ML20206P302

Revised Rept of Reactor Operations for Jan-Dec 1998 for Ford Nuclear Reactor Michigan Memorial - Phoenix Project Univ of Mi,Ann Arbor

ML20206P302
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Site:	University of Michigan
Issue date:	12/31/1998
From:	MICHIGAN, UNIV. OF, ANN ARBOR, MI
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REPORT OF REACTOR OPERATIONS January 1,1998 - December 31,1998 FORD NUCLEAR REACTOR MICHIGAN MEMORIAL - PHOENIX PROJECT THE UNIVERSITY OF MICHIGAN ANN ARBOR March 1999 Revir.ed May 1999 Prepared For The U.S. Nuclear Regulatory Commission 9905180209 990511 PDR ADOCK 05000002 R PDR

_ _ _ _ _ _ )

Repon of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 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 reactor schedule of ten days of continuous operation at licensed power of two megawatts .

followed by four days of shutdown resulted in 5739.7 reactor operating hours, 5,442.2 operating hours at full power, 11,120.5 accumulated megawatt hours, and an overall reactor availability of 62.3 percent for the calendar year.

There were two reponable occurrences in 1998: Number 19, Reactor Operation with In-Operable Alarm Circut on Bridge Radiation Monitor - dated 30 Jul 98. Number 20, Inadequate implementation of Calorimerty Procedures - dated 9 Oct 98.

There were 17 unscheduled reactor shutdowns during the year.

There were no radioactive effluent releases above 10CFR20 limits. The maximum radiation dose equivalent received by an individual at the facility was 0.75 tem Total Effective Dose Equivalent.

The total radiation dose equivalent for all of the workers at the facility was 7.29 rem Total Effective Dose Equivalent.

i

FORD NUCLEAR REACTOR Docket No. 50-2 License No. R-28 i REPORT OF REACTOR OPERATIONS January 1,1998 - December 31,1998 This report reviews the operation of the University of Michigan's Ford Nuclear Reactor for the period January I to December 31, 1998. The report is to meet the requirement of Technical Specifications for the Ford Nuclear Reactor. The format for the sections that follow conforms to

Section 6.6.1 of Technical Specifications.

The Ford Nuclear Reactor 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 served and the 588 who died in World War II, encourages and supports research on the peaceful uses of nuclear energy and its social implications. In addition 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 purpose.

The operation of the Ford Nuclear Reactor provides major assistance to a wide variety of research and educational programs. The reactor provides neutron irradiation services and neutron beamport experimental facilities for use by faculty, students, and researchers from the University of Michigan, other universities, and industrial research organizations. Reactor staff members teach classes related to nuclear reactors and the Ford Nuclear Reactor in panicular 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 the year 1020 people participated in 67 tours.

The operating schedule of the reactor enables a sustained high level of panicipation by research groups. Continued support by the Department of Energy through the University Research Reactor Assistance Program (Contract No. J-AF-4000-000 (DE-AC02-76ER00385)) and the Reactor Facility Cost Sharing Program (Contract No. DE-FG07-80ER10724) has been essential to maintaining operation of the reactor facility.

1 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998

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 followed by four 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, a cycle consisting of four days or 96 full-power operating hours per week at licensed power followed by three days of shutdown maintenance was established in order to eliminate the periodic shutdown maintenance weeks needed in the previous cycle. Beginning July 1,1987, the reactor operating cycle n: turned to ten day operation at full power followed by four days of shutdown maintenance. This calendar year began with cycle 402 and ended with cycle 414. A cycle 4 covers four weeks; two of the ten day - four day sequences.

The reactor operates at a maximum power level of two megawatts which produces a peak thermal flux of approximately 2x10" n/cm /sec. 2 An equilibrium core configuration consists of approximately 41 standard and 4 control,19.75% enriched, 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"x 3"x 24".

Fuel elements are retired after burnup levels of approximately 35-40% are reached. Fuel burnup rate is approximately 2.46 gm U235/ day at two megawatts.

1.1 Facility Design Changes Afodification Request 126, Cooling Tower Fan Upgrade. One of the three cooling tower fan motors was n:placed and a variable speed solid state inverter placed in line to allow for temperature control.

Afodipcation Request 127, Cooling Tower Sump Level Indicator Replacement.

The original cooling tower sump level indicator was replaced with a new electronic level indicator with a pneumatic signal conditioner.

Afodipcation Request 128. Isolation of Interior FNR Building Drains from the Storm Sewer System. Several times during the summer of 1997, storm water backed up into the FNR building due to heavy thunderstorms. Investigation revealed that the storm sewers system was partially blocked due to recent construction in the area. It is also believed that the storm sewer system may be undersized for the current loading in the area. The two storm drain connections on the FNR first floor were sealed off from the storm sewer system. One line was blanked and the building drains diverted to the sanitary sewer system. The other line was fitted with a manual isolation valve.

Afodification Request 129, Primary Pump biotor Replacement. The primary pump and motor were replaced due to a bearing seizure in the motor, biodipcation Request 130, Replacement of Bridge Radiation bionitor. The bridge radiation monitor was replaced due to aging of the previous unit.

2 Revised May 1999

f Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 l 1.2 Equipment and Fuel Performance Characteristics l

l I

Reactor equipment and fuel exhibited no abnormal characteristics. Replacement of '

expended fuel elements resulted in an annual use of six standard fuel elements and two control fuel elements.

l No new standard or control fuel elements were received.

l There were no spent fuel shipments.

1.3 Safety-Related Procedure Changes I Safety-related procedures am those associated with operation, calibration, and maintenance of the primary coolant, the reactor safety system, the shim-safety rods, all scram functions, the high temperature auto rundown function, and the pool level rundown.

Calibration and Maintenance Procedures

1. CP-206, Safety System Period Channel C Calibration was revised to broaden the acceptable band for setting the 80% high power recorder switch. This revision was made because of Amendment No. 44 to the FNR License (Tin = 114 F @ 1.6 MW Rundown).

l 2. CP-216 - Ludlum Area Monitoring Ccdibration Proceduree was revised to include a function test of the High Reading Radiation Recorder (HRRR) reactor console alarm as part of the cormctive actions from Reportable Occurance No.19.

