ML20150C695

Rept of Reactor Operations,1987

ML20150C695
Person / Time
Site:	University of Michigan
Issue date:	12/31/1987
From:	Kerr W MICHIGAN, UNIV. OF, ANN ARBOR, MI
To:	Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 8803210179
Download: ML20150C695 (22)

Text

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1. f A 1 REFORT 0F REACTOR OPERATIONS

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January 1. 1987'to DecemDer'31, 1987 1

FORD NUCLEAR REACTOR HICHIGAN MEMORIAL - PHOEHlX PROJECT THE UNIVERSITY OF MICHIGAN ANN ARBOR Marcn 1988 l

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ABSTRACT Technical Specifications f or the Ford Nuclear Reactor (FNR) require the annual submission of this review of reactor operations to the

-U. S. Nuc l ear Regulatory -Commis s ion - (NRC ) .

.The11967 reactor schedule of_ ten days of continuous operation at 11 censed power of two megawatts followed by four days of-shutdown resulted in 5773.3 reactor operating hours, 5247.9 operating hours at full power, 9721.8 accumulated megawatt hours, and an overall reactor availability of 66 percent for the calendar year.

Seven regular fuel. elements, two contrcl rod fuel elements, and 232 pounds of heavy water were required for operation. Twenty-four new fuel elements were received during 1987. There were two spent fuel shipments during the year with a total of 46 high enriched uranium (HEU) elements. With these shipments, all HEU f uel was removed f rom the facility.

There were no reportable occurrences in 1987. Regular maintenance, surveillance tests, and experiments were completed in a normal manner.

There were 17 unscheduled reactor shutdowns during 1987.

There were no radioactive effluent releases aDove 10CFR20 1)mits.

The maximum whole body exposure received by an Indivioual at the facility was 1.520 rem.

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FORD HUCLEAR REACTOR Docket No. 50-2 License.No. R-26 HEPORT10F REACTOR OPERATIONS

' January 1, 1967 - December 31, 1987 This report reviews the operation of the University of Michigan's Ford Nuclear Reactor for.the period January i to December 31, 1987.

The report is to meet the requirement of Technical Specifications for the Ford. Nuclear Reactor. The format for the sections that f ollow conf orms to Section 6.6. (1) of Technical Specif ications.

The Ford Nuclear Reactor is operated by the Michigan Memorial-Phoenix Project of the University of Michigan. The Pro.)ect, established in 1949.as a memorial to students and alumni of the University who died in World War II, encourages and supports research on the peacef ul uses of nuclear energy and its social implications, in addition to the Ford Nuclear Reactor (FHR), the Project operates the Phoenix Memorial Laboratory (PML). These laboratories, together with a f aculty research grant program, are the means by which the Project carries out its purpose.

During 1967, as in previous years, the operation of the Ford Nuclear Reactor has provided major assistance to a wide variety of research and educational programs. The reactor provides neutron irradiation services and neutron beamport experimental f acilities f or use by faculty, students, and researchers from tne University of Michigan, other universities, and Industrial research organizations. Reactor staff members teach classes related to nuclear reactors and the Ford Nuclear Reactor in particular and assist in reactor-related laboratories.

Tours are provided for school children, university students, and the public at large as part of a public education program. During 1987.

569 peop.e participated in 58 tours. .

The operating schedule of the reactor enables a sustained high level of participation by research groups. Continued support by the l Department of Energy through the Univers2ty Research Reactor Assistance Program (Contract No. DE-ACO2-7 6ER003e51 and the Reactor

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Facility Cost Sharing Program (Contract No. DE-FG02-80ER10724) has l been essential to maintalning operation of the reactor f acility.

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1. - OPERATIONS

SUMMARY

in January, 1966, a continuous operating cycle was adopted for the Ford Nuclear. Reactor at its licensed power level of.two megawatts. The cycle consisted of approximately 25 days at full power followed by three' days of shutdown maintenance. In June, 1975, a reduced operating cycle consisting of ten days at full power f ollowed by f our days of shutdown maintenance was adopted. A' typical week consisted of 120 full-power operating hours. In July, 1983, the reactor operating schedule was changed to Monday through Friday at licensed power and weekend shutdowns. Periodic maintenance weeks were scheduled during the year. In January, 1985,- the cycle consisting of ff our days or 96 f ull-power operating hours per week at licensed power fo!! owed by three days of shutdown maintenance was restored in order to eliminate the periodic shutdown maintenance weeks needed.in the previous cycle.

