ML19253B025

Annual Operating Rept for Jul 1978-June 1979

ML19253B025
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Site:	MIT Nuclear Research Reactor
Issue date:	08/29/1979
From:	MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE
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MIT RESEARCH REACTOR ANNUAL REFORT TO UNITED STATES NUCLEAR REGULATORY COMMISSION FOR THE PERIOD Ji?Y 1,1978 - JUNE 30,1979 BY REACTOR STAFF August 29, 1979 d

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. ... ~ ..-.. . . . . . . . . . . . . . . . . . _ . ..... . ... --

TABLE OF CONTENTS Section .

Introduction 1 A. Summary of Operating Experience 2 B. Reactor Operation 7 C. Shutdowns and Scrans 7 J

D. Major Maintenance 9 E. Section 50.59 Changes, Tests and 11 Experiments F. Environ = ental Surveys 16 G. Radiation Exposures and Surveys 17 H. Radioactive Effluents is a s:> M .r >g, Ug

. . _ - . . . . - . . - - . . . - . ~. . .. _ . .-. . - . - . _ _

1 Introduction This report has been prepared by the staff of the Massachusetts Institute of Technology Research Reactor for submission to the Director of Region 1. United States Nuclear Regulatory Commission, Office of Inspection and Enforcement, in compliance with the tequirements 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, 1956, the first year was devote & to startup experiments, calibration and a gradual rise to one megawatt, the initially licensed maximum power. Routine three-shift operation (Monday-Friday) commenced in July 1959. The authorized power level was_ increased to two megawatts in 1962 and five megawatts (the design power level) in 1965.

Studies of an improved design were first undertaken in 1967. The con-cept 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 maxinizing the peak of ther=al neutrons in the heavy water at the ends of the beam port re-entrant thimbles and for enhancement of the neutron flux, particularly the fast componentv 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 U Al internetallic 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 shield, cooling system, 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 cocponents for the modified reactor were procured and the MITR-I was shut down on :Sy 24, 1974, having logged 250,445 megawatt hours during nearly 16 years of operation.

The old core tank, associated piping, top shielding, control rods and drives, and some experimental facilities were disassembled, removed and subsequently replaced with new equipment. After preoperational tests were conducted on all systems, the U. S. Nuclear Regulatory Commission issued Amendment No. 10 to Facility Operating License No. R-37 on July 23, 1975.

This is the fourth annual report required by the Technical Specifications, and it covers the period July 1, 1978 through June 30, 1979. Previous reports, along with the "MITR-II Startup Report" (Report No. MlTNE-198, February 14, 1977) have covered the startup testing period and the tran-sition to relatively routine reactor operation. This report covers the second full year of routine reactor operation at the 5 MW licensed power level. It was a year in which the safety and reliability of reactor operation fully met the requirements of reactor users.

84.ODO

2 A summary of operating experience and other activities and related statis~ical data are provided in the following Sections A - H of this report.

A.

SUMMARY

OF OPERATING EXPERIENCE

1. General During the period covered by this report (July 1, 1978 - June 30, 1979), the mil Research Reactor, MITR.-II, was operated on a routine, four days per week schedule, normally at a nutinal 5 MW. It was the second full year of no mal operation, the startup program for the MITR-II having been completed during FY77 (November 1976).

During FY79, 9 Tuesday - Saturday schedule was in effect, and prevailed with a few exceptions throughout that year. This permitted maintenance and experiment changes, protective system surveillance tests, and partial com-pletion of the checklists on the Monday day and evening shifts, followed by an early secrt on Tuesday. Some maintenance and surveillance testing was also accomplished after shutdown on Saturday, which generally occurred in the forencon.

In December 1978 the reactor operating hours were reduced from G5-90 hours per week to about 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> per week as a measure to save fuel, since the fuel fabricator's delivery schedule was slipping badly, and the ex-haustion of the MITR fuel inventory by Fall 1979 appeared to be a real possibility. For the year, the reactor averaged 79.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per week at full power.~~ At the time of eriting this report, the fuel fabricator is well along in production, the delivery schedule is much more certain, and the reactor returned to a 90-hour week in July.

