rest of the containment building.

The hemispherical portion of the containment building was painted in FY91. The condition of the paint on the remaining portion of the containment building was acceptable and did not requtre painting.

While performing a routine surveillance arocedure on the containment cathodic protection system, the readings on some of the e ectrodes were found to be low but still within specified limits. An outside company that specializes in cathodic protection was hired to evaluate the effectiveness and condition of our system. The preliminary report from the companr has been received and it suggests methods for improving our system.

Investigation of the containment cathodic protectico system will be continued in FY92.

One of the main heat-exchangers, llE 1 A, was drained and visually inspected for crud buildup prior to and following annual cleanup of the cooling tower basins and main hiat-exchangers, it was concluded from these inspections that the current chemistry centrol program used in the secondary system is effective and also that the annual peruxide cleaning of the heat-exchangers remains effective.

As an improvement to the control room, new equipment was acquired to Irplace the primary core outlet temperature recorder. De new recorder has a large dial indicator, a digital readout, and a 24-hour circular chart. The primary piping was drained to allow installation of a new thermowell for the thermo-resistance probe. The recorder itself was installed in the control room with the digital readout mounted on the main console in direct view of the operator.

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

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E. SECTION 50.59 CilAN(11Lc r TESTS. AND EXPERlhWD1 This section contains a der,cription of each change to the facility or procedures ind of the conduct of tests and experiments canied cut under the conditions 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 prmedures as described in the SAR are documented in the MITR records by means of " Safety Revie~w Fonns". These have been paraphrased for this report and are identified on the following pages for ready reference if funtwr infomation should be required with irgard to any item.

Pertinent pages in the SAR 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, Ofnce of Nuclear Reactor Regulation, USNRC, he conduct of tests and experinents on the reactor are nomudly documented in. the experiments and irradiation files. For experiments carried out under the provisions of 10 CFR 50.59, the review and approval is dccumented by means of the Safety Review Fonn.

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

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Pressurized Coolant Chemistry Looo (PCCl3 SR #0-86-9 (04/21/88),0-88J4 (07/2F188),0-88 5 (09/09/88),0-88-14 (12/07/88),0-89 2 (01/06/89), 0 89 3 (01/19/89), 0-89-6 (01/24/89),0 89 9 (06/02/89,. 89 14 (06/19/89),

0-90 6 (03/20/90). 0 90-7 (03/20/90), 0 90 8 (03/20/90), 0 90-9 (03/20S 0), 0 90-25 (l2/10NO),0-90 26 (12/18/90),0-90-27 (12/18NO),0 91 8 (05/21N1).

This project involves the design, installation, and operation of a pressurized light-water loop in the MITR core for the purpose of studying the production, activation, and e transpon of corrosien to detennine the optimum method for reducing the creation of activated corrosion products (crud) and thereby reducing radiation fields associated with 3ressurized water reactors (PWRs). The ultimate goalis 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 ufety question existed because no failure or accident associated with the PCCL could lead to an accident or failure involving reactor components. Details of that detemilnation, together with safety review #0 86-9, were submitted to the U.S. Nu.' lear Regulatory Commission on 04/21/88.

Subsec uent to the determination that r,o unreviewed safety question existed, specific procec;ures for PCCL operation were prepared. nese included:

- Pmcedure for Ex-Core Testing Supplement to the Safety Evaluation Repen PreoperationalTest Procedure

- Abnormal Operating Procedures for the PCCL

- Procedures for PCCL Stanup/ Shutdown Procedures for PCCL Installation / Removal Experiments using the PCCL began in April 1989 and have been quite successful. During the period covered by this repon, several changes were made to the ex-core ponion of the PCCL. Rese were:

- Addition of a splashguani to prevent condensate that forms on the underside of the reactor lid from dropping onto the PCCL lid. Should condensate fall onto the PCCL lid, the result mig 1t be the formation of radionuclide particles of

' exceptionally small size. These would pass through the rough filters in the reactor's air purge system and cause excess radiation levels in that system's fine fihers. %c splashguard is anchoird to preclude its coming loose. (SR #0-90 25)

- Installation of an inolttion transformer between the PCCL heater power source and the heater power controller to prevent arcing in the event of a short to ground fnxn De heater lead. Reliability of the heater is meirased because the transformer eliminates the ground return pi.h from the heater power controller when a short occurs. (SR #0-90-26)