Beginning July 1, 1987, the reactor operating cycle returned to ten day operation at full power followed by four days of shutdown maintenance. Calendar year 1987 began with cycle 272 and ended with cycle 284. A cycle covers f our weeks; two of the ten day - four day sequences.

The reactor operates at a maximum power level of two mega whicDproducesapeakthermalfluxofapproximately3x10yatts n/cmd /sec. An equilibrium core configuration consists of approximately 40, 19.75% enrichment, plate type fuel elements.

Standard elements contain 167 gm of U235 in 18 aluminum clad fuel plates. Control elements, which have control rod guide channels, have nine plates and-contain 83 gm of U235. Overall active fuel element dimensions are approximately 3"x3"x24".

Standard fuel elements are retired after burnup levels of approximately 20-30X are reached. Control elements are retired annually at burnup levels of approximately 25-35X.

Fuel burnup rate is approximately 2.46 gm U235/ day at two megawatts.

1.1 Facility Design Changes There were no facility design changes in 1987.

1.2 Equipment and Fuel Performance Characteristics Reactor equipment and tuel exhibited no abnormal characteristics during 1987. Replacement of retired fuel elements resulted in an annual use of approximately seven standard elements and two control elements.

During 1987, 24 new fuel elements were received.

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There were two spent fuel shipments in 1987 with a total of 46 elements. With these shipments, all high enrichment uranium (HEU) f uel has been removed f rom the site.

In 1987, the use of the hel.'y water ref lector tank on one f ace of the reactor core required a throughput of 232 pounds of heavy water. Fresh heavy water is used to replace heavy water in the tank in order to maintain a tritium level in the tank not greater than the 50 curie limit imposed by Technical Specifications.

1.3 Safety Related Operating Procedure Changes There were no safety related operating procedure changes made in 1987.

1.4 Maintenance, Surveillance Tests, and Inspection Results as Required by Technical Specifications.

Maintenance, surveillance tests, and inspections required by Technical Specifications were completed at the prescribed intervals. Procedures, data sheets, and a maintenance schedule / record provide documentation.

1.5 Summary of Changes, Tests, and Experiments for Which NRC Authori2ation was Required.

None 1.6 Operating Staff Changes On June 1, 1987, the f acility staf f was reorganized. Reed R.

l Burn was appointed Nuclear Reactor !.aboratory Manager with i

responsibility for the Ford Nuclear Reactor and the Phoenix l Menorial Laboratory. Gary M. Cook was appointed Assistant l Manager for Operations, and Philip A. Simpson was appointed Assistant Manager for Research Support Activities.

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The following additional reactor operations staff changes occurred June 1, 1987:

Name From To Stephen E. Nowak Lead Reactor Operator Relief Operator Robert E. Touchberry Reactor Operator Lead Reactor Operator Page-5 f .

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Three~new reactor operators were hired in 1987:

Name Date of Hire John C. Pulford March 16, 1987 Andrea E. Cross June 1, 1987 Robert M. Thebert June 1, 1987 Radiation Control. Service 13 rotating the services of the healt.h physics staff. In 1987, the health physicist and health physics technician were rotated as f ollows:

Name Start date Departure Date Mark Driscoll February, 1987 P.lilip Keavy May, 1987 Kcnneth Conway January, 1987 Jeffery Hadley May. 1987 1.7 Reportable Occurrences There were no reportable occurrences during 1987.

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2. POWER GENERATION

SUMMARY

The following table summarizes reactor power generation for 1987.