The reactor was operated throughout the year with 24 elements in the core. The remaining three positions were occupied either by irradiation facilities or by solid aluminum dummy elements. Reactivity was gained (to compens ate for burnup) by making five refuelings in which two or three fresh eleme nts were substituted for partially burned elements in the B-ring.

The latter were held in the core tank storage ring for subsequent use as replacements for C-ring elements. In a sixth case, and also in one of the above five (ases, the element in Position A-2 was replaced by another having less Surnup. In a seventh case, the A-2 ele =ent and the 15 C-ring elements were all inverted (flipped end for end), which gained reactivity and also took advantage of this unique feature of the elements in order to achieve more uniform burnup.

Three other shuffles of fuel into and out of the core ~took place in June and July 1979 for the purpose of removing and identifying an element with faulty cladding which had been leaking small amounts of fission pro-duct gases into the primary coolant. Two other core alterations were related to changes in the type or location of irradiation facilities.

As in FY78, the reactor was operated throughout the period without the fixed hafnium absorbers, which were designed to achieve a maximum MSZ60

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peaking of the thermal neutron flux in the heavy water reflector beneath the core. These had been removed in November 1976 in order to gain the reactivity necessary to support more in-core facilities.

2. , Experiments The MITR-II was used throughout the year for erperiments and irradiations in support of research and training progra=s at MIT and elsewhere.

Experiments and irradiations of the following types were conducted:

a) Neutron diffraction npectrometer alignment and studies (3 ports).

b) Sh121 ding studies and component alignment for an inelastic scattering spectrometer for molecular dynamics studies.

c) Reactor physics experiments in facilities equipped with neutron fission converters for fast breedet reactor blanket studies.

d) Dosimetry measurements of the neutron beam in the medical therapy facility.

tj Dosimetry measurements for pneu=atic rabbits and other irradiation facilities.

f) Irradiations of biological, geological, oceanographic, and medical

--specimens for neutron activation analysis purposes.

g) Activation of ablation monitor wires for re-entry vehicles.

h) Production of molybdenum-99 and gold-198.

1) Irradiation of tissue specimens on particle track detectors for plutonium radiobiology, and other studies, j) Use of the facility in reactor operator training.

k) Irradiation damage studies of candidate fusion reactor materials.

1) Studies of fatigue failure as a function of surface bombardment and bulk irradiation damage.

3. Changes to Facility Design A third primary coolant heat exchanger, which operates in parallel with the two previous exchangers, was installed during the year. NRC approval for the installation was received in the form of Amendment No. 14 to the Facility Operating (Licence No. R-37, dated August 25, 1978.

After preoperational tests in October, the unit was placed in service and performs as predicted. It provides insurance against a long-term power reduction if any of the three exchangers fails, since two are normally sufficient for 5 MW operation, and it is possible to operate with any two or with all three.

MOrDI.

4 As indicated in last year's report, a study cf the feasibility of increasing the uranium loading in the MITR-II fuel proved satisfactory, and specifications for fabrication now call for 34 gra=s U-235 per plate, 510 grams per element,as compared to 29.7 grams and 445 grams respectively in the first and only other fuel procurement. The new loading results in 41.2 w/o U in the core, based on a 7% void fraction, and corresponds to the maximum loading in Advanced Test Reactor (ATR) fuel. The higher loading will increase the reactivity available from the fuel and increase its lifetime (subject to the maximum burnup permitted by the Technical Specifications).

Other changes in the facility design are reported in Section E.

4. Changes in Performance Cbsrecteristics Performance

• characteristics for the MITR-II were reported in the "MITR-II Startup Report", and no significant changes have occurred since that time.

5. Changes in Operating Procedures Related to Safety There. were two amendments to Facility Operating License No. R-37 during the year (in addition to Amendment No. 14 authorizing the in-stallation of a third primary coolant heat exchanger). Amendment No. 15 incorporates the " Security Plan for the MIT Reactor" into the License.

Amendsent No.15 adds a definition of the term " frequency" os applied to surveillance tests and other periodic activities.