- he in-core portion of the PCCL is contained in a titanium can. A change was approved to the method by which these cans are manufactured. His change

., . . . j 16 - i I

reduced the amount of machining and hence the amount of cold work impaned to i the can. His in turn lessens the probability of warpage once the can is in use. l

- (SR #0-90 27) i

- A reduction in the coil resistance of the PCCL's heater was necessitated by a  !

change in the available dimensions of the nichcome heating element. This l' tum j resulted in a need to increase the setpoint of the heater's power supply circuit  :

breakers from 70 to 100 Amperes. The parameters of the modified heater are j within the envelope of those originally approved and allinterlocks were retained.  ;

(SR #0 918) >

None of these changes involved an unreviewed safety question. All served to increase the t reliability of the experiment.. j i

Experiments that make use of the PCCL facility continued throughout this entire repoaing  !

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Bollinn Coolant Chemhtrv Imo (BCCL)

~

SR Mi8914 (06/19/89),0 89 20 (12/20/89),0-90-17 (09/17/90),0-90-18 (09/14/90),

OM20 (10/1580)

This amject involves the design, installation, and operation of a boiling light water

. loop in the MITR core for the purpose of studying t, e production, activation, and transport of corrosion products. The effect of various wat r chemistries will be examined to determine the optimum method for reducing the creation of activated conosion products (crud) and thereby reducing radiation fields associated with boiling water reactws (BWRs).

The ultimate goal is to reduce radiation exposures to BWR maintenance personnel. j In 1988 and 1989, the Reactor Staff made a detennination that boiling within an in-core facility is not contrary to the technical specifications provided that teactivity 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 l carefully controlled experiment was proposed to demonstrate that boiling within an in-core j

. facility would not adversely affect reactor operation. Following both a determination that l

. no unreviewed safety question was involved and approval by the MIT Reactor Safeguards  !

' Committee, this expenment was conducted. *lhe results were as expected.

The final safety evaluation repon for the BCCL was completed on 8 March 1989

- and approved by the MITR Staff. On 12/20/89, the MIT Reactor Safeguards Committee j 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 fonvarded to the U.S. Nuclear Regulatory Commission pursuant to 10 CFR 50.59(b)(2).

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

Preoperational Test Procedure.

- Abnormal Operating Pmcedures for the BCCL

- Procedure for BCCL Startup.

Other necessary procedures such as BCCL shutdown and installationhemoval are the same as those previously develo xd and approved for the PCCL Experinents using the BCCL began in October 1990 anc all went as expectxt.

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Emeriments Rehted to Neutmn Capture Thergy [

SR #0-89-4 (01/23/89),0 89 8 (03/01/89),0-9l 7 (05/0@l)

L in conjunction with the "Iuits - New England Medical Center and with the support i of the U.S. Department of Er.ergy MIT is designing an epithermal neutron beam for the i treatment of brain cancer (glioblastoma). ' thermal beams hav been used successfully for l this treatruent in Japan. The reason for designing an epithennal beam is to allow tunwr  ;

treatment without having to subjt;ct the patient to surgery involving removal of a portion of  ;

the skull. Also, an epithermal beam gives greater penetration. Thus far, the research has i consisted of simulation studies using Monte Carlo codes and experiments using the MIT  !

Reactor's medical beam to verify those studies. Two facility changes were previously l made. These were.  ;

i (1) Installation of a liner and a support plate in the medical therapy beam. F These wtae installed to pemtit the subsequent installation of candidate filter  !

materials for producing an epithermal beam. (SR #0-89-4)  !

(2) Installatior. of a candidate filter. Currently, filters containing sulphur and aluminum with small quantities of lithium and cadmium appear to give the best results. One such filter was installed on a trial basis for the purpose of confirming the results of the simulation studies. (SR 689-8)

Neither of these design changes was judged to involve an unreviewed safety question.

Efforts during the present reporting period concentrated on filter design, evaluation of the treatment's efficacy using ammal models, dosimetry and boron distribution studies, and patient treatment planning. In May 1991, an improved filter was built and installed (SR #0-91-7). Standvd procedures for filter installauon and removal were also prepared as pan of that effort. This change in filter design was judged not to involve an unreviewed safety question.