3. CP-219 - Ludlum 395 Area Monitoring Calibration Procedure was revised to include a function test of the HRRR reactor console alarm as part of the corrective actions from Reportable Occurance No.19.

Operating Procedures

1. OP-101-ReactorStart-Up was revised to instmet the operator to check pool j temperatures >90 F but <114 F prior to startup. This revision incorporated Amendment No. 44 to the FNR License (Tin > 114 F @ 1.6 4 MW Rundown). j 1.4 Maintenance, Surveillance Tests, and Inspection Results as Required by Technical Specifications. l 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.

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

None l 3 Revised May 1999 l

l l

Report of Reactor Operations ,

Ford Nuclear Reactor l January 1 - December 31,1998 l 1.6 Operating Staff Changes The following reactor opemtions staff changes occurred:

Newiv Hired Position h Deren Swartz Reactor OperatorII 12/07/98 Transferred Position h Trobi J. Hall Engineering Tech. II 06/30/98 1.7 Reportable Occurrences 1.7.1 Reportable Occurrence No. 19, Reactor Operation with in-Operable Alarm Circuit on the Bridge Radiation Monitor Description of the Event At approximately 15:45 on July 20, 1998 it was brought to the acting reactor manager's attention that the 0000 - 0800 shift crew had disabled the local alarm circuit in the Ford Nuclear Reactor's bridge radiation monitor at approximately 01:00 without authorization. The monitor had been giving spurious alarms due to temperature and humidity sensitivity. The reactor was shut down at 15:48 and a source check of the bridge monitor was performed. Results of this check showed that the remote radiation recorder alarm for this system, located in the control room, was also not operating.

A further check showed that the radiation recorder alarm for the bridge monitor was incorrectly programmed. The reactor was kept shut down pending further investigation.

Discussions with the reactor operators resulted in finding that the bridge monitor had previously given spurious alarms on Friday morning, June 26.

They were authorized at that time to operate the reactor with the bridge radiation monitor local alarm turned off from late that morning until the shutdown at 15:47 later on Friday afternoon. That is why they thought they could disable the local alarm on July 20. We believe the reactor may have been operated without a functioning bridge radiation alarm during this period also.

The bridge radiation monitor local alarm was restored to operation prior to the reactor start up on the following Tuesday, June 30. The bridge radiation monitor is Required Safety Related Equipment as defined in the Technical Specifications and is required to be functioning during reactor operation with an alarm setpoint not to exceed 50 mr/hr. The normal setpoint is 40 mr/hr. The reactor was operated for approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> Monday, July 20 and very likely was operated for approximately six hours on Friday, June 26 without an alarm function on the bridge radiation monitor.

The radiation measurement capability of the bridge radiation monitor was not disabled at any time during these periods. The bridge monitor radiation levels are checked by the operators and recorded on the Operational Checklist every two hours when the reactor is operating. A review of the 4 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 chart record for the two periods when the alarm function was defeated showed normal radiation levels for the bridge radiation monitor during the intervening time between Operational Checks.

Full compliance was restored on Tuesday, July 21,1998.

l l

Immediate actions taken l

The gamma radiation measurement circuit in the monitor outputs to a local meter, an internal alarm circuit (local alarm) and the radiation recorder. The radiation recorder accepts inputs from the various Tech. Spec. required radiation monitors that are located throughout PMIJFNR and gives the control room operator the ability to remotely observe the radiation monitors.

In addition,'the radiation recorder also has alarms that can be separately programmed for each input. These alarms indicate at the recorder and supplement the local alarms. The radiation recorder also has three alarm output relays, two of which are programmed to actuate under various circumstances. Output miay 1 is connected to actuate the High Reading Radiation Recorder (HRRR) alarm on the control panel. Output miay 2 is connected to the Stack Alarm on the control panel. All of the radiation monitor channels have radiation recorder alarms programmed. These alamis are programmed to actuate recorder output relay I giving the HRRR alarm in addition to the recorder alarm and local alarm. The PML stack 2 gaseous activity monitor (Stack GAD) is also programmed to actuate recorder output relay 2 giving the stack alarm.

The calibration of the bridge radiation monitor was checked on Tuesday, i

July 21. It was during this calibration check that the spurious alarms were isolated to the local alarm circuit in the monitor. The monitor was removed from service and replaced with a calibrated fully functional monitor unit.

The radiation recorder alarm programming was checked after the discovery that the alarm was not functioning on July 20, 1998. The check showed that the bridge monitor channel alarm programming was incorrect and that the recorder alarm and HRRR alarm were not active. The correct programming was restored. Between July 21 and July 27,1998 all radiation monitoring instruments were given source and functional tests to verify their alarm capability.

A memo was routed to all FNR/PML staff reminding them that absolutely no alarms are to be bypassed or alarm set points modified without the explicit approval of reactor management. The memo also reminded them of the need to properly document any changes that are made.

Calibration and Maintenance Procedures CP-216, Ludlum Area Monitoring l

Calibration Procedure and CP-219, Ludlum 395 Area Monitoring Calibration Procedure, the calibration procedures for the facility's area radiation monitors, were modified to include a function test of the HRRR reactor console alarm.

l 5 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor l January 1 - December 31,1998 l

Root Causes and Long Term Actions to Prevent a Recurrence l

I Two items have been identified as root causes. First, an inadequate implementation of procedure CP-308, Equipment Out of Operation. An immediate change was made to CP-308 instructing operators to promptly inform the On-call Supervisor or reactor management of faulty or improperly operating equipment. A memo was routed to the staff regarding I

implementation of CP-308. A further, more broad modification will be made to CP-308, after due consideration, that will provide reactor I

management a mechanism to track actions taken to repair or replace faulty l

equipment in a timely manner.

l The second root cause is an incomplete understanding of the importance of f l assuring that all equipment, set points, alarms, and surveillance j requirements stipulated in the FNR License and Technical Specifications are l

met and the absolute requirement to obtain authorization prior to taking any potentially compromising action in regard to the same. A closed book 1 training exercise (test) on the License and Technical Specifications was l administered to remind all Operator and Senior Licensed personnel of the l l Tech. Spec. requirements and the need for compliance.