Full Power Operating Operating Megawatt Peretnt Cycle inclusive Dates Hours Hours Hours _ Availability 272 1/4/87-1/31/87 363.3 254.5 555.0 54 273 2/1/87-2/28/87 363.5 326.8 662.2 54 274 3/1/87-3/28/87 386.4 333.0 676.8 58

.275 3/29/87-4/25/87 380.1 334.1 680.7 57 276 4/26/87-5/23/87 346.1 320.4 665.3 52 277 5/24/87-6/20/87 370.1 320.4 65:1.4 55 278 6/21/87-7/18/87 426.6 350.1 740.2 63 279 7/19/87-8/15/87 470.4 448.1 903.2 70 280 8/16/87-9/5/87 300.9 288.7 584.7 45 281 9/6/87-10/03/87 '475.8 463.9 932.8 71 282 10/04/87-10/31/87 477.3 446.9 903.3 71 l

283 11/1/87-11/28/87 456.0 437.2 879.4 68 284 11/29/87-1/2/88 507.4 486.9 986.1 76 l 5773.3 5247.9 9721.8 66 l

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3. UNSCHEDULED REACTOR SHUTDOWN

SUMMARY

The following table summarizes unscheduled reactor shutdowns for 1987.

3.1 Unscheduled Shutdowns Total Unscheduled Shutdowns 17 Operating Hours per Shutdown 239.6 3.2 Shutdown Types l Single Rod Drop....................... 0 Multiple Rod Drop (NAR)......... ..... 3 Operator Action....................... 2 Operator Error........................ 3 Process Equipment..................... 3 Reactor Controls... .................. 4 Electric Power Failure................ 2 l 3.3 Shutdown Type Definitions Single Rod Drop and Multiple Rod Drop (NAR)

An unscheduled shutdown caused by the release of one or more of the reactor shim-safety rods from its electromagnet, and f or which at the time of the rod j

release, no specific component malf unction can be Ident1 fled as having caused the release.

l Operator Action A condition exists (usually some minor difficulty with l

an experiment) for which the operator on duty judges that shutdown of the reactor is required until the j dif f iculty is corrected.

1 l Operator Error The operator on duty makes a judgement or manipulative error which results in shutdown of the reactor.

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f a Process Equipment Shutdown caused by malfunctions in the process equipment interlocks of the reactor control system.

Reactor Controls Shutdown initiated by malfunctions of the control and detection equipment directly associated with the reactor safety and control system.

Electrical Power Failure Shutdown caused by Interruptions in the electrical power supply of the reactor f acility.

3.4 Cycle Summary of Unscheduled Shutdowns Cycle 272 There were three unscheduled shutdowns.during cycle 272. Two of these were single rod drops from "A" level safety channel. The cause was determined to be in the linear level recorder. For some undetermined reason, the recorder circuitry or drive system would intermittently not follow small signal changes in the upward direction. The servo-controller which received its deviation signal from the linear level system would see a continuous low power and in turn call for a raise of the control rod. This caused reactor power to increase. When the first scram occurred, it was not clear to the operators what had caused the power

! to spike up. The restart checks were all satisfactory, and the-reactor was restarted. A short l

time later, as one operator was loading a sample and l monitoring the control rod reaction, the reactor scrammed again. The reactor remained shutdown, and a systematic check and replacement of the components in the linear level recorder were completed. There were no Indications that any of the recorder electronic components were f aulty. The mechanical parts of the recorder were also checked for wear and binding.

l Nothing was found. The reactor was restarted without i dif ficulty. A new servo-controller which does not

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depend on the recorder for a dev13 tion signal was l ordered and installed.

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C~ s The third unscheduled shutdown was traced to a fault in the wiring-in the "Lrldge Not Clamped" scram circuit. By design, any fault to ground will cause a

-scram. In this case the' fault was traced to a loose bulb socket. The reactor bridge supports the core structure and must remain clamped in place during operation at full power.

Cycle 273 There was one unscheduled shutdown during cycle 273.

The shutdown was a single rod drop for no apparent reason. Upon investigation, the operations staf f found that they-could simulate a similar rod drop by momentarliy opening the ground loop between components of the Log H neutron detector system. It is possible that this was the cause for a number of the rod drops that occurred during the last few months. A permanent ground strap was installed with the hope of eliminating the problem.