Experiments were conducted throughout the year to evaluate the efficacy of the proposed treatment. 'Ihis entailed inoculating specially bred mice with tumor and then treating them. Those receiving the combinanon of boron compound and neutrons (3000 - 4000 RBE-Rads) have shown excellent survival. All other groups (no treatment, boron compound only, boron compound plus ganuna radiation, gamma radiation only, neutron radiation only) have not.

l The Massachusetts Institute of Technology (MIT) has been in written contact with (Refer to the U.S. Nuclear correspondence dated Regulatory Commission concerning Asthis0403/91, has been noted04/16/91, 06/0 areviously, it is MIT's posiden that une of its neutron beam for medical therapy is allowed xcause (a) the MIT Resean:h Reactor is licensed under both 50.21(a) and 50.21(c) of 10

' CFR, (b) an internal administrative structure with the ex penise and activity appropriate to I review the design and use of such a beam for medical t;wrayy is in place and functional, and (c) mechanistas are in place whereby NRC may monitor c ccisions made concerning the design and use of this beam.

In Scatember 1991, MIT and its partner the Tufts - New England Medical Center (NEMC) wiL1 formally begin the arduous process of obtaining approvals for patient treatment well as the U. usink.the Food and DrugMilR neutron beam fmm the cognizant MIT and NEMC co Administration.

l i

19-Djcital Computer Control of Reactors Under Steadv State and Transient Conditions SR# ht 813 (11/17/81), M-814 (12/10/81), E 82 2 (OlA)8/82), E 82 3 (02/24/82),

E 82 4 (03A)3/82), E 82 5 (04/14/82), E 82 6 (07/13/82),0 83 5 (02/03/83), E 831 (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 (l1/28/86),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)

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

.alidation procedures, en:;ures 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 demonstrate the impmvements that can be achieved in reactor contml.

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 contmiler were conducted in 1983-1984 using the facility's regulating rod which was of relatively low reactivity worth (0.2% AK/K).

Following the successiul completion of these tests, facility operating license amendment l No. 24 was obtained fmm NRC (April 2,1985). It permits:

1 (1) Closed loop contml 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 loo) control of one of the shim blades and/or the regulating rod provided t1st the overall controller is designed so that reactivity is constrained sufficiently to permit control of reactor power within desired or j authorized limits.

A successful experimentation pmgram is now continuing under the provisions of this license amendment. A pmtocol is observed in which the NRC-licen:cd supervisory controller is used to monitor, and if necessary override, other novc! controllers diat are still in development. Tests of novel contmllers are conducted under the pmvisions of technical l specification #6.4 which requires that iractivity be constrained to ensure " feasibility of

, control." Signal implementation is accomplished using a variable speed step sing motor.

l This motor is insta!!ed prior to the tests and removed upon their comp'etion. An

inde xndent hard wired cacult is used to monitor motor speed and preclude an ovenpeed

' concition, his arrangement for the conduct of these tests has been ap; roved by the MIT Reactor Safeguards Committee.

There were three major advancements associated with the research program on digital control du;ing FY91. nese were (1) an extensive upgrade to the digital control

-20 I system's haniwarr, (2) development and use of the perturled reactivity method for the on-Ime estimation of reactivity during automated power increases from suberitical, and (3) the comparative evaluation of many forms of undel based feedforwani control. Each of these advances is sumnurized below.

De hardwar: associated with the original MITR digital control system consisted of an LSI II/23 minicomputer and associated instrumentation. This system was used to perform all required control tasks and, while it performed superlatively for many years, was recently judged to have become obsolete. Specific denciencies were that a higher rate of numerical thmughput was needed, that both software telated to safety and software under development for control law evaluation were run on the same machine, that simulations had to be run on a separate machine, and that the associated instruments only covered one scale on the power range instmments. In 19891990, a multi-computer / single-task system was designed to allow continued growth of the research. It consists of five interconnected computers and has been designated as the Advanced Control Computer System. he five computers are:

(i) Rack-Mount 80386: This is an IBM AT computer that is used for data acquisition, execution of software essential to safety such as the code to implement the requirements of MITR Technical Specification #6.4, and the wnting of data to disk. Softwart 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 contml 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: his 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 PROLOO, LISP, or C.

(iv) IBM-XT 8088: his computer's mic is to receive validated signals fmm the data acquisition computer and to display nudel-based predictive infomution or a safety pammeter display on its screen.

(v) LSI-11/23: his 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 pasaed 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 19851icense arreA=nt issued by NRC for digital contml experiments at MIT and that no t;nreviewed safety question was involved. As pan of the installation of this new system, several preoperational test packages were pirpared 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.

