ML19305E512
| ML19305E512 | |
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| Site: | University of Michigan |
| Issue date: | 03/31/1980 |
| From: | MICHIGAN, UNIV. OF, ANN ARBOR, MI |
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Text
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8005200 OO 7 O
l MICHIGAN MEMORIAL PHOENIX PROJECT l
THE UNIVERSITY OF MICHIGAN REPORT ON REACTOR OPERATIONS January 1, 1979 to December 31, 1979 f4
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FORD NUCLEAR REACTOR
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ABSTRACT Technical Specifications for the Ford Nuclear Reactor (FNR) require the annual submission of this review of reactor operations to the U.S. Nuclear Regulatory Commission (NRC).
The 1979 reactor operating schedule of 10 days at licensed power followed by four days of shutdown and maintenance resulted in 5,829.5 reactor operating hours, 5,196.4 operating hours at full power, 10,392.8 accumulated megawatt hours, and an overall reactor availability of 67 percent for the calendar year.
Seventeen fuel elements and 204 pounds of heavy water were required for operation.
No unusual events occurred while operating the reactor during 1979.
Regular maintenance, surveillance tests, and experiments were completed in a normal manner.
Three reactor operators resigned and were replaced.
There were 28 unscheduled reactor shutdowns in 1979.
There were no radioactive effluent releases above 10 CFR 20 limits.
The maximum whole body exposure received by an individual at the facility was 0.87 rem.
ii
TABLE OF CONTENTS Page TITLE PAGE i
ABSTRACT 11 TABLE OF CONTENTS iii INTRODUCTION 1
- 1. OPERATIONS
SUMMARY
2 1.1 Facility Design Changes 2
1.2 Equipment and Fuel Performance Characteristics 3
1.3 Safety Related Operating Procedure Changes 3
1.4 Maintenance, Surveillance Tests and Inspection Results as Required by Technical Specifications 3
1.5 Summary of Changes, Tests, and Experiments for Which NRC Authorization was Required 3
1.6 Operating Staff Changes 4
1.7 Reportable Occurrences 4
- 2. POWER GENERATION
SUMMARY
5
- 3. UNSCHEDULED REACTOR SHUTDOWNS
SUMMARY
6
- 4. CORRECTIVE MAINTENANCE ON SAFETY-RELATED SYSTEMS AND COMPONENTS 12
- 5. CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR NRC APPROVAL 12
- 6. RADIOACTIVE EFFLUENT RELEASES 13 6.1 Gaseous Effluents 13 6.2 Iodine Releases 13 6.3 Particulate Releases 14 6.4 Liquid Effluents 14 6.5 Environmental Monitoring.
15 6.6 Occupational Personnel Radiation Exposures 18 iii
EDPD ICCIEAR RE7CIOR Docket 50-2 License R-28 REPCRP O! FDCIOR OPERATICNS January 1, 1979 - December 31, 1979
'Ihis report reviews the operaticn of the Lhiversity of Michigan's Ford Nuclear Reactor for the period January 1 to December 31, 1979. The report is to meet the requirement of Technical Specifications for the Ford Nuclear Reactor. The for:nat for the sections that follow confones to Section 6.6.a of Technical Specificaticns.
The Ford Nuclear Feactor is operated by the Michigan Memorial-Phoenix Project of the University of Michigan. The Project, established in 1948 as a memorial to students and alumni of the University who died in World War II, encourages and supports research on the peaceful uses of nuclear energy and its social inplicaticns. In additicn to the Ford Nuclear Reactor (FNR),
the Project operates the Phoenix Memorial laboratory (PML). These laboratories, together with a faculty research grant program, are the means by which the Project carries out its purposes.
During 1979, as in previous years, the operation of the Ford Nuclear Paactor has provided major assistance to a wide variety of research and educational programs. 'Ihe ENR provides neutrcn irradiation services and neutron beam port experimental facilities for use by faculty, students, and researchers frcm the University of Michigan, other universities, and industrial research organizaticns. ENR staff mrbers teach classes related to nuclear reactors and the EUR 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 educaticn program. During 1979, approximately 2,862 pecple participated in 154 tours.