Wi'.h respect to operating procedures, a summary of those related to safety is given below:

a) A revised Trocedure 1.16, "Requalification Program for Licensed Personnel", was adopted after approval by USNRC on May 26, 1978 (SR # 0-78-16).

b) A step was added to surveillance Procedure 6.1.1, Emergency Cooling System, to attach lead seals to the ECCS valves after returning them to the normal position on completion of the semi-annuti test. Steps were added to the startup checklists, Procedures 3.1.1.2 and 3.1.2.2, to check the valve alignments and seals (SR # 0-78-17). This was in response to a recommendation in Occurrence Report 50-20/78-4.

c) A new Procedure 4.4.4.9 was written to specify the steps required for operation of the containment building pressure relief system (SR # 0-78-20). Prior written procedures had dealt only with the system use for other purposes, i.e. weekly exercise of the char-coal filters and annual building pressure test.

d) Startup ?rocedures 3.1.1.2 and 3.1.2.2 were revised to provide for recording of the shutdown margin on those checklists in order L

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• 5 to give weakl, documentation of its value (SR # 0-78-21). They were furtler revised to incorporate steps related to startup using the (nird primary heat exchanger (SR # 0-78-26). The abnormal Occurrence Procedures (AOP's) and the Test and Calibra-tion Procedures were also revised as required to include the third exchanger (SR # 0-78-28). '

e) Administrative Procedure 1.10.8, " Experiments and Irradiations",

was revised to incorporate the requirements for use of the Building NW13 pneumatic send station, including approvals, daily pre-use tests, and new record forms. (SR #0-78-25).

f) A new AOP, #5.7.9, was prepared and approved to prov ide guidance to the operator in the event sf an alarm on a " fatigue cracking

• experiment installed in an in-core irradiation thimble (SR # 0-78-29 and SR # 0-79-2).

g) A new procedure, " Operating Procedure for Shipping MITR-I Core Tank", 3.10.1, was prepared and approved to cover the sectioning and shipping of the tank (SR # 0-79-1). The sectioning has been nearly completed, but shipping has been delayed due to diversion of the planned shipping cask to Three Mile Island.

h) Startup Procedures 3.1.1.2, 3.1.2.2 and 3.1.3 were revised to clarify the method of setting the level trips on the safety channels and to incorporate specifically a step for rechecking the trips at

. full power (SR # 0-79-4). These changes were in response to a re-commendation in Occurrence Report #50-20/79-1.

1) A new " Procedure for Flux Mapping", 6.5.17, was developed and approved (SR # 0-79-5) for measuring the flux shape with U-Al foils.

j) A new Procedure 6.5.18, Control Blade Thickness Measurement, was developed and approved in order to provide a written procedure for this measurement (SR # 0-79-6).

k) Miscellaneous minor changes to operating procedures and equipment were approved and implemented throughout the year.

6. Surveillance Tests and Inspections There are many written procedures in use for surveillance tests and inspections required by the Technical Specifications. These procedures provide a detailed machod for conducting each test or inspection and specify an acceptance criterion which must be met in order for tue 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-three such tests and calibrations are conducted on an annual, semi-annual or quarterly basis.

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

During the reporting period, the surveillance frequency has been at least equal to that required by the Technical Specifications.

The results of tests and inspection < required by the Technical Specifications have been satisfactory with two exceptions (Reportable occurrences # 50-20/79-2 and # 50-20/79-3), where NRC was notified in accordance with Technical Specification 7.13 in each case. Conditions having safety implications were found as the result of other tests and inspectiona and bere likewise reported to NRC (Reportable Occurrences

1. 50-20/78-4, # 50-20/79-1, and # 50-20/79-4).

943ZG4

7 B. Reactor Operation Information on energy generated and on reactor operating hours is tabulcted below:

Quarter Total 1 2 , 3 4

1. Energy Generated (MWD):

a) MITR-II (MIT FY79) 219.0 211.8 188.9 198.2 817.9 (normally at 4.9 MW) b) MITR-II (MIT FY76-78) 1,734.7 c) MITR-I (MIT FY59-74) 10,435.2 d) Cumulative, MITR-I & MITR-II 12,987.8

2. Hours of Operation gMIT FY1979, MITR-II a) At Power (>0.5 MW) 1165.4 1061.2 936.0 988.2 4,150.8 for research b) Low Power (<0.5 MW) 39.8 61.9 35.6 72.4 209.7 for training (1) and test c) Total critical 1205.2 1123.1 971.6 1060.6 4,360.5 Note: (1) : These hours do not include training conducted while the reactor is at full power for research purpose s (spectrometer, etc.) or fo r isotope production. Such hours are included in previous line.