In addition to this upgraded hardware, an auto ranging digital picoamneter has been installed to measure teactor neutronic power. his instrument pmvides both the level and range of the power signal. Moreover, it switches scales automatically and thus

. - 21 1

i facilitates the development of control strategies for automated stanups in which operation i i

over nuny decades of power is required. This instrument was subjected to a preoperational test in which its accuracy was verified. l The second major advancement made in FY 91 relative to the MITR digital control i

system was the development and demonstration of the 'penurbed reactivity method' which provides an accurate means for estimating a reactor's initial degree of suberiticality. This I technique, which is applicable to reactors described by space-independent kinetics, entails l perturbing a shutdown reactor by the insertion of reactivity at a known rate and then i estimating the initial degree of suberiticality from observation of the resulting reactor i i

period. It is therefore similar in certain respects to the conventional inverse kinetics approach. liowever,it differs in two imponant aspects. First, the penurbation must be ,

made at a known rate. Second, the net reactivity, p(t), present in the reactor is treated  !

separately using superposition as: j i

p(t) = pukn(t) + p g (t)  !

where p,g(t) is the reactivity present in the core excluding the seactivity associated with  !

the known penurtation, and l p g(t) is the reactivity associated with the known penurbation.

would nonnally be generated by moving a calibrated control device  !

The quantity connected to a pYigital controller. Also, it would normally be computed by means of a i balance using data from previously perfemed calibrations. It is this separauon of the net reactivity into two components that makes the perturbed reactivity method useful for  ;

automated startups. Specifically, there are no significant reactivity feedback mechanisms in a suberitical core. Hence, in a shutdown reactor, the quantity p uwill be the initial degree of suberiticality and it will remain constant during the startup. Yhus, repeated estimates can be made of this quantity and signal smoothing techniques can be applied to improve accuracy.

Implementation of the penurbed reactivity method requires a system model. That used is the alternate dynamic period equation which is a form of point kinetics. The method was first demonstrated via simulation. On line trials den followed in which it was used in an open loop manner. Closed loop trials were then successfully conducted in which automated power increases weie made from suberitical conditions. Both the MITR Staff and the MIT Reactor Safeguards Committee concluded that use of this method of reactivity estimation did not involve an unreviewd safety question.

De third major advance made in FY 91 relative to the MITR digital contml program was the comparauve experimental evaluation of model based feedforward control methods, included in this comparison were error-driven, pure feedforward, hybrid feedforward /

feedback, variable structure, and period generated control laws. All were tested in accordance with the now standard MITR protocol for the evaluation of novel control strategies. Namely, the NRC-licensed supervisory controller is used to tronitor, and if necessary override, the novel control law. The tests of this panicular class of control law were in every way similar to tests conducted earlier of ollwr control laws under this same protocol No unreviewed safety question was judged to exist.

Research on the digital control of nuclear reactors is slated to continue during the upcoming fiscal year. Emphasis will be on completing the installation of the upgraded control system.

22-4 Revicion of Fission Density Limit SR 68812 (12/01/88) i The fission density limit for the UAlx fuel used by the MIT Research Reactor is I

1.8 1021 fissions /cc. Researth conducted by the Idaho National Engineering Laberatory (Nucl. Tech.. 49,136-149, June 1980) shows that r limit of 2.3 1021 fissions /cc is technically justified. Anelysis of the MITR fuel cycle showed that increasing the MITR ,

12 fission density limit to 2.3 10 fissions /cc wouki eventually reduce the overall number of elements in the cycle. Accordingly, a safety analysis was prepared and, following review ,

and approval by the MfT Reactor Safeguards Committee, st;bmitted to the U.S. Nuclear  !

Regulatory Commission (NRC) on 13 February 1989. On 27 November 1989, NRC requested additional information. That material was forwarded on 6 July 1990. On 14 January 1991, the NRC requested ftrther additional information. The MITR Staff is in the ,

pmcess of preparing this material for transmission to NRC.  ;

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Irradiation Assisted Stress Corrosion Cracking HASCC) Exocriment ,

SR #0-89 15 (06/19/89),0-90-21 (10/22/90),0-90-22 (10/22/90),0-90-23 (l1/05NO) r This pmject involves the design, installation, and operation of an in-core facility for l the purpose of studying irradiation assisted stress corrosion cracking and in particular the combmed 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. l De IASCC is currently under design. When finished, it will provide the capability for placing steel specimens under a constant, measurable strain rate with those samples contained m 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 ,