The operating schedule of the reactor enables a sustained high level of participation by research groups. Ccntinued support by the Department of Energy through the University Research Reactor Assistance Program (Contract No.
EY-76-C-02-0385) and the Reactor Facility Cost Sharing Prcgram (Ccntract No.
EY-76-C-02-2117) has been essential to maintaining operation of the reactor facility.
_1_
1.
TEPATICES SDNAKI In January,1966, a ecntinuous cperating cycle was adopted for the ENR at its licensed power level of 2 megawatts. The cycle consisted of approximately 25 days at full power followed by three days of shutdown maintenance. In Juns,1975, a reduced operating cycle consisting of 10 days at full power followed by fcur days of shutdown maintenance was i
adopted. A typical week ccnsists of 120 full power operating hours.
In March,1977, reactor power was reduced to 1 megawatt in order to ccnserwn fuel. In June, 1978, operating pcwer was returned to 2 megawatts.
For maintenance and record keeping purposes, a cycle consistr of 28 days.
4
'Iwo 14-day operating cycles ccr@ose ene 28-day maintenance cycle.
Calendar year 1979 began with Cycle 168 and ended with Cycle 180.
The reactor operates at a maximum power level of 2 MW which produces 13 2
a peak flux of approximately 5 x 10 n/cm /sec. A typical core l
configuration consists of 35-40, 93% enrichrent, plate type fuel elements. Standard elements ccntain 140 gm of U-235 in 18 aluminum clad fuel plates. Control elements, which have a control rod guida channel, have nine plates and ccntain 70 gm of U-235. Overall fuel element dimensions are approximately 3" x 3" x 24".
Standard fuel elements are retired after burnup levels of approximately 17% are reached. Control elements are retired after burnup levels of approximately 35%. Fuel burnup rate is approxirately 2.46 grn/ day at 2MN.
1.1 Facility Design Changes 1.1.1 Reactor Overhead Crane
'Ihe reactor crane was upgraded frcan 10 to 15 tcns j
capacity. In addition to the installaticn of a ccamercially available 15 ten crane and hoist, the l
"I" beam support structure was upgraded with welded angle supports as reccmended by a civil engineering censultant. A safety analysis of the installation was performed. '
1.2 Equignent and Fuel Perforrance Characteristics Reactor equipment and fuel exhibited no abnor:nal characteristics during 1979. Peplacemnt of retired fuel elements resulted in an annual use of apprcximately one control and sixteen standard elements.
During 1979, a total of 29 new fuel elements were received en the following schedule.
Date Elements U-235 Icading Supplier Feb. 2 6
140 gm Atomics International Feb. 6 4
169 gm Penn State Feb. 26 4
140 gm Atcmics International March 14 4
169 gm Pera State June 13 4
169 gm Penn State Oct. 11 7
140 gm Atcmics International There were no spent fuel shipments in 1979.
In 1979, the use of the heavy water reflector tank on cne face of the INR core required a throughput of 204 pounds of heavy water.
Fresh heavy water is used to replace heavy water in the tank in order to traintain a tritium level in the tank not greater than the 50 curie limit inposed by Technical Specifications.
No heavy water shipments were made in 1979.
1.3 Safety Pelated Operating Procedure Chances None.
1.4 Maintenance, Surmillance Tests, and Inspecticn Results as Required by Technical Specifications Maintenance, surveillance tests, and inspections required by Technical Specifications were ccrapleted at prescribed intervals.
Procedures, data sheets, and a maintenance schedule / record provide documentation.
1.5 Summary of Changes, Tests, and Experiments for which NPC, Authorizaticn was Required None.
1.6 Operating Staff Chances During 1979, the follcwing reactor operations ' staff changes occurred.
Resianed Effective Date Peplacement A. Cecil Drecsler March 23, 1979 Pcbert Johnson David P. Waalkes May 18, 1979 Thomas Eis Roger A. West July 15, 1979 Christcpher C. Branncn 1.7 Reportable occurrences There w re no reportable occurrences in 1979.