C. Shutdowns and Scrams During the period of this report there were 33 inadvertent scrams, and ten un-scheduled power reductions or shutdowns. If the multiple failures due to power supplies and old channel 2 failures (le (ii) and (Ii) are eliminated, the total is

~

not far above last year's 22.

The term " scram" refers to shutting down of the reactor through protective system action when the reactor is at power or at least critical, while the term

" reduction" or " shutdown" refers to an unscheduled power reduction or shutdown to suberitical by the reactor operator in response to an abnormal gondition indication.

Rod drops without protective system action are included in shutdowns.

The following summary of scrams and shutdowns is provided in approximately the same format as last year in order to facilitate a comparison.

I. Nuclear Safety System a) Period channels during normal startup, resulting from electrical 3 noise b) Level channel high trip due to noise 1 c) Electric Company power dips 2 d) Level channel tripped on high level due to trip being sc slightly 4 under 5 MW (should be about 5.5 MW) 943255

8

-e) Electronic component failure in:

I) Levri channel i II) Channel power supplies; rebuilt 7 f) Withdraw permit circuit open for no '

2 apparent reason g) Operator error in deactivating channel I h) Technician error in doing maintenance 1 i) Channel 2, caused by old chamber failing; 7 chamber replaced 4

j) Low Voltage Chamber Power supply scram caused 1 by water from experirent cooling line leaking into instrument port Subtotal 30 II. Process System a) Primary outlet temperature scram when cooling 1 tower fan tripped off due to vibration b) Low level core tank scrams caused by failure 2 of insulation in level probe; insulated Subtotal 3

-III. Ot_her Scrams or Unscheduled Shutdowns a) Operator shutdown by "All Rods In":

(i) To investigate a decrease in Channel #2 1 chamber output caused by leak in experi-ment cooling line (ii) High reading on core purge monitor; replaced 1 element with faulty cladding (iii) After trip of containment building exhaust 1 fan due to overload when fan speed was in-creased slightly; overload trip level in-creased within rated limit.

(iiii) To investigate loss of helium supply for 1 irridiation thimble; supply line repaired b) Operator lowered power to 500 KW:

(1) To check functiontng of recombiner 1 (ii) To check local indication of flow in 1 auxiliary heat exchangers OdbrSbO

9 c) Operator lowered power to 2.5 MW:

(1) To investigate and correct sticking of regulating rod direction relay 1 (ii) To investigate rising temperature in Fatigue Cracking Experiment . 2 (iii) When cooling tower fan tripped off (no apparent reason) 1 Subtotal 10 Total 43 d

D. Major Maintenance Major maintenance projects during FY79, including the effect, if any, on safe operation of the reactor, are described in this section.

1) A program to upgrade instrumentation was continued during Ff79. Because of excessive maintenance and sosolescence on older units, some of which had been in use since the initial operation and for which it is in-creasingly difficult to obtain spare parts, several comnonents wetc re-placed with new units having equivalent or improved' characteristics.

These included the primary flow - AT recorder (which also provides a signal to a new thermal power digital readout), a new regulating rod controller for automatic maintenance of power at a pre-set level, new picoaemeters for channels #7 and #8, digital readonts to replace the vertical indicators for selected cooling system pressures, and mis-cellaneous minor changes. Additional chacges were also initiated and will be accomplished in FY80, such as a naa radiation maaitor multi-point re-corder, simulated period generator for testing and calibrating period channels, and solid-state count rate amplifiers and scalers for the startup channels.

2) Last year's report referred to a program for inspection and repair of the control blade drive mechanisms, required because of a weld failure in one and the subsequent rewelding of that one and five others. Reference was made to two remaining units, then serving as spares; these were re-paired early in FY79, thus completing the program. A more detailed description of the problem and a safety evaluation were given in the FY77 report.