A non-cited violation was given for this event based upon the corrective action, self identification, non-willfulness, and non-repetitive nature of this event.

1.7.2 Reportable Occurrence No. 20, Inadequate Implementation of Calorimertry Procedures Description ofthe Event On Friday, September 25,1998 at 18:39 the FNR was started up in preparation for 2.0 MW operation following a three-week shutdown. At 19:29 reactor power was increased from low power to a nominal 1.0 MW (50% on the 2 MW range of the Linear Level channel) and the reactor placed in automatic control for a calorimetric determination of core thermal power through 0P-106, Power Level Determination.

The calorimetry was completed and analyzed to determine the tme 1.0 MW Linear Level control setpoint at approximately 20:51. At 20:58 the setpoint for automatic control was reduced from 50% to 49% to bring the true reactor power to 1.0 MW as determined by the calorimeter. The swing-shift Lead Reactor Operator contacted the Assistant Manager for Operations, who was the On-Call Supervisor that night, and obtained permission to increase reactor power to 2.0 MW. Power was increased and at 21:12 the reactor was put into automatic control at a Linear Level channel setpoint of 99%.

The calorimetry procedure required a Linear Level automatic control setpoint of 98%.

At approximately 23:30 the Interim Reactor Manager visited the Control Room to check on progress prior to leaving the facility. The Lead Reactor Operator brought to the Manager's attention that he believed the power level to be low. He felt this was the case because the primary coolant temperature rise AT across the core was significantly less than what would be expected l

i l  !

i 6 Revised May 1999

]

l 1

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 at 2.0 MW. The AT was 12.8 0F at 23:00. The 2 MW core AT at the 3 current flow condition of 1000 gpm is expected to be approximately 13.8 )

oF. In addition, a physical adjustment to the position of Power Level channelion chambers had been required, to bring the indicated power level up to the 2 MW operating band of 2.0 to 2.1 indicated power. A physical adjustment to the ion chamber positions had not been anticipated. There l were no changes to the core configuration or maintenance performed on tie Power Level channels that would have affected their indicated power to the extent of requiring a physical adjustment to their position subsequent to the last reactor operation at 2.0 MW on September 4.

The Interim Reactor Manager reviewed the calorimetry data and did not see any obvious errors. He told the Lead Operator to maintain the current power level correcting for shim rod shadow and that the discrepancy would be investigated on Monday, September 28. The Interim Reactor Manager was not informed at the time nor was he aware of the unnecessary 1%

addition to the Linear Level setpoint the Lead Operator made when power was increased from 1 to 2 MW at 21:12.

I Saturday morning, September 26, the Assistant Manager for Operations reviewed the calorimetry data with the intent of setting the target operating band for the core AT. It ivas at this point that he noticed the incorrect Linear i Level setpoint (99% instead of 98%) that had been initially established for 2 MW at 21:12 on September 25. At 10:47 the Assistant Manager for Operations lowered the Linear Level setpoint by 1% to compensate for the prior error and contacted the Interim Reactor Manager to discuss the event and determine what additional actions should be taken.  ;

i

Safety Sigmficance of the Event l At the time of discovery it was believed that the reactor might have operated for a period of approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, between 21

12 Friday, September ,

25, and 10:47 Saturday, September 26, at a steady-state power level of 2.02 l MW thermal, in violation of the license power level of 2.0 MW thermal. ,

The 1% power increase is believed to be well within experimental i uncertainties associated with our method of determining thermal power )

level. In addition, the safety significance of operating at 2.02 MW thermal ,

is minimal. The FNR Safety Analysis Report shows a steady-state power l level of 4.68 MW at the mimmum pn, mary coolant flow of 900 gpm and 18 feet of pool head is required before boiling may occur in the hot channel.

A second calorimetry experiment was performed on October 1,1998 to establish the true reactor power level. A conservative analysis of the second calorimetry data indicates the reactor operated at an actual power not exceeding 1.97 MW during the 14-hour period before the setpoint error was corrected.

The FNR management feels obligated, however, to consider the possibility that a different combination of reactor system parameters, subject to a similar procedural error, could have resulted in a larger power adjustment, e.g., a few per cent, possibly resulting in an overpower condition. Thus, we consider the incorrect implementation of the calorimetry procedure, OP-l 7 Revised May 1999 l

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 106, and delayed discovery of the error as a serious mistake on the part of ,

the operations staff and management.

]

Immediate Corrective Actions l

Upon discovery on Saturday, September 26, of the error made in the Linear Level setpoint, reactor power was reduced by 1% to compensate for the error. With this 1% power reduction, the actual reactor power level during  ;

the subsequent reactor operation until the end of the operation cycle on l Friday, October 2, is estimated to have been less than 97.5% of license l power or 1.95 MW. l Root Causes of the Event In their written descriptions of the event and in oral discussions with the FNR management, the Lead Operator and Reactor Operator, responsible for the procedural error, did not demonstrate sufficient understanding of the i sigmficance of the calorimetry in establishing thermal power level of the '

reactor. There were reasons to suggest that reactor system parameters deviated somewhat from normal ranges, as indicated by the need to adjust the Power Level channel positions. This, however, did not justify any ad hoc adjustment in the Linear Ixvel setpoint, although the Lead Operator indicated that he had no intention of increasing the power level beyond the erroneous 1% adjustment.

There is also indication that the two operators on swing shift did not have sufficient discussion on the system parameters observed and steps required for properimplementation of the calorimetry procedure. Furthermore, they did not bring to the attention of the On-Call Supervisor any concerns they had regarding the results of the calorimetry or the 1% erroneous adjustment made to the Linear Level setpoint. This indicates lack of proper communication and questioning attitudes between the operators themselves and between the operators and supervisory personnel. Finally, the operators did not follow the steps delineated in OP-106 correctly, indicating their lack of attention to detail.