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2.
POER Ga'ERATICN SGNARY The follwing table sumarizes reactor power generaticn for 1979.
Full Power Operating Cperating Megawatt Percent Cycle Inclusive Dates Hours Ecurs Hours Availability 168 Jan. 3 - Jan. 30 466.2 410.7 821.4 69%
169 Jan.30 - Feb. 27 488.7 454.4 908.8 73%
170 Feb.27 - Mar. 27 448.0 387.1 774.2 67%
171 Mar.27 - Apr. 24 465.0 444.8 889.6 69%
172 Apr.24 - May 22 471.1 433.1 866.2 70%
173 May 22 - Jun. 18 447.8 395.8 791.6 67%
174 Jun.18 - Jul.17 447.9 367.0 734.0 67%
175 Jul.17 - Aug. 14 486.9 453.9 907.8 72%
176 Aug.14 - Sep. 11 454.3 331.3 662.6 68%
177 Sep.ll - Oct.
9 493.1 436.0 872.0 73%
178 Oct. 9 - Nov. 6 476.5 445.3 890.6 71%
179 Nov. 6 - Dec.
4 362.7 325.2 650.4 54%
180 Dec. 4 - Jan. 2 321.3 311.8 623.6 48%
Tol'AL 5829.5 5196.4 10,392.8 67%
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3.
UNSGEDUIED PEACIOR SHLTDCkN SCE'AW
'the follcwing table surrearizes unscheduled reactor shutdowns for 1979.
Unscheduled Shutdowns
'Ibtal Unscheduled Shutdowns.......
28 Shutdowns /100 Operating Hours......
.46 Shutdown Types 4
Single Pod Drop (NAR)
O Multiple Rod Drop (NAR)
Operator Action.............
12 4
Operator Error Process Equipment...
O Reactor Ccntrols 5
Electric Power Failure 3
Shutdown Type Definitions Single Rod Drcp and Multiple Rod Drop (NAR)
An unscheduled shutdown caused by the release of one or more of the reactor shim safety rods fran its electromagnet, and for which, at the time of the rod release, no specific capcnent ralfunctica can be identified as causing the release.
Operator Action A condition exists (usually sone minor difficulty with an experiment) for which the operator on duty judges that shutdown of the reactor is required until the difficulty is corrected.
Operator Error The cperator on duty nakes a judgecent or manipulative error which results in shutdcun of the reactor.
Process Equicment Shutdown caused by malfunctions in the process equipnent interlocks of the reactor control system.
Reactor Centrols Shutdcwn 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 facility.
Cycle Sunrary of Unscheduled Shutdowns Cycle 168 2ere were six unscheduled shutdowns during cycle 168. Three of these were caused by the primary water flow measuring channel when it mcarentarily dropped belcw the trip point. It has been determined that a corrbination of circumstances lead to the problem. All of the trips came frcm prirary flow No. 2 circuit.
The primary flow No.1 circuit continued to indicate nonral water ficw through the core during this pericd. A calibration check of the system indicated that while it was within normal specifications the trip point was near the upper limit. This fact along with the norral oscillations of the flow apparently caused the shutdowns.
Two of the remaining unscheduled shutdcwns were the results of cperator acticn in response to abnorral ccnditions. The first occurred when the operator noticed abnorral readings on a CSA during a routine check. The instrument was replaced and the reactor returned to power. The problem was later traced to a bad vacuum tube in the instrument. The second operator actica shutdown occurred when the operator noticed that the fast speed motor on the linear level recorder was not functioning. The reactor was shutdown to allow replacement of the motor.
The final unscheduled shutdown was the result of an operator error. 'Iko merrbers of the reactor operations staff were attempting to operate a strall recorder in the control rocm. As it was being connected to the power supply a receptacle was shorted causing a voltage transient in scme of the reactor control circuits. his transient was seen by the reactor safety circuit which in turn dropped B and C shim safety rods. We reactor was returned to power without further difficulty. It was decided that the recorder in question would not be used as planned. -
Cycle 169 There was one unschedulei shutdown during cycle 169. This shutdown was initiated by the reactor operator when a narber of the cperations crew discovered that an in-core sample was not attached to its support wire. After the reactor was shut down, the irradiation facility was removed and the sanple retrieved. The reactor was returned to power without difficulty.