3) Work continued on plans and preparations for the disposal of additional radioactive components removed from the MlTR-I reactor during its modification to MITR-II. The principal remaining component is the old core tank which has been in storage in the spent fuel pool. After review of available and suitable shipping containers, arrangements were "3de with Chem-Nuclear Systems, Inc., for rental of its CNS1 LL-50-100

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cask. In order to .cilize this cask (or any other feasible container), it was necessary t.; tut the old core tank into sections small enough to fit the LL-50-100. An underwater plasma cut ting torch was procured, and the tank was cut c ircumferentially rorialy in half. As,a result of the Three Mile Islard cecident, however, Chem-Nu icar diverted the cask for use at that react 2r, and it is expected that tae cask will not be available to MIT until the coming Fall. The core tank and miscellaneous other components from the MITR-I will be transferred (at least two shipments) to Chem-Nuclear for dir,rosal at a licensed site by burial. Disposal of the tank within a few z., .4 is a necessitv in order to clear the way for the transfer of spent luel from the reactor to the spent fuel pool early in calendar .1980.

4) The auxiliary ingake damper was slightly modified because of a failure to clest an signal during testing (Occurrence Report 50-20/78-3). The problem was caused by sticking of the pivot, which is designed to persit closing of tae dssoa.r by gravi *.y. The pivot was changed in March 1979 to incor-pccat e br anze bushings and teflon thrust washe s and has operated eatisfactorily during vaekly tests since then.

E' rao gaskets related to building containment were renewed during the year.

The gisket on the main exhaust damper began to show signs of west which rJ gh lead to leakage when inspected in April. It was replaced, shown by funpectica and ter' to provide the necessary seal; it was rechecked a weet later as part or the concainment leak test. The gasket on the outer track'2ack door, which had been significantly flattened by more than 20 rears of pressure,was replaced in March 1979. The new gasket, when tested as pa:t of tha containment leak test the following month, was found to be the source of a significant leak. The gasket pressure was increased, and then successfully leak tested separately the following month. Meanwhile.

the inner door provided the required containment integrity.

6) Many other routine 4

.intenance and preventive maintenance jobs were done throughout the year.

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11 E. Section 50.59 Changes, Test and Exoeriments This section contains a description of each change to the facility or procedures and of the conduct of tests and experiments carried out under the conditions of Section 50.59 of 10 CFR 50, together with a sum =ary of the safety evaluation in each case. .

The review and approval of changes in the facility and in the procedures as described in the SAR are documented in the MITR records by means of " Safety Review Forms". These have been paraphrased for this report and are identified on the following pages for ready reference if further information should be required with regard to any item. Pertinent pages in tha SAR have been or are being revised to reflect these changes, and they will be forwarded to the Director of Nuclear Reactor Regulation, USNRC.

The conduct of tests and experiments on the reactor are documented in the experiments and irradiations files. During FY 1978 all experiments have been done in accordance wit!. the descriptions provided in Section 10 of the SAR,

" Experimental Facilities".

343OhJ

12

1. SR # M-78-3 (4/7/78)

Inpta11ation and Use of Hot Cells in the Containment Building As described in last year's report, a hot cell (actually a pair of adjacent hot cells) has Leen designed for installation on the main floor of the containment building. It is to be used for such activities as the inspection and testing of experimental materials and for the transfer of irradiated materials from irradiation capsules to shipping containers.

Last Spring a number of existing shield blocks were modified and a few new ones were fabricated so as to fit properly together and to accommodate viewing windows, manipuletors, ventilation, etc. Assembly of the blocks was not begun until July of this year, but it is now almost complete, and one of the cells is expected to be given its preoperational testing very shortly prior to being placed in operation.

Reactor Staff approval:

Initial: 4/13/78 Final: 8/23/78 MIT Reactor Safeguards Committee approval 9/6/78, subject to Subcommittee review and approval of specific items I

(

13

2) SR # 0-77-13 (7/1/77)

Pneumatic Tube Send Station for Building NW13 This system and the procedures for assuring its safe,t.se were described in last year's report. During FY79 the facility was completed, tested and placed in routine use.