Subsequent Corrective Actions The Lead Operator responsible for the incident was relieved of Lead Operator responsibilities for a period of time. Additional training of the Lead Operator and Reactor Operator, who were on swing shift at the time of the incident, has been completed. Following this training the Lead Operator and Reactor Operator demonstrated adequate knowledge and were reinstated to their original positions.

A staff meeting was held on October 5 to discuss and emphasize the importance of (a) teamwork and communication, (b) thoroughness and attention to detail, and (c) understanding of the physical basis for operational procedures. A summary of the staff meeting was distributed as a memorandum to the entire staff on October 6. The memorandum has supplemented the existing codes of performance for the FNR staff in the retraining of the entire licensed staff.

1 8 Revised May 1999 l

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 No violation was issued for this event.

1.8 Non-Routine Occurrences 1.8.1 Hot Demineralizer Leak Causes 75 Gallon Loss of Pool Water to Radioactive Liquid Retention Tank System Description ofElement On May 10,1998 at 0250 reactor operators discovered a leak in one of the two resin columns comprising the "B" Hot Demineralizer (Hot DI) system.

l The leak resulted in the loss of approximately 75 gallons of reactor pool i water to the facility's radioactive liquid waste retention tank system (total reactor pool water volume is approximately 48,000 gallons). Hot DI "B" was isolated and 2Mw reactor operation continued. Safe operation of the reactor was not threatened at any time during this occurrence. No i radioactivity was released from the facility as a result of this occurrence.

The leak was discovered while the reactor operators were investigating the l cause of an unusually large observed decrease in pool water level amounting to one-half (1/2) inch during the prior two hours. The normal pool water loss rate is about one and one-half (1-1/2) inches per day.

l An automatic shutoff valve that was installed with the most mcent Hot DI  !

system upgrade was not in service because of past erratic behavior in its )

flow sensor.

~

Hot DI System The FNR Hot DI system consists of two sets of fiberglass reinforced resin tanks containing mixed cation and anion ion exchange resin beads. The purpose of the Hot DI system is to maintain the pH and conductivity the pnmary coolant water. The tanks are physically located in the basement of the reactor building. The Hot DI system taps off of the primary coolant retum piping diverting a small fraction (14 gallons / minute) of primary coolant water to the Hot DI's. Nominal primary coolant flow rate is 1050 gallons / minute. Valves are installed that allow isolation of the Hot DI system from the primary coolant system. A check valve prevents back flow from the pool through the Hot DI retum line. Most of the Hot DI system has been switched to Schedule 80 chlorinated polyvinyl chloride (CPVC) pipe, fittings, and valves. During normal operation one set of DI columns will be put on-line when conductivity of the pool water reaches 2.5 mho/cm. After the conductivity has been brought down to 1.5 mho/cm the DI's will be bypassed until conductivity again reaches 2.5 mho/cm.

When the on-line set of columns (either "A" or "B") is no longer effective the other set is put into service. The out-of-service set of columns are restored after a suitable decay period by retiring the upstream column, moving the downstream column to the upstream position, and replacing the downstream column with a tank containing fresh ion exchange resins.

9 Revised May 1999 4

l Report of Reactor Operations j Ford Nuclear Reactor '

January 1 - December 31,1998 Safety Analysis Testing conducted on May 18 during the shutdown maintenance period irranediately following the even showed that the maximum primary coolant loss rate through a guillotine break in the Hot DI piping is 15.9 gallons / minute. This loss rate would cause a decrease in the water height above the core at the rate of 6.4 inches per hour (one inch of pool height is equal to 150 gallons of water).

The water would flow into pits and floor drains in the reactor basement that drain to a pair of sumps. Sump pumps transfer the water to the 3000 gallon radioactive liquid retention tanks in the Phoenix Memorial Laboratory building. Twenty-nine gallons is pumped each time the sump pumps actuate. An alarm also annunciates in the reactor control room when the reactor control circuits am energized and either sump pump actuates. The normal frequency at which the sumps pump out and alarm is about once per eight hour shift.

Reactor Operating A worst case leak in the Hot DI system would cause repeated sump alarms and would quickly be investigated, discovered and isolated by the reactor operators when the reactor is operating. This operator action would take place well before the low pool level alarm at minus five inches or the automatic reactor rundown at minus 12 inches would occur.

ReactorSecured Two pool level alarms are active when the reactor is secured. A local alarm is actuated at a pool level of about minus seven inches. The second alarm, actuated by the same sensor that gives the pool level reactor rundown signal at minus 12 inches,is sent directly to the alarm panel at the UM Depanment of Public Safety which is staffed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day.

The worst case Hot DI leak situation when the mactor is secured would occur if the local alann actuates just after the security guard has passed through on his rounds. In this situation, knowledge that a problem exists will probably not be known until the minus 12 inch alarm actuates. Given a one hour response time between Public Safety contacting the On-Call Supervisor (OCS) and the OCS isolating the Hot DI system, an additional 6.4 inches would be lost from the pool. The pool level would still be at least 18 feet 5 inches above the core when the leak is isolated.

Radiation levels at the surface of the pool would not be significantly altered by a drop in pool level from 20 feet above the core to 18 feel 5 inches. The intensity of direct gamma radiation from the core would increase by a factor of approximately 60 at the average fission gamma energy of 700 kev.

However, the contribution direct core gammas make to the total dose rate at the pool surface is insignificant as compared to the dose due to activated impurities (primarily Na-24) in the pool water. The reactor has operated with the pool level lowered by as much as five inches under special l

circumstances. No discernible increase in surface radiation level was detected at the time. Normal pool surface radiation levels are 10 to 15

mR/hr.

I 10 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 Reactor pool water leaking from the Hot DI system would be captured by the facility's radioactive liquid waste system. There are no direct pathways between the basement of the FNR building, where the Hot DI's are located, and unrestricted areas. Therefore, no radioactivity would be released to unrestricted areas.