Cycle 170 There was cne unscheduled shutdown during cycle 170. This shut-dcwn was initiated by the reacter building exhaust radiation mcnitor. The alarm was silenced inrediately and the false alarm announcement was given en the building public address system.
The cperaticns crew were aware that this nonitor had been inter-mitently erratic for scme tire and were unable to make an effective repair. The reactor was returned to power without further difficulty. A new detector-photo nultiplier for the system has been ordered frcm the supplier.
Cycle 171 The caly unscheduled shutdown during cycle 171 was due to a mcmentary electrical power interruption to the facility. The reactor was returned to pcwer without further difficulty.
Cycle 172 There were two unschedulal shutdowns during cycle 172; both were as the result of an operator action. The first occurred when the reactor operator noticed a small reduction (4%) of indicated power cn the two safety charnels. The cperator decided to reduce power and investigate. Shortly before this was noticed the crew had performed a routine chmbr adjustIrent.
After checking the cha:rber adjusting mechanisms and verifying their prcper orientaticn the reactor was returned to power.
It is believed that the linear level chamixer which had just been adjusted moved slightly in its mcunt which caused reactor power to be lowered by the servo-centroller. The crews have been cautioned to check the chamber alignments and orientations after each adjustment to verify that they are in a stable ecnditicn..
The second cperator action shutdown occurred 2en the cperator noticed that cne of the samples being irradiated in the core had Wme detached fran its rotating mechanism. The reactor was shutdtsn and the sample retrieved frcm the irradiatien location and stored. The reactor was then returned to power.
Cycle 173 The first unscheduled shutdown.in cycle 173 occurred when the operator noti d that.the supporting wire for an in-core sarple had cane 1cose fran a motor drive and had fallen cnto the reactor The reactor was shutdown and the sample ccntainer re-core.
trieved. It was subsequently discovered that the silicon can-tainer had broken which made it necessary to remove the irradiation facility fran the core in crder to remove the samples fran it.
The reactor was returned to pcwer after reinstalling the irradiaticn facility.
The seccnd unscheduled shutdown was an operator action after hearire an intermittent radiaticn alarm and noticing that coe of the in-core egeriments had caused a small amount of ccntaminated water to bubble out a vent line cnto the floor. The experiment was renoved fran the core and the area around the spill isolated until Health Physics perscnnel were alerted. The reactor was returned to power. The spill was cleaned up by Health Physics perscnnel during the day with cnly slight contaminaticn to the surrunding area and the two reactor operators. There was no airborne radio-activity detected and no excessive radiation exposure to facility persconel.
The third unscheduled shutdcun was due to a short duraticn power failure to the facility. The reactor was returned to power with out further difficulty.
The fourth unscheduled shutdcun was due to an operator error wh"n during a calorimeter r m, the primary purtp was mistakenly turned off instead of the seccndary punp. The error was corrected and the reactor returned to pcuer.
The fifth unscheduled shutdown was for No Apparent Reascn when "A" shim red dropped during a startup. The probable reason was a noise spike when the shim rod raise handle was actuated. The restart was acccuplished without further difficulty.
The final unscheduled shutdown was by operator action when water was noticed in the vent line of an in-core experiment. The sample was removed frcn the core. During removal the operator noticed that a hose fitting had leaked scme cantaminated water ento the floor. The satple was set aside and the area secured.
The reactor was returned to power.
Cycle 174 The first unscheduled shutdcwn was a single rod drop caused by electronic noise in the safety amplifier cireait. The cause of the noise spike is believed to be the shim rod raise-lower switch. The reactor was restarted without difficulty.
The second unscheduled shutdown was by cserator action when it was decided that area radiation monitor No. 3 **as not functioning properly. The operator noted a lower than normal reading during a routine check and decided to shut down the reactor. The problem was traced to a loose connector in the unit. The reactor was restarted without difficulty.