Cycle 274 There were four unscheduled shutdowns during cycle 274. Three of these were rod drops for no apparent reason. During the maintenance period the operations staff looked carefully at the system and was able to trace the problem to a faulty ground loop. The saf ety system components have a safety Interlock circuit that will turn off magnet current to the rods if one of the components is disconnected. This circuit uses the chassis ground connections as one leg of the loop. It was discovered that at least one of the drawers in the l

Safety circuit did not always make a positive ground

' connection when it was fastened in place. To ellmlnate the possibility of f uture dif f iculty, a braided ground strap was run to each of the components in the safety system and securely grounded. There have been no further problems with this circult.

The last unscheduled shutdown was an automatic rod l Insertior. (auto-rundown ) from the linear level

recorder. The linear level detector has been erratic and it is believed that the problem was caused by a short term spike in the signal. The detector has since been replaced and repaired.

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l 4 a Cycle 275-There were two unscheduled shutdowns during cycle 275.

The first occurred.when the reactor building high radiation. alarm was tripped inadvertently. A. member.

of the operations'staf f.was checking the radiatlon

-level of a' sample from the reactor pool. The sample was lowered back into the pool as soon as the alarm sounded. The reactor was returned to power immediately. The second unscheduled shutdown was an operator action when he noticed the linear level recorder behaving in an erratic manner. The operations crew had been aware for some time that the linear level chamber had been showing signs of intermittent failure. See Cycle 274 summary. The chamber was replaced with a spare unit and the system has not caused further problems. A permanent replacement chamber was purchased and installed.

Cycle 278 There was one unscheduled shutdown during cycle 278.

This shutdown was caused by a temporary loss of offsite electrical power during a thunderstorm. All reactor systems were checked and the reactor returned to power without difficulty.

Cycle 279 There was one unscheduled shutdown during cycle 279.

The shutdown occurred when an operator, during a routine check of the reactor building exhaust i radiation monitor, inadvertently initiated a building alarm. The building alarm alarms and shuts the reactor down. The alarm was immediately recognized as being f alse and was silenced. This error occurred when an attempt to reset the alarm test was made before the i

l test signal decayed below the alarm setpoint. The operator involved was re-instructed as to the proper procedure. The reactor was restarted without dif f icult/.

Cycle 280 There were three unscheduled shutdowns during cycle 280. The f1rst was an operator action when, during a normal check, a member of the operations crew noticed a small piece of unknown material on the reactor core, i

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,J The reactor was returned to full power. The second unscheduled. shutdown was'due to a momentary electrical

' power interruption (cause unknown). The reactor was returned to full power immediately. The third unscheduled shutdown was caused by the failure of a bulb socket in the "Header Up/No Water Flow" scram circult. This circuit is designed such that if a scram' condition is reached, current f low is induced in the circuit which in turn is sensed by the safety system which then turns off the magnet power. In this instance the Indicating bulb socket shorted to ground causing a current flow which caused the scram to occur. The operations crew replaced the bulb socket which cleared the problem. The reactor was returned to power All of the bulb sockets in the scram circuit loop.were then replaced during the next reactor maintenance period.

Cycle 282 There were two unscheduled shutdowns during cycle 282.

The f1rst was an auto-rundown from the temperature recorder for no apparent reason. The recorder indicated one point approximately 100 F above the normal trace. The signal returned to its normal value on the next cycle. (The temperature recorder cycles

! through all points at one minute intervals.) The recorder was checked and nothing was f ound to be abnormal. The problem has not reoccurred. The reactor power was reduced for approximately 5 minutes after which the operator returned to two megawatts.

The second unscheduled shutdown occurred when an experimenter inadvertently shorted the primary flow signal which tripped the "High Power /No Water Flow" l

Scram. The experimenter was in the process of calibrating the inputs for computer readouts and apparently shorted the primary flow contacts which in

l. turn caused the shutdown. In the future this process will not be allowed while the reactor is at power.

The reactor was returned to two megawatts without dif f iculty.

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6. RADIOACTIVE EFFLUENT RELEASE Quantitles and types of radioactive effluent releases, environmental monitoring locations and data, and occupational personnel radiation exposures are provided in this section.