The conclusion is that a leak from the Hot DI system will have no reactor safety significance.

Possible Root Cause In the course of testing the Hot DI system following this event, it was observed that a partial vacuum could be drawn on the resin columns depending on how the system's valves were operated. This caused an inward flexing of the resin tank. Such flexing may have initiated the crack in the failed resin column.

Immediate Actions Taken Immediate actions taken were: 1) to isolate and secure Hot DI "B", and 2) to isolate and secure the entire Hot DI system during periods when the reactor was not operating until funher testing could be completed.

Long Term Actions to Prevent a Recurrence The vendor who supplied the resin columns was contacted. The burst pressure was quoted to be 250 psi. The maximum recommended working pressure was said to be 100 psi. A 100 psi pressure test is performed on each DI resin column before it is put into service.

A number of tests were performed on the entire Hot DI system on May 18, 26 and June 2.

The entire Hot DI system was isolated and then pressurized and hydrostatically leak tested at 85 psi for more than one hour on May 18. No leaks were found. The system was hydrostatically tested a second time on June 2. One fitting containing a conductivity cell cracked at a pressure between 100 and 120 psi. The fitting was replaced by a CPVC pipe nipple and the system was tested again at 120 psi. No further leaks were detected.

The maximum operating pressure in the Hot DI system occurs in the piping immediately after the return pump. The Hot DI return pump is normally operated with a back pressure of 50 - 60 psi to prevent cavitation. Maximum pressure will occur when the pumps are running and the Hot DI return i isolation valve is shut. Under this condition a pressure of 80 psi was observed on May 18. The June 2 leak test showed the Hot DI system piping is capable of withstanding the highest anticipated operating pressure without failure.

1 A new auto-shutoff valve and flow sensor using a different design have been procured and installed, but-are not yet wired electrically. This will make the Hot DI system self-isolating in case of a major leak when the installation is complete.

I1 Revised May 1999

I 7 Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 OP-211, Operating the Hot Demineralizer System has been modified to secure the inlet and outlet pumps prior to changing valve line-ups in the system. This should eliminate unnecessary flexing of the resin tanks.

! 2. POWER GENERATION

SUMMARY

The following table summarizes reactor annual power generation.

, Oxrating Full Power Megawatt Percent i Cycle Inclusive Dates ,1qua Ooerating Hours Availability Hours l 415 01/13/98- 514.6 487.7 979.0 72.6 02/11/98 416 02/11/98- 471.2 460.1 923.6 68.5 03/10/98 417 03/10/98- 411.1 274.0 553.6 40.1 04/07/98 418 04/07/98- 120.8 5.9 17.6 0.9 05/05/98 419 05/05/98- 412.7 402.9 810.2 60.0 06/01/98 420 06/01/98- 493.0 482.9 968.6 71.9 06/29/98 421 06/29/98- 327.7 314.8 634.6 46.8 07/27/98 422 07/27/98- 264.1 213.2 500.8 31.7 08/24/98 423 08/24/98- 215.6 192.3 386.3 28.6 09/21/98 424 09/21/98- 378.0 367.5 738.3 54.7 10/19/98 425 10/19/98- 429.4 421.9 845.2 62.8 11/16/98 426 11/16/98- 436.8 424.7 853.4 63.2 12/14/98 427 12/14/98- 198.8 189.7 381.2 28.2 01/11/99 Totals: 4673.8 4237.6 8592.4 59.8

3. UNSCHEDULED REACTOR SHUTDOWN

SUMMARY

The following table summarizes unscheduled reactor shutdowns.

3.1 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 for which at the time of the rod release, no specific component malfunction and no apparent reason (NAR) can be identified as having caused the release.

12 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 Operator Action A condition exists (usually some minor difficulty with an experiment) for which the operator on dutyjudges that shutdown of the reactor is required until the difficulty is corrected.

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

Process Eauipment Failure Shutdown caused by a malfunction in the process equipment interlocks of the reactor control system.

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

Electrical Power Failure Shutdown caused by interruption in the reactor facility electric power supply.

3.2 Cycle Summary of Unscheduled Shutdowns Cycle 417 March 25,1998. The reactor shutdown was due to a high power - no flow signal caused by the failure of the primary pump. The primary pump motor controller blew two main-line fuses when the shaft end motor bearing overheated and seized.

The root cause of this unscheduled shutdown was a manufacturing defect in the motor. The reactor was not operated in forced convection mode following the replacement of the primary pump and motor pending a License Change (Amendment #44). Process Equipment Failure.

Cvele 419 May 5,1998. There was a reactor scram due to Safety Channel A. This shutdown was attributed to electronic noise. Cables and connectors were checked prior to restart, no problems were found. The reactor was returned to operation at 2MW without difficulty. Process Equipment Failure.

Cycle 420 June 24,1998. The reactor shutdown due to a tornado warning. After the tornado warning was lifted by the University of Michigan Department of Public Safety, the reactor returned to normal operation at 2MW. Operator Action.

13 Revised May 1999 i

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 Cycle 421 July 20,1998. The reactor shutdown due to erratic operation of the Reactor Bridge Radiation Monitor alarm circuit. A replacement radiation monitor (formerly the basement monitor) was installed in its place. The replacement monitor was calibrated, and found acceptable. Restart of the reactor was delayed due to other unrelated events (loss of electrical power and emergency generator problems).

Process Equipment Failure.

Cycle 422 1

August 4,1998. The reactor shutdown was due to Nitrogen-16 detector failure.

! The Nitrogen-16 indication was used to maintain constant reactor power level during rod shims. A Temporary Operating Instruction was written which used rod shadow tables to help operators maintain a constant reactor power level during shims. The reactor was restarted without difficulties. Troubleshooting and repairs were initiated for the Nitrogen-16 instrument channel. Reactor Controls.

August i1,1998. The reactor shutdown was due to control rod inward drift.