Cycle 175 The first unscheduled shutdown was due to operator error while unloading a sample frcm the reactor. The sample was to be moved to the underwater unloading tray but during the transfer the operator inadvertently raised it close enough to the surface to trigger the fuel vault radiation monitor thereby causing the building alarm to sound and the reactor to be shutdown. The trip point of this monitor is 5 mr/hr. The Health Physics staff reccnstructed the incident in an atte:rpt to detennine the radiation dose the operator received. The conclusion was that because of the very short duraticn of the exposure that the dose was probably lcw. In order to verify this the operators film badge was sent to be analyzed. This analysis indicated that he had received 40 mr in the eighteen day period which included the incident.
The operations staff has been rerunded of the problems associated with sanple mcuecent in the pool. The building evacuation procedure was irrplemented and carried out as planned. The reactor was restarted without further difficulty.
The secord unscheduled shutdown was caused by electrcnic noise picked up by the safety system which caused a shim rod to drop into the core. Tre reactor was restarted without further difficulty.
The final shutdchn during the cycle was by operator acticn when the control rocm cperator was requested to shutdown by an cperator cn the reactor bridge. The cperator was in the process of loading an experiment into the core when he drcpped an allen wrench cnto the core. The wrench was removed and the reactor restarted.
Operations personnel have been reminded of their responsibilities while working on the reactor bridge and near the reactor pool.
Cycle 176 The first unscheduled shutdcEn was for No Apparent Reascn; "B" rcd dropped during a 1cw pcwer run experittent. The reactor was checked out and restarted without further difficulty.
The second unscheduled shutdown occurred during a neutrcn radi-ography experiment cn the reacter bridge. The power lead for "C" magnet was touched causing the reactor to scram. Subseg e t investigation revealed the power lead to be open but held in place by the lead insulation. The power lead was replaced.
Cycle 177 The first unscheduled shutdown was for No Apparent Peason.
There were no abnormal indications cn the ccntrol panel other than "B" rod had dropped. The reactor was restarted without further difficulty.
The second unscheduled shutdown was an operatcr action when the control rocm cperator was told by the assistant operator that an in-core experiment holder had becane detached frcm its mount and the cable had fallen ento the core. The cable was retrieved ard the reactor returned to power. The design of this system has been altered to ncw include a device for retaining the cable.
Cycle 178
'Ihe single unscheduled shutdown in cycle 178 was caused by a loss of electric power to the facility. The power was out i
empletely for approximately one hour initially and for an additional five minutes later in the afternoon. The prcblem was traced to a Detroit Edison equipment failure in another building. During the cutage the e:mrgency generator responded normally and was able to carry the required load. The security system maintained surveillance during the outage after the initial alarms went out to the security office. When power was restored the reactor control system was checked out and the reactor returned to power.
Cycle 179
'Ihere were no unscheduled shutdowns in cycle 179.
Cycle 180 There were no unscheduled shutdesns in cycle 180.
i 4.
QRRIITIVE MAINIENANCE CM SAFETY-REIATED SYSTHE AND COMPONENTS None 5.
CHANGES, TESTS, AND EXPERDENTS CARRIED CUT WITHOUT PRIOR IEC APPPGAL 5.1 Building Alarm Test.
01 July 18, an unannounced test of the facility building alann was capleted. Approximately fifty people were evacuated in less than two minutes. The reactor was returned to power at the conclusion of the test.
6.
PADIOTCTIVE EFFID NP PE EASES Quantities and types of radioactive effluent releases, envirarental monitoring locaticns and data, and occupational perscnnel radiation exposures are provided in this secticn.
6.1 Gas Effluents Quantity Unit Gaseous effluent concentrations are averaged over the period of
. reactor operation. Gaseous effluent ccreentrations do not include 400 dilution factor, a.
Total gross radioactivity 148.10 Ci b.
Avera;e ccncentraticn released during nonnal steady state reactor operaticn.
3.00E-7 pCi/cc c.