Approvals: listed in FY78 report.

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14

3. SR # 0-79-7 (3/19/79)

Change in D 0 Upper Reflector Temperature Alarm, DT-5 2

A thermocouple, DT-5, was installed in the D 0 upper reflector for the 2

purpose of monitoring its tenperature in comparison with other D 0 and primary 2

system tenperatures during the initial startup and rise to power of MITR-II.

It was found that DT-5 temperatures are within a few degrees of the reflector outlet and core inlet temperatures; this is to be expected due to mixing in the former case and due to the large heat transfer surface formed by the core tank between the H,0 and D 2 0 in the second case. Consequently, on the basis of experience, DT-3 can now be considered redundant and no longer required.

4 DT-5 output was originally recorded on Temperature Recorder No. 3. When maintenance became excessive, DT-5 was transferred to an indicating meter having.an adjustable up-scale alarm. Thus, the alarm feature was retained, but the recording function was not. For the reasons mentioned above, this change is not considered to have any safety significance. Since both the recording and alarm features, however, are mentioned in the SAR, the change is being reported under 10 CFR 50.59.

Reactor Staff approved 4/18/79 3d3 7sN

15

4. SR # E-79-1 (1/18/79)

Replacement of Power Galvanometer with Picoatmeter One of the improvements in reactor instrumentation =ade during the year was the replace =ent of the power galvanometer in the emergency, power channel (#8) with a solid-state, digital-readout picoa==eter. The picoa= meter has its own battery supply, which powers the instrument in the event of loss of normal instru-ment power and also of standby power.

When the picoa==eter was installed, it was decided for two reasons to abandon that part of the sttrtup interlock system which required that the power galvano-meter be switched to maxi =um sensitivity. The picoa=ceter switching circuits do not contain provisions for an interlock and, second, the sensitivity of the lowest scale is such that it is, not normally desirable to use it. Also, the scale desired for startup depends upon the length of the shutdown. Startup checklists have been revised to include a step to set Channel 8 in the lowest usable range.

This is a 10 CFR 50.59 chan t . because SAR Section 7.3 lists the power galvano-meter as part of the startup interlock system and Fig. 7.3a3 shows the interlock.

These references will be deleted in a future SAR change. This represents an improve-ment, not a reduction, in safety.

Reactor Staff approval 1/23/79 343/N O

16 ENVIRONMENTAL SURVEYS F. Environmental surveys, outside the facility, were performed using area monitors. The systems (located approximately in a h mile radius from the reactor site) consist of calibrated G.M. detectors with associated elect-ronics and recorders.

The detectable radiation levels due to Argon-41 are listed below:

Site Julv 1. 1978 - June 30. 1979 North 1.2 mR/ year South 1.0 mR/ year East 2.9 mR/ year West 1.8 mR/ year Creen (East) 0.8 mR/ year

' AVE'IAGE 1.5 mR/ year The ratio of FY 1979 average (1.5 mR/ year) over the FY 1978 average (1.9 mR/ year) reflects a 21% decrease in the measured radiation levels.

This was to be expected, at least in part, since the energy generated decreased by 13.1% (from 941.4 NWD to 817.9 MWD). The remaining 9%

decrease (to give a total of 21%) is probably not significant, since the decrease in the Ar-41 stack release was proportional to the decrease in energy generated within 1%.

343Z74

17 RADIATION EXPOSURES AND SURVEYS WITilIN T'4E FACILITY '

G. A su= mary of radiation exposures received by facility personnel and experimenters is given below:

Whole Body Exposure Range (Rens) Period 7/01/78 - 6/30/79 No. of Personnel No Measurable 96 Measurable - Exposure Less than 0.1 17 0.1 - 0.25 9 0.25 - 0.5 10 0.5 - 0.75 6 0.75 - 1.0 2 1.0 - 2.0 2 TOTAL 142 Suc=ary of the results of radiation and contamination surveys from July 1978 to June 1979:

During the 1978-1979 period, the Reactor Radiation Protection Office continued to provide the routine radiation protection services necessary for full power (5 megawatts) operation of the reactor. The routine services (performed on a daily, weekly, or monthly schedule) include the following:

1. Collection and analysis of air samples taken within the reactor containment shell, and in the exhaust ventilation system.