6.1 Gaseous Effluents Gaseous effluent concentrations are averaged over a period of one year.

Quantity Unit

a. Total gross radioactivity. 57.2 C1 D. Average concentration released during normal steady state reactcr operation. 1.09E-7 pC1/cc

c. Average release rate during normal steady state reactor operation. 1.79 UC1/sec

d. Maximum Instantaneous concentration during special operations, tests, Not and experiments. Applicable pC1/cc

e. Percent of 41Ar MPC (4.0x10-8 pC1/cc) without dilution factor. 271 Percent

f. Percent of 41Ar MPC with 400 dilution factor. 0.678 Percent 6.2 lodine Releases

a. Total lodine radioactivity by nuclide based upon a representative isotopic analysis. (Required if lodine is identified in primary coolant samples or if fueled experiments are conducted at the facility). None pC1/ml Page-13

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b. 131I odine releases related to steady state reactor operation

-(Sample C-3, main reactor exhaust.

stack).

(1) Total lodine released. 598 pC1 (2) Average concentration released. 1.70x10-12 pC1/cc (3 ) Percent of 131 1 MPC (1.0x10-10 pC1/cc) without

' dilution factor. 1.70 Percent (4) Percent of 131 1 MPC with 400 dilution factor. 4.25x10-3 Percent C. lodine releases related to combined steady state reactor operation and radiation laboratory activities (Sample C-2, combined secondary reactor exhaust and partial radiation laboratory exhaust).

(1) Total lodine released. 2.21 mC 1.

(2) Average concentration released 1.26E-ii pC1/cc (3) Percent of 131 1 MPC without dilution factor. 12.8 Percent (4) Percent of 131 1 MPC with 400 dilution factor. .032 Percent

d. 125 1 and 131 1 releases not related to reactor operations (Samples C-4 + C C-2, radiation laboratory exhaust stacks).

(1) Total lodine released. 43.5 mC1 (2) Average concentration released. 1.35E-10 pC1/cc (3 ) Percent of combined 125 1 and 131 1 MPCs (8.0x10-11 and ,

1.0x10-10 pC1/cc, respec-tively) without dilution f actor 137 Percent (4) Percent of 125 I and 131 1 MPCs 0.34 Percent with 400 dilution factor.

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t-F 6.3 Particulate Releases Gross alpha activity is required to be measured if.the:

operational or experimental program could. result in'the-release of alpha emitters.

Quantity Unit

a. ' Total gross beta gamma radioactivity. 110.0 pC 1-b.- Gross' alpha radioactivity. Not required

c. Total gross radicactivity of nuclides with half lives greater

> than eight days, 1.30E-13 pC1/cc

d. Percent of MPC (1.0x10-10 UC/cc) for part-1culate radioactivlty with half lives greater than eight days without dilution factor. .130 Percent

e. Percent of MPC for particulate radioactivity with half lives greater than eight dtys with 400 dilution factor. 3.25x10-4 Percent 6.4 Liquid Effluents Total alpha radioactivity is required to be measured if the operational or experimental program could result in the release of alpha emitters. Release concentrations are averaged over the period of the release.

a. Total gross beta-gamma radioactivity. ~3 64. 7 mC1 (1) Distilled 3H . . . . . . . . . . . . 8 6 . 7 2 mC 1 (2) Evaporated gross beta-gamma.............. 278.0 mC1

b. Average gross beta-gamma concentration. 7.192x10-4 pC1/mi (1) Distilled 3H............ 1.703x10-4 pC1/ml (2) Evaporated gross 5.489x10-4 pC1/ml beta-gamma..............

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-Quantity' Unit

c. Maximum gross beta gamma concentration. 7.409x10-3 yC1/ml (1) Distilled 3H ............ 2.077xiO-3 pC1/ml (2) Evaporated gross beta gamma.............. 5.332x10-3 pC1/mi-

d. Total alpha radioactivity. Not required

e. Average alpha concentration. Not required

f. Total liquid waste volume. 5.06E8 mi

g. Total dilution water volume added prior to release f rom the f acility. O mi

h. Approximate total dilution water volume of North Campus sanitary sewer. 1.0x10+12 ml

1. Isotopic analysis and average isotopic concentration of waste storage tanks with average gross beta gamma concentrations in excess of 9.0x10-4 pC1/ml. Not required J. Percent of MPC (3.0xiO~4 pC1/ml) without dilution based on gross beta gamma radioactivity leaving the retention tank. 183.0 Percent X. Percent of MPC with 300 dilution factor based on gross beta-gamma radioactivity leaving university property. 0.61 Percent