Average release rate during normal steady state reactor cperaticn.
7.91 pci/sec d.
Maximtma instantaneous ccncentration during special operaticns, tests, and experiments.
None pCi/cc e.
Percent of MPC without diluticn factor.
716 f.
Percent of MPC with 400 dilution factor.
1.79 6.2 Icdine Peleases a.
Total icdine radioactivity by nuclide based upon a represen-tative isotopic analysis. Paquired if iodine is identified in primary coolant samples or if fueled experiments are ecnducted at the facility.
Not required.
b.
Iodine releases not related to reactor cp raticn.
18.01 mci c.
Percent of MPC without dilution factor.
21.6 d.
Percent of MPC with 400 dilution factor
.054 6.3 Particulate Releases Quantity Unit Gross alpha activity is required if the operaticnal or experi-mental program could result in the release of alpha emitters.
Particulate releases do not include 400 diluticn factor.
a.
Total gross beta-ga:tma radio-activity.
281.54 pCi b.
Gross alpha radioactivity.
Not Required c.
Total gross radioactivity of nuclides with half lives greater than eight days.
3.38E-13 pCi/cc d.
Percent of MPC for particulate radioactivity with half lives greater than eight days.
0.338 e.
Per nt of MPC with 400 diluticn factor.
8.45E-4 6.4 Liauld Effluents Total alpha ralicactivity is regaired if the operational or experimental program could result in the release of alpha enitters.
Release concentrations are averaged over the period of the release.
Liquid concentrations do not include 300 dilution factor, a.
Total gross beta-gamna radioactivity 118.92 nCi b.
Average gross beta-gamna ecncen-tration.
3.19E-4 pCi/ml c.
Maximum gross beta-gamna ccncen-tration.
3.00E-3 pCi/ml d.
Total alpha radioactivity Not required e.
Average alpha concentration Not required f.
Total liquid waste volume.
3.73EV '
ml g.
Total dilution water volume prior to release frcm the facility.
O m1 h.
Total dilution water volume of North Campus sanitary sewer.
1.12E411 ml i.
Isotcpic analysis and average isotopic ccncentration of waste storage tanks with average gross beta-ganna cencen-4 traticns in excess of 9 x 10 pCi/ml.
Not Required.
Quantity Unit j.
Percent of MPC withcut dilution based on gross beta-ganraa radio-activity.
106.2 k.
Percent of MPC with 300 diluticn factor based en gress beta-ganma radioactivity.
.354 6.5 Envircmental Monitoring The environmental mcnitoring program for the Ford Nuclear Reactor facility ccnsists of direct radiatien monitors (TID) and air sampling stations located around the facility and selected water and sewer sampling statiens.
TID's The TIDs location at stations in the vicinity of the facility are collected at two month intervals and sent to a ccr:mercial dosimetry ccmpany for analysis. At the submittal date of this report, TID data through Septerier 30, 1979 was available.
Staticn Description PwHng Unit Highest staticn (absolute) 18 mrem /2 ro Iowest staticn (absolute) 4 mran/2 ro
~
Average of all stations 9
mrem /2 to Average background 8.2 mrem /2 mo Average annual dose (above background) 4.8 mrem /yr.
TM average annual dose is determined by subtracting background ircm the average of all staticns and multiplying by six since the measured doses are for two month periods. Background is taken at a distance in excess of three miles frcra the reactor.
Dust: Four air grab sa:rples are collected weekly to the north (NorthWood Apartments), east (Research Administration Building),
south (Institute of Science and Technology), and west (Chrysler Center) of the reactor facility. Each millipore filter sample is counted for net beta activity. Approximate radioactivity ccncentraticns are 4.29E-14 pCi/cc which is.043% of MPC.
Icdine - 131: Air grab samples are collected mcnthly at the same locations that dust samples are collected. Each of the four charcoal cartridges is ccunted for Iodine - 131 with a e xlium iodide detector. Approximate radioactivity ccncentraticns are 8.82E-14 pCi/cc which is.088% of MPC.