, 1991. For the present, the steel specimens that will be used for the IASCC te.ts are being  :

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 l designed in 1979 and has been used previously for sample preconditioning.  ;

Three safety reviews were completed during FY 91 that were specific to the l IASCC. All were related to the samples undergoing preconditioning. These were a [

preoperational test of the sample irradiation facility, the issuance of abnormal operating i procedures to cover possible contingencies associated with the preconditioning, and a ,

pmcedure to test the efficacy of controlling sample temperature by adjusting the pressure of a heliutrVearbon dioxide cover gas. None involved an unreviewed safety question.

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F. ENVIRONMENTAL SURVEYS i Environmental surveys, outside the facility, were conducted using ladiation monitors and dosimetry devices. 'Ihc monitors located within a quaner mile radius of the facility consist of GM based detectors with associated electmnics. The dosimetry devices are film badges. The detectable radiation levels per sector due to Ar-41 are presented below. 'Ite quarterly monitors (film badges) indicated no detectable radiation for the a period in question.

Silt Exposure (07/01/90-06B0/91)

Nonh 0.03 mR/ year East 0.32 mR/ year South 0.01 mR/ year West 0.03 mR/ year Green (east) 0.01 mR/ year i

Fiscal Year Avenges 1991 0.1 mR 1990 0.1 mR 1989 0.2 mR 1988 0.2 mR 1987 1.2 mR ,

1986 1.8 mR

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G. RADIATION EXPOSURES AND SURVEYS WITilIN Ti1E FACILITY j A sununary of mdlation exposures received by facility personnel and experimenters I is given below:  :

July 1.1990 June 30.1991 Whole Body Exnosure Range (Rems) Number of Personnel  !

r No measurable ....................................................................109 l Mee s urabl e - < 0.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6 {

0.1 - 0.25 ..................................................................... 6 l

0.25 - 0.5 ....................................................................12 l 0.5 - 0.75 ..................................................................... 3 )

0.75 - 1.0 ..................................................................... 0 1.25 - 1.50 ..................................................................... 1 IDtal Man Rem = 9.41 Total Personnel = 177 From July 1,1990 through June 30, 1991, tiie Reactor Radiation Protection Office -

provided radiation protection services for the facility which included power and non. power j operational surveillance (performed on daily, weekly, monthly, quarterly, and other t frequencies as required), maintenance activities, and experimental proj,ect support. Specific j examples of these activides include, but are not limited to, the following: l r

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 cooling towers, D20 system, primary system, shield coolant system, heat exchangers, fuel storage facility, wasic storage *.anks, and expenmental systems. i c 3. Performance of radiation and contamination surveys, radioactive waste {

collection, calibration of area radiation monitors, calibration of effluent monitors, calibration of radiation survey instruments, and establishing /

posting radiological control areas. ]

4. Provide radiation protection services during fuel inovements, in. core  !

. caperimenta, nample irradiations, beam port use, ion coltunn removal, etc.  !

'Ihe results of all surveys and surveillance conducted have been within the guidelines j established for the facility.  !

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11. RADIOACTIVE EFFLUENTS ,

his i section summarizes the nature and amount of liquid, gaseous, and solid radioactive wastes released or discharged fmm the facility. l t

1. I kid Waste  ;

Liquid radioactive wastes generated at the facility are discharged only to the sanitary l sewer serving the facility. There were two sources of such wastes during the year: the i

cooling tower blowdown and the liquid waste storage tanks. All of the liquid volumes are measured, by far the largest being the 6,183.105 liters discharged during FY91 from the i cooling towers. (larger quantities of non radioactive waste water are discharged to the  !

sanitary sewer system by other pans of MIT, but no credit for such dilution is taken since i the volume is not routinely measured.) j All releases were in accordance with Technical Specification 3.81, including (

Part 20, Title 20, Code of Federal Regulations. All activities were substantially below the i limits sxcified in 10 CFR 20.303, but the monthly tritium releases are reported in -

Table lh3 in accordance with Technical Specification 3.81 because its concentration l exceeded 3x104 pCi/ml. j

2. Cagous Waste  ;

Gaseous radioactivity is discharged to the atmosphere from the containment ,

building exhaust stack and by evaporation from the cooling towers. 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 diludon faciar of 3000. Also, (

all were substantially below the limits of 10 CFR 20, Appendix B, Note 5, with the j exception of Ar-41, which is reported in the following Table 11 1. The 684.4 Ci of Ar-41 were released at an average concentration of 0.18x10-8 pCi/ml for the year. This i represents 4.4% of MPC (4x10-8 Ci/ml) and, given the 29% increase in reactor operating hours from FY90 to FY91,is consistent with the previous year's release of $42.6 Ca. This .