1. Percent of combined MPCs without dilut:on at the point of release from the retention tanks into the building sewer system based on distilled 3 H + gross beta gamma radioactivity (0.1 and 3.0x10~4

, PC1/ml respectively). 183.2 Percent

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4 Quantity Unit

m. Percent of combined MPCs with 300 dilution factor based on distilled 3H+ gross beta-gamma radioactiv-Ity leaving university property .6107 Percent 6.5 Environmental Monitoring The environmental monitoring program for the Ford Nuclear Reactor f acility consists of direct radiation monitors (TLD) and air sampling stations located around the f acility and selected water and sewer sampling stations,

a. TLD Environmental Monitors The TLDs located at stations in the vicinity of the facility are collected at two month intervals and sent to a commercial dosimetry company for analysis. At the submittal date of this report, TLD data through November 30, 1987 were available.

Station Description Reading Unit Highest station (absolute) 21 mrem /2 mo Lowest station (absolute) 3 mrem /2 mo Average of all stations 9.72 mrem /2 mo Average background 8.58 mram/2 mo Average annual dose 0 mrem / year (above background)

The average annual dose is determined by subtracting background from the average of all stations and multiplying by six since the measured doses are for two month periods. Background is taken at a distance in excess of three miles from the reactor. As no station was statistically greater than background, the average annual dose is reported as zero (0).

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b. Dust Samples Four air grab samples are collected weekly from continuously operating monitors located to the north (Northwood Apartments), aast (Industrial and Operating Engineering), south (Institute of Science and Technology), and west (Chrysler Center) of the reactor f acility. Each millipore filter sample is counted for net beta activity.

Approximate radioactivity concentrations are 1.64x10'4 pC1/cc, which is 0.02% of MPC.

c. Water Samples A daily 100 ml tap water sample is collected at the University of Michigan School of Public Health, giving a composite monthly sample of 3,000 ml. Near the first of each month, a 3,000 mi grab sample of river water is collected above Ann Arbor (Pump Station). Near the fifteenth of each month, 3,000 ml grab samples of river water are collected below the sewage treatment plant (Superior Township) and at the Dixboro bridge (Dixboro).

Each of the four samples is oven-dried onto a '

planchet for net beta analysis. Approxim radioactivity concentrations are 7.45x10~ ate 7pC1/ml, which is .25% of MPC.

d. Sewage Samples Ann Arbor sewage plant personnel collect two 100 ml samples daily; one raw and one treated sewage.

Two composite 1,500 mi samples are picked up f rom the sewage plant on the 15th and 30th of the month for analysis. Each sample is oven-dried onto a planchet for net beta analysis. The average radioactivity concentration measured in 1986 was 2.26E-8 pC1/ml, which is 0.75% of MPC.

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! e. Number of locations at which levels were found to be significantly higher than the remaining '

locations.

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f. Highest, lowest, and average concentrations or radiation levels for the sampling point with the highest average.  ;

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TLD Highest Lowest Average Unit-Inst'itute of- 8.5 0.5 4.25' mrem above Science and background Technology 2 months.

DUST Research 5.68E-13 3.1x10-15 2.05E-14 pC1/cc Administration i

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Dixboro 2.55E-8 6.07E-9 8.44E-9 pC1/ml Bridge SEWAGE - Ann Arbor Waste Water Treatment Plant Raw Sewage 3.15E-8 1.72E-8 2.28E-8 pC1/ml Treated Sewage 4.12E-8 1.39E-8 2.23E-8 pC1/mi l

g. Maximum Cumulative Radiation Dose The maximum cumulative radiation dose which could have been received by an individual continuously present in an unrestricted area during reactor operations from direct radiation exposure, exposure to gaseous effluents, and exposure to

! 11guld ef fluents:

l (1) Direct radiation exposure to such an l Individual is negligible since a survey of accessible areas around the reactor building l

Shows no detectable radiation dose rates above backgrcund.

(2) Gaseous Effluents The maximum dose-equivalent isbaseduppg1 l

anassumedcont}Dyousexposuretothe (2640 pC1) and J21 (44,000 pC1) released from all sources at the PML/FNR

! facility. The total air volume released from all stacks was 8.45E14 cc. From this, combined effluent concentrations of l -

3.70E-12UC1/ccpgd5.21{3{ipC1/ccare calculated f or 1 1 and 1, respectively.