Water: A daily 100 m1 tap water sample is collected at the University School of Public Health, making a ccrnposite monthly sample of 3000 ml.
On the first of each month, a 3000 ml grab sample of river water is collected above Ann Arbor (Pump Station).
On the fifteenth of each month, a 3000 ml grab sample of river water is collectal above the Ann Arbor seaage treatment plant (Dixboro) and another 3000 ml grab sample is collected belcw the sewage treatrent plant (Superior). Each of the four samples is oven-dried on a planchet for net M ta analysis. Approx 2 mate radioactivity ccncentrations are 4.78E-9 pCi/cc which is.16%
of MPC.
Sewace: Ann Arbor sewage plant perscnnel collect two 100 m1 sampleo daily; one raw and cne treated sewage. 'IWo ccrcposite 1500 m1 samples are picked up fran the sewage plant on the fifteenth and thirtieth of the month for analysis. Each sarple is oven-dried cn a planchet for net beta analysis. Approximate radioactivity concentraticns are 13.20E-9 pCi/ml which is.44% of MPC.
a.
Number of locatiens at which levels were found to be significantly higher than the remaining locations.
Ncne b.
Highest, lowest, and the annual concentraticns or levels of radiation for the sampling point with the highest average.
Annual Highest Icwest Average Unit TID mrem above Research Ad:ninistraticn Building 6
0 2
bkgd/2 mo Dust Research Administraticn Building 1.77E-13 3.20E-15 4.79E-14 pCi/cc Water Superior Road Bridge 9.31E-9 9.60E-10 5.29E-9 pCi/ml Sewage Ann Arbor Waste Water Treatnent Plant 2.56E-8 9.60E-10 1.32E-8 pCi/ml c.
The maximum cumulative radiaticn dose which could have been received by an individual continuously present in an unrestricted area during reactor operaticn frcm:
Direct Radiation The direct radiatica dcse is zero since a survey of accessible areas around the reactor building shows no detectable radiaticn dose rates above background.
Gaseous Effluent The maximum dose is based upon an assuned ccntinuous exposure to the average Iodine-131 ccncentration of 2.16E-ll pCi/cc released to the environment. The percentage of MPC based up;n that concentraticn is 21.6%. The maximum annual dose is.216 mren based upon NCRP 69; 1 MPC Icdine-131 yields 1 ren thyroid dose.
Liquid Effluent The annual dose frcm liquid effluents is zero. The most likely source of exposure would be an external dose to sewage treatrent personnel in the vicinity of raw sewage. The radicactivity concentrations at the Ann Arbor sewage plant yield no detectable dose rate on survey instruments, d.
If levels of radioactive materials in envircnnental nedia, as determined by an environmental mcnitoring program, indicate the likelihood of public intakes in excess of 1% of those that could result frcm continu-aus exposure to the concentration values listed in Appendix B, Table II, 10 CFR 20 estimates of the likely resultant exposure to individuals and to pcpulatica groups and assumptions upcn which estimates are bared.
Not applicable.
I l
6.6 Occupational Personnel Radiation Exposures.
The total ntrber of facility occupaticnal perscnnel for whan perscnnel mcnitoring was provided was 121 whole body and 50 extrenities. No radiation exposures greater than 50 mren were received at the facility by individuals under tha age of 18. A sumary of whole bcriy and extrenity exposures based upcn data fran January 1 - Decerber 31, 1979 follows.
Nunber of Individuals
~
Estimate Whole Body Exposure Range (ren) in Each Range l
No measurable exposure 65 Iess than 0.10 32 0.10 - 0.25 11 0.25 - 0.50 9
0.50 - 0.75 3
0.75 - 1.00 1
Greater than 1.00 0
TOTAL 121 Estimated Extremity Exposure Range (re:n)
(Maxinun of either hand)
No measurable exposure 8
Inss than 0.10 9
0.10 - 1.00 27 1.00 - 2.00 4
2.00 - 3.00 2
3.00 - 4.00 0
{
4.00 - 5.00 0
Greater than 5.00 0
TorAL 50 i
,