2. Collection and analysis of water samples taken from the reactor cooling towers, D3 0 system, vaste storage tanks, shield coolant, heat exchangers, luel storage facility, and the primary system.

3. Performance of radiation and contamination surveys, radioactive waste collection, calibration of reactor radiation monitoring systems, and servicing of radiation survey meters.

The results of all surveys described above have been within guide lines established for the facility, o s o W-r;tt. wire it*

s

18 H. RADI0 ACTIVE EFFLUENTS The nature and amounts of radioactive effluents from the MITR during FY79 are su=marized in Table H-la, b and c.

For the activity in liquids rel' eased to the sanitary severage system, the amounts are given on lines 1, 2, 3(a) and 3(b). In calculating concentrations no credit is taken for dilution by non-radioactive waste water from the Nuclear Engineering Building on the reactor facility site or from the remainder of the MIT Cambridge campus, since these are not routinely measured. The volumes of water discharged from the waste tanks and the cooling tower blowdown are measured, however, and are given on lines 3(a) and 3(b). The concentrations for nuclides other than tritium did not exceed 3 x 10-6 pCi/ml, when credit is taken for dilution of the waste tank water by the measured cooling tower

,water, both of which discharge into the sewer at the same point. The nuclides identified were mostly activated corrosion products and are listed on line 2.

The principal gaseous nuclide is Ar-41 from the stack. The annual average concentration as a percent of MPC (61.3%) is down slightly from last year (70.3%)'

because the reactor generated 817.9 MWD compared to 941.4 MUD the previous year for reasons given in Section A-1. The curies per unit of energy generated.: vere almost identical, 10.4 Ci/MRD in FY79 and 10.3 Ci/MRD in FY78.

Other gaseous effluents are reported in the balance of Table H-la and in Table H-lb. The sum of the fractions of MFC add up to approximately 1%. Values are calculated from analyses made of the core purge gas (air flowing across the top of the core tank and through the primary coolant storage tank at 5-6 CFM).

Concentrations here are 1400 times greater than after dilution in'the building exhaust (8500 CFM), and it is possible to detect *lne Kr, Xe, and I nuclides reported in the table. Periodic measuraments :ppear to indicate an equilibrium condition, although the concentrations were higher in May and June 1979 before a fuel element with faulty cladding was identified and removed from the core.

Such measurements are continuing in order to detect any similar trends which may develop.

The activity in solid waste shipments are reported in Table H-lc.

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r- _ _ Table il -la h L~,.3

SUMMARY

OF MITit RADI0 ACTIVE EFFLUENTS i F15 CAL YEAR ).9?j MFC 1978: 1979: 1970-(pCi/ml Unita July . Aug. Sept. Oct. Nov. Dqc.a Jan. Feb. hr2 Apr.  !!}y Jo.ne. Tot s l {}

AeLisity in liquids released to sanitary sewerage system:

1. actal gross 3, excluding 11 (C1) 0.001 0.000( ' NDA } 0.000 NDA NDA 0.000 NDA NDA NDA 0.001 NDA 0.002

2. Jeecific nuclides other than 11 (C1) 0.001 NDA 'NDA NDA NDA NDA NDA NDA NDA NDA 0.001 NDA 0.002 (Cr-51, Na-24, Zn-65 Fe-59 and Hn-54 and Co-60 identified) 3 3(a). H from waste tanks (C1) 0.002 No No 0.011 No No 0.003 No No No 0.03 No 0.046 Average concentration 1 x 10' (x10-4pC1/ml) 7.8 disch, disch.15.3 disch. disch. 7.5 disch, disch. disch.29.7 disch. 16.7(4)