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-L- ; *-A The applicable MPCs for 125 1 and 131 1 releasesinairtounrestric}gdareasare 8.0x10-11 UC1/cc and 1.0x10- pC1/cc.

Combined fractional MPC for the two isotopes is 55.9% without-allowance for dilution.

Applying the allowable dilution factor of 400, gaseous effluent concentration Is .140X of MPC.

Continual exposure to 1.0 MPC of 131 1 for a period of one year would yield a thyroid dose-equivalent of 1.0 rem. Thus, the maximum dose equivalent to an Individual from gaseous effluents in 198 may be calculated to be 1.40 mrem. This calculation pqqsumes 100% o the gaseous eff!uent to be IsII.

Actual .aaximum theoretic 1' dose equivalent wouldbesomewha}2gowerdtetothe contribution of I t o t'.le t o t a l .

(3) Liquid Effluent The annual dose from 11guld effluents is zero. The most likely source of exposure would be an external dose to sewage treatment personnel in the vicinity of raw sewa8e. The radioactivity ccncentrations measured at the Ann Arbor sewage plant yleid no exposure rate above background using a survey Instrument,

h. If levels of radioactive materials in environmental media, and determined by an environmental monitoring program, indicate the likelihood cf pub!1c intake in excess of 1X of those that could result from continuous exposure ,

to the concentration values listed in Appendix B, Table 1, 10CFR20, estimate the likely i

resultant exposure to individuals and to population groups and the assumptions upon which those estimates are based.

l Not Applicable.

6.6 Occupational Personnel Radiation Exposures The total number of facility operational personnel for l whom personnel monitoring was provided was 226 whole l body and 85 extremities. (Individuals for whom extremity monitoring was provided received TLD ring l dosimeters f or each hand). No radiation exposures t

greater thar 50 mrem were received at the facility by individuals under the age of 18.

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/ i A summary of whole body and extremity exposures based upon data f rom January 1 through December 31, 1987 is as f ollows:

Estimated Whole Body Number of Individuals Exposure Range frem) in Each Range No measureable exposure....................... 122 Measureable exposure, less than 0.10................................. 78 0.10 - 0.25.................................... 13 0.25 - 0.50...................... .............. 3 0.50 - 0.75..................................... 0 0.75 - 1,00..................................... 3 1.00 - 2.00..................................... 7 Greater than 2.00............................... O Total 226 (Maximum exposure: 1.520 ret)

Estimated Extremity Exposure Range (rem) Humber of Individuals (Maximum of Either Hand) in Each Range No measureable exposure.. ......... ....... . .26 Measureable exposure, less than 0.10. .. . ............ ............ 18 0.10 - 1.00....... ............... . .......... 23 1.00 - P.00..... ........ ......... ... ....... 13 2.00 - 3.00..... ...... .............. ....... ,4 3.00 - 4.00.. ........ ................... ... 0 4.00 - 5,00............................. ..... .0 5.00 - 6.00.................................. .. 0 6.00 - 7.00........... ......................... 1 Total 85 (Maximum exposure: 6.970 rem)

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J o e . r Tile UNIVERSITY OF MICllIGAN Michigan Memorial-Phoenix Project Office of the Director Ann Arbor, Michigan 48109-2100 (313) 764-6213 FRICnliy E0 Jing I 5I])GitiI w-2%rp

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m .lish U.S. Nuclear Regulatory Commission UKbd D Region III Attn: A. Bert Davis 4 Regional Administrator

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799 Roosevelt Road .

Glen Ellyn, Illinois 60137 Dear Sir; The enclosed REPORT ON REACTOR OPERATIONS for the period January 1, 1987 to December 31, 1987 is submitted to comply with Section 6.6 of the Ford Nuclear Reactor Technical Specifications.

Sincerely, Wh-William Kerr Director Michigan Memorial-Phoenix Project xc: USNRC, Washington, D.C.

FNR Safety Review Committee FNR llealth Physicist Enclosure (1) i i pg o en g

)