- reflects the continued success of our efforts to identify and climinate sources of Ar-41.

3. Solid Wacie Only one shipment of solid waste was made during the year, information on which  !

is provided in the following Table H-2. [

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Total activity less tritium in the liquid effluents (cooling tower blowdown, waste -

E storage tank discharges, and engineering lab sink discharges) amounted to 0.007 Ci for .

.- FY91. 'Ihe total intium was 0.115 Ci. 'Ihe total effluent water volume was 6.22x106 ,

liters, giving an average tritium concentration of 18.4x104 pCi/ml.

L Ihe above liquid waste discharges are provided on a monthly basis in tlw following ram a.3. ,

t i

--g w e -

, , , , , . - - - -w--,--.. _ ,

+m._.- m- m- ,

.--sa . n , , - , , n ., ,-,m -r - w,, -

---m, n_wn-,,m-vam-- n

- 27 TABLE 111 ARGON-41 STACK Rf'I PASES FISCAL YEAR 1991 I

I Ar-41 Average Discharged Corv:emrationU) I i

(Curies) GiCi/ml)

July 1990 33.5 0.11 x 10 August 26.0 0.07 September 13.9 0.05 October 33.0 0.11 November 53.6 0.14 December 40.5. 0.13 January 1991 53.1 0.14 Fettuary 76.1 0.25 March 83.1 0.28 April 100.2 0.33 May 82.9 0.22 June 88.5 0.29-

~

Totals (12 Months) 684.4 0.18 x 10-8 MPC(Table II, Column I) 4 x 10 s

% MPC 4.4%

' 0)After authorized dilution factor (3000).

TABLE 112 SUhihiARY OF hirILII RADIOACrlVE SOIID WASTE SillPhiENTS FISCAL YEAR 1991 ,

Description Units Shipment #1 Total Volume Cubic Feet 116.5 116.5 Weight Pounds 4093 4093  :

Activity 0) Curies 0.125 0.125 Date of shipment December 10,1990 Disposition to licensee for burial U.S. Ecology, Inc.

0) Radioactive waste includes: dry active waste comprised of irradiated components, contaminated items, and solidified wastes. The principal radionuclides are activation '

products such as "W 'ICr,65Zn, IUSb,187W,95h,95Mb,3H,463c, etc, f

6 y - - - - - - . , , .-

-r-._.-- ,

. _. _-.m. . _ . _ . _ - _ . . _ . _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _

is a+' .. l 29  ;

TABLE 113 MOULD WASH DISCllARGES f FISCAL YEAR 1991  !

~~

Total Total Volume Average Activity Tritium of Efiluent Tritium t less Tritium Activity Watedi) Concentration f (x104 Ci) (x10 3 Ci) (x104 liters) (x104 pCi/ml)

July 1990 2728 7.13 25.7 27.8 f

i Aug. 1.04 0.07 13.4 0.5 t

- t Sept. 3.1 3.05 10.9 27.9 {

Oct. NDAG) 0.02 20.8 0.11 l

Nov. NDAG) 0.15 40.3 0.37

{

Dec. 3.3 87.6 47.7 184.0  :

l Jan.1991 ND4G) 0.78 44.5 1.75  !

I Feb. NDAG) 0.97 56.6 1.71 l Mar. 27.8 1.05 90.2 1.16 i i

~ Apr. 2023 5.19 84.3 6.16  !

May 1281 5.92 74.4 7.96  !

1 June 1029 2.65 113.3 22.8 j i

12 months 7096 114.6 622.1 18.4 j 1

U) Volume of effluent fmm cooling towers, waste tanks, and NW12139 Ent;ineering lab [

sink. Does not include other diluent from MIT estimated at 2.7 million ga.lons/ day. 1 G) No Detectable Activity; less than 1.26x104 pCVml beta for each sample.  !

i f

i T

_I

_. ,- m , - - . . . , . . . . . . . . - . . . . , _ - , _ , - , , ,....,_,m,_.