Troubleshooting indicated dirty terminal blocks on the control rod drive motor. The terminal blocks were cleaned, and the drift ceased. The reactor was restarted without difficulty. Reactor Controls.

August 17,1998. The reactor shutdown was again due to control rod inward drift.

Troubleshooting began. The rod drive motor was found to be OK. The attached worm gear drive was replaced due to signs of wear on the gear surfaces. The control rod drive assembly was exercised and operated satisfactorily. No more drift was observed. The reactor was restarted without difficulty. Reactor Controls.

Cycle 423 August 25,1998. The reactor shutdown was due to control rod drift. Slow inward drift of the control rod was noted by the operators while performing rod calibrations. The operators shutdown the reactor. The control rod drive mechanism was disassembled and inspected. A gear cavity found empty, was repacked with grease. The control rod drive mechanism was reassembled and tested. The reactor was restarted without difficulty. Operator Action.

August 27,1998. The reactor shutdown was due to Primary Pump tripping on overload. The primary pump and motor were inspected and appeared normal. The motor controller thermal overloads were reset, and the pump restarted. The reactor was restarted without difficulty. Process Equipment Failure.

Cycle 424 October 7,1998. The reactor was shutdown due to a small piece of debris, source unknown. The debris covered <25% of one channel and <10% of the element.

The fuel element (in location L36) was removed from the core in accordance with procedure AP-301 and shaken to dislodge the debris. The fuel element was then returned to its fonner location in the core. The reactor was then restarted without difficulty. Operator Action. l 14 Revised May 1999

Repon of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 October 10,1998. The reactor was shutdown in the morning due to a small yellow paint chip. The debris covered <25% of one channel and <10% of the element. The fuel element (in location L20) was removed from the core in accordance with procedure AP-301 and shaken to dislodge the paint chip. The paint chip was observed to be nearly neutral in buoyancy. The fuel element was then returned to its former location in the core. The reactor was then restarted without difficulty.

Operator Action.

October 10,1998. The reactor was shutdown in the afternoon due to reappearance of the yellow paint chip. The debris covered <25% of one channel and <10% of the element. The primary pump was secured and the debris drifted away from core.

The primary pump was then tumed on again. The paint chip was gone. The reactor was then restarted without difficulty. Note: The paint chip did not reappear as of March 1999. Operator Action.

Cycle 425 October 23,1998. The reactor shutdown was due to an inadvertent primary pump trip. While attempting to deenergize the fluorescent lights in room 2107 due to a smelly (" burning") ballast, staff cycled through the breakers in a 120 VAC distribution panel in the FNR second floor hallway, near the south stairwell. One of the breakers provides power to auxiliary control circuits for the primary and secondary pumps in room 2111. With the auxiliary control power off, the pumps stopped, thereby causing a high power /no flow reactor scram. After the ballast circuit was isolated, the primary pump was restarted and the reactor restarted without difficulty. The affected breaker was labeled to prevent a similar event in the future. Operator Error.

October 25, 1998. The reactor shutdown was due to a short intenuption of electrical power from Detroit Edison. The reactor was restarted without difficulty.

Electrical Power Failure.

October 28,1998. The reactor shutdown upon discovery that the reactor building ventilation dampers would not automatically close. University of Michigan Plant Depanment personnel were called to work on the dampers. They disconnected the air cylinder, then removed and disassembled it. The air cylinder components were all cleaned and lubricated. The air cylinder was then reassembled and reinstalled.

Plant Department installed a new three-way solenoid control valve and a new in-line oiler as replacement parts. The ventilation damper controls were exercised. The dampers worked smoothly and normally after the repairs were completed. Based on the successful refurbishment, the Assistant Manager for Reactor Operations authorized reactor restart and operation resumed November 3. Process Equipment Failure.  !

November 11,1998. The reactor shutdown was due to a thermal overload tripping {

the primary pump, thereby causing a high power /no flow reactor scram. The primary pump and controller were inspected and appeared normal. The motor controller thermal overloads were reset, and the pump restarted. The reactor was j restarted without difficulty. Corrective maintenance was scheduled on the electrical controller. A University of Michigan Plant Department electrician performed an '

inspection of the controller, checked all the in-line temninations, and performed a thermal scan of the internals to find hot spots (none found). A ventilation fan and l

15 Revised May 1999 i

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Report of Reactor Operations ;

Ford Nuclear Reactor January 1 - December 31,1998 exhaust grille were installed in the controller enclosure to provide some ventilation and reduce internal heat buildup. Process Equipment Failure.

Cycle 427 December 16,1998. The reactor shutdown was due to a high power trip on Safety Channel A. Suspected to be caused by electronic noise. Reactor Controls.

3.3 Characterization of Unscheduled Shutdowns Single Rod Drop (NAR) 0 Multiple Rod Drop (NAR) 0 Operator Action 5 Operator Error 1 Process Equipment Failure 6 Reactor Controls 4 Electric Power Failure 1 3.3 Unscheduled Shutdowns Total Unscheduled Shutdowns 17 Average Operating Hours Betvceen Unscheduled Shutdowns 233.9

4. CORRECTIVE MAINTENANCE ON SAFETY RELATED SYSTEMS AND COMPONENTS None

5. CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR NRC APPROVAL PURSUANT TO 10CFR50.59(a) 5.1 Modification Request 126, Cooling Tower Fan Upgrade

Description:

One of the three cooling tower fan motors was replaced and a variable speed solid state inverter placed in line to allow for temperature control.

Safety Evaluation: This modification made no change in the facility as described in the Safety Analysis and Technical Specifications for the Ford Nuclear Reactor.

16 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 5.2 Modification Request 127, Cooling Tower Sump Level Indicator Replacement

Description:

The original cooling tower sump level indicator was replaced with a new electronic level indicator with a pneumatic signal conditioner.

Safety Evaluation: This modification made no change in the facility as described in the Safety Analysis and Technical Specifications for the Ford Nuclear Reactor.