Volume of effluent water (5) 4 (x10 liters) 0.82 0.71 0.33 1.01 2.81 (b)  !! from cooling towers (C1) 4 0.005 0.005 0.004 0.004 0.005 0.005 0.003 0.003 0.002 0.001 0.001 0.001 0.039 Average concentration 1 x 10'1 (8) (x10_ pC1/ml) 0.078 0.071 0.062 0.078 0.087 0.091 0.071 0.090 0.034 0.029 0.026 0.014 0.004 Volume of effluent water (5) (x104 liters) 57.5 65.0 63.1 52.9 57.3 56.3 45.1 36.6 52.4 34.7 46.7 44.0 611.5 Ac_ttvit M ng aseouw wiste:

'I

1. Ar from stack (C1)-a 603 782 650 674 828 616 787 733 619 728 871 605 Average concentration I 8501 4 x 10~a (x10 pel/ml) 3.32 2.39 2.48 2.57 2.53 2.35 2.40 2.46 2.08 2.44 2.34 2.03 2.45 I ##

3 61.3% MN 2(a). 11 f rom stack _7 (C1) 0.60 0.87 0.46 0.45 0.84 0.51 0.55 0.67 0.92 0.82 (x10,33pC1/ml) 2.29 d.65 0.67 8.01 Average Concentration 2 x 10 2.66 1.77 1,72 2.55 1.95 1.97 1.85 2.26 2.23 2.46 2.75 2.21 0.011 lice (b). 11 f rom cooling tower ,7 (C1) 0.020 0.018 0.016 0.016 0.016 0.019 0.011 0.018 Average Cancentaction 2 x 10 (x10,33pC1/ml) 14.9 10.w 10.9 11.1 9.9 14.7 7.4 15.1' 0.007 0.004 0.000 0.003 6.1 3.8 3.5 2.5 0.154 9.21 0.05% Nec

.otes: (1) 10CFR20 (2) 0.000 indicates less than 0.0005 C1. (3) NDA - No Detectable Activity (4) Weighted Average of individual discharges.

(5) Does not include other diluent from MIT estimated at 2.7 million gals / day. (6) Average concentrations of gaseous wastes include authorized dilution factor of 3000. (7) Fiscal year totals are averaged over 12 months for gaseous releases.

(8) Technical Specification 3.8-1.b limits cooling tower concentration to 1 x 10-3 pC1/ml.

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Table Il-lb g!MMARY OF HITR RADIDACTIVE EFFLUErfrS FISCAL YEAR 1979 Activity in Caseous Uante Estimates of annual releaccs from stack for other nuclides based on representative samples:

Average Conc.

Nuclide MPC (uCi/ml) (uCi /ml) 7. MPC Curies

-8 ~11 3 (a) Br 3 x 10 0.047 x 10 0.0016, 0.17 Om (b) ur 0.01 x 10' O.002 x 10' O.02 0.006

~ ~

(c) Br 4 x 10 0.001 x 10 0.00003 0.004

~

4 (a) "Kr 10 x 10" 2.75 x 10 0.028 10.1

-8 (b) Kr 2 x 10 5.63 x 10' O.282 20.7

-8 (c) Kr 2 x 10 6.29 x 10' O.315 23.1

-11 5 (a) Xe 30 x 10' 5.05 x 10 0.017 18.5

~

(b) *Xe 30 x 10 2.04 x 10" 0.007 7.5 135 ~

(c) Xe 10 x 10 3.78 x 10" 0.038 13.9

~

(d) *X c. 3 x 10' 1.44 x 10 0.048 5.3

-8 ~

(c) Xe 3 x 10 9.99 x 10 " 0.333 36.7 (f) 1 0.01 x 10" 0.008 x 10"I 0.08 0.00004 TOTAL 1.17 .

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Tabic II-1c SIRCtARY OF MITR RADIOACTIVE SOI.ID WASTE SilIINENTS FISCAT, YEAR 1979

1978 1979 linits August March Totc1

1. Solid waste packaged (Cu.Ft.) 113 199* 312 2 Total activity (irradiated components,ioneghange resins, etc.) Co (C1) 0.057 0.108 0.165 51 65 Cr, 55-59Fe, 7 , ,

3.

(a) Dates of shipment 8/25 3/20 2 shipments (b) Disposition To licensee for Same as in burial August CC 4a c:a EJ w si ~

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