5.3 Modification Request 128, Isolation ofInterior FNR Building Drainsfrom the Storm Sewer System

Description:

Several times during the summer of 1997, storm water backed up into the FNR building due to heavy thunderstorms. Investigation revealed that the storm sewers system was partially blocked due to recent construction in the area. It is also believed that the storm sewer system may be undersized for the current loading in the area. The two storm drain connections on the FNR first floor were sealed off from the storm sewer system. One line was blanked and the building drains diverted to the sanitary sewer system. The other line was fitted with a manual isolation valve.

Safety Evaluation: This modification made no change in the facility as described in the Safety Analysis and Technical Specifications for the Ford Nuclear Reactor.

5.4 Modification Request 129, Primary Pump Motor Replacement

Description:

The primary pump and motor were replaced due to a bearing seizure in the motor.

Safety Evaluation: The safety evaluation identified an unresolved safety question, namely that changes in primary coolant forced convection flow may improve core cooling and decrease the margin of safety between the primary coolant temperature Limiting Safety System Setting and its associated Safety Limit (forced convection mode). Amendment No. 44 to the FNR License and Technical Specifications was submitted to the NRC and approved. This license amendment added a Limiting Safety System Setting (forced convection mode) for core inlet temperature of 144 F when reactor power level is greater than or equal to 1.6 MWth.

5.5 Modification Request 130, Replacement of Bridge Radiation Monitor.

Description:

The bridge radiation monitor was replaced due to aging of the previous umt.

Safety Evaluation: The safety evaluation describes the bridge radiation monitor as a monitor which measures gamma radiation levels and is connected to the radiation recorder in the control room. The bridge radiation monitor causes a "High Reading Radiation Recorder" alarm if the setpoint is exceeded. The upgrade of the bridge radiation monitor was determined not to be a change to the facility or its operation as described in the Safety Analysis and Technical Specifications for the Ford Nuclear Reactor nor an unreviewed safety question so the change out was implemented under 10 CFR 50.59.

17 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 5.6 Temporary Operating Instruction, Emergency Generator Manual Operation and Fuel Vault Monitoring.

Description:

This temporary operating instruction was written to provide procedural guidance on starting the emergency generator manually while the emergency generator auto-start sequencer was being replaced. Additionally this procedure provides guidance to operations staff on the additional measures to ensure the security of new reactor fuel stored in the fuel vault upon a loss of electricity to the facility.

Safety Evaluation: The safety analysis describes the emergency generator, associated bus transfer and essentialloads supplied. The manual starting of the

( emergency generator was determined to be a change to the facility or its l operation as described in the safety analysis. The subsequent analysis

(

determined that the emergency generator is not required to shutdown or maintain shutdown of the reactor systems or to minimize the consequences of an accident or a transient. Manual operation of the emergency generator would not create an unanalyzed condition as its operation is transparent to the electrical distribution system. An unreviewed safety question was not identified, so the temporary operating instmetion for manual starting of the emergency generator was implemented under 10 CFR 50.59.

5.7 Modification Request 131, Emergency Generator Auto-Start Repair.

Description:

The emergency generator auto-start sequencer was replaced with a modern solid state controller.

Safety Evaluation: The safety analysis describes the emergency generator, associated bus transfer and essentialloads supplied. The upgrade of the auto l start sequencer was determined not to be a change to the facility or its operation as described in the Safety Analysis and Technical Specifications for the Ford Nuclear Reactor and the new auto start sequencer had no impact on the generator's ability to supply the identified essentialloads so the change out was implemented under 10 CFR 50.59.

5.8 Modification Request 132, Install Air Filter and Upgrade the FNR Damper Cylinder Air Line.

I l

Description:

The reactor air line to the common pneumatic cylinder which operates the supply and intake dampers was changed to copper and an in-line air filter was installed.

Safety Evaluation: The safety analysis described the purpose and operation of the  !

pneumatic cylinder and the reactor air system. Niether the material change of  !

the reactor air piping nor the installation of a new air filter was determined to I change the facility or its operation as described in the Safety Analysis and Technical Specifications for the Ford Nuclear Reactor nor to be an unreviewed

)

safety question so the change was implemented under 10 CFR 50.59.  ;

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i 18 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998

6. RADIOACTIVE EFFLUENT RELEASE Quantities 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 4'Ar Releases l Gaseous effluent concentrations are averaged over a period of one year.

l Quantity l Unit l l l

l l a. Total gross radioactivity. 3.56x 10' Ci

b. Average concentration released. 1.03x10-7 pCi/ml

c. Average release rate. 1.12 pCi/sec

d. Maximum instantaneous concentration during Not pCi/mi i special operations, tests, and experiments. Applicable '

e. Percent ofAr ERL (Effluent Release Limits) 1035 Percent (1.0x 10-8 pCi/ml) without dilution factor.

f. Percent ofAr ERL with 400 dilution factor. 2.59 Percent 6.2 Radiohalogen Releases

a. Total iodine radioactivity by nuclide based upon a representative isotopic analysis. (Required if iodine is identified in primary coolant samples or if fueled experiments are conducted at the facility). Based on this criteria, this section of the report is not required. The analysis is based on primary I coolant activity following one week of decay.

Iodine-131 was not identified in the one week count of the primary coolant samples.

Xenon-133 was not identified in the one week count of the primary coolant samples.

The pool water analyses show no indication ofleaking fuel.

b. "' Iodine releases related to steady state reactor operation (Sample C-3, main reactor exhaust stack).

l Quantity Unit l

1. Total "'I release. 245 pCi

2. Average concentration released. 1.13x10 2 Ci/ml 0.56 Percent

3. Percent of "'I ERL (2.0x10- Ci/ml) without dilution factor.

4. Percent of "'I ERL with 400 dilution 0.00141 Percent factor.

19 Revised May 1999

Report of Reactor Operations i

Ford Nuclear Reactor January 1 - December 31,1998

c. Radiohalogen 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 C-2 stack radiohalogen releases.

Br-82 4,977 pCi I-123 3% Ci I-125 499 Ci I-131 2,757 Ci Hg-203 79 Ci

2. Average concentration released. .

i l Quantity Unit l Br-82 3.52x10-" pCi/ml  ;

I-123 2.16x10' pCi/ml l I-125 3.17x10 I pCi/ml I-131 1.95x 10-" Ci/ml  !

Hg-203 5.58 x 10- Ci/ml

3. Percent of ERL without the dilution factor.

Br-82 0.70 Percent I-123 0.01 Percent I-125 1.06 Percent I-131 9.74 Percent Hg-203 0.06 Percent

4. Percent of ERL with factor of 400 dilution factor.

Br-82 0.00176 Percent I-123 0.00003 Percent I-125 0.00264 Percent I131 0.02435 Percent Hg-203 0.00014 Percent 20 Revised May 1999

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' Repon of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 l d. Total Facility Release of Radiohalogens.  ;

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l 1. Total facility radiohalogen releases.

Br-82 6,797 Ci I-123 18,223 Ci I-125 6,312 pCi 1-131 69,510 pCi Hg-203 1,779 pCi

2. Average concentration released.

i Br-82 1.60 pCi/ml I-123 '4.30x 10 " pCihnl I-125 9.63x 10' Ci/ml I-131 1.06x 10* pCi/ml Hg-203 4.20x 10- Ci/ml

3. Percent of ERL without the dilution factor.

l Quantity Unit l Br-82 0.32 Percent I-123 0.21 Percent I-125 3.21 Percent I-131 53.02 Percent Hg-203 0.42 Percent l TOTAL 57.18 Percent

4. Percent of ERL with factor of 400 dilution factor.

Br-82 0.00080 Percent

, I-123 0.00054 Percent I-125 0.00802 Percent I-131 0.13254 Percent Hg-203 0.00105 Percent TOTAL 0.14295 Percent I

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21 Revised May 1999 l

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Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 6.3 Particulate Releases Particulate activity for nuclides with half lives greater than eight days.

a. Total gross radioactivity. F376 Ci

b. Average concentration. 5.87x10* Ci/ml

c. Percent of "'I ERL (1.0x10-i2 Ci/ml) 58.75 Percent without dilution factor.

d. Percent of ERL with 400 dilution 0.147 Percent factor.

Gross alpha activity is required to be measured if the operational or experimental program could result in the release of alpha emitters.

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e. Gross alpha radioactivity. Not Required I l l 6.4 Liquid Effluents No radioactive liquid effluents were released from the facility in 1998.

6.5 Accident Evaluation Monitoring The accident evaluation monitoring program for the Ford Nuclear Reactor facility consists of direct radiation monitors (TLD), air sampling staticns located around the facility, and selected water and sewer sampling stations.

a. TLD Monitors TLDs located at stations to the north (lawn adjacent to the reactor building),

northeast (fluids), east (Beal Avenue), south (Glazier Way), and west (School of Music) of the reactor facility are collected and sent to a commercial dosimetry company for analysis. The values reported have a deploy control TLD subtracted. Background (UM Botanical Gardens) has not been subtracted from the TLD values.

Annual Quarterly  :

Total Mean i Location Direction (mrem) (mrem) i FNR Lawn North 37.7 9.4 l Fluids

• Northeast 40.9 10.2 Beal

• East 31.8 8.0 Glazier Way South 25.7 6.4 i School of Music West 27.0 6.8 Environmental Control i (UM Botanical Gardens) 19.7 4.9  !

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• The 2"' Quarter Beal dosimeter was not processed with the rest of the TLD's and had no control subtracted.

• The 3"d Quarter Fluids dosimeter was flagged by the processor for low l

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energy photon exposure (<100 kev).

1 22 Revised May 1999

I Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 Background is taken at a distance in excess of one mile from the reactor at The University of Michigan Botanical Gardens. None of the readings for the indicator locations were statistically distinguishable from the background readings (Student's T-Test).

b. Dust Samples Five air grab samples are collected weekly from continuously operating j

, monitors located to the north (Northwood Apartments), east (Industrial and  ;

Operations Engineering), northeast (Laundry), south (Institute of Science l and Technology), and west (Media Union) of the reactor facility. Each filter j sample is counted for net beta activity. There are 47 samples included in this  :

report for each location. Gas proportional counter backgrounds have been  !

subtracted from the concentrations reported. Environmental background (University of Michigan Botanical Gardens) has not been subtracted from the mean radioactivity concentrations shown below.

I Mean Station Description Concentration Unit Northwood (N) 1.54x 10"4 Ci/ml Industrial and Operations Engineering (E) 2.69x10 44 Ci/ml l Media Union (W) 2.51x104 ' Ci/ml 44 Institute of Science and Technology (S) 2.49x10 Ci/ml Laundry (NE) 2.45x1044 Ci/ml Environmental Control (Background) 3.00x10 44 Ci/ml The result of air sampling expressed in percentages of the Effluent Release Limits are shown below.

Percent Station Description ERL Unit Northwood (N) 1.54 Percent Industrial and Operations Engineering (E) 2.69 Percent Media Union (W) 2.51 Percent Institute of Science and Technology (S) 2.49 Percent Laundry (NE) 2.45 Percent Environmental Control (Background) 3.00 Percent

c. Water Samples No Radioactive liquid effluents were released from the facility in 1998.

d. Sewage Samples No Radioactive liquid effluents were released from the facility in 1998.

23 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998

e. 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 l operations from direct radiation exposure, exposure to gaseous effluents, and exposure to liquid effluents:

1. Direct radiation exposure to such an individual is negligible since a survey of occupied areas around the reactor building shows I insignificant radiation dose rates above background from the reactor. ,

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2. Airborne Effluents The airborne effluents from the reactor and the contiguous laboratory facility are as follows:

Total %ERL  % ERL Isotope Release Concentratio Undiluted Dduted (pCi) A (pCi/ml)

Ar-41 3.56E+07 1.03E-07 1034.58 2.59000 Br-82 6794.4 1.60E-11 0.32 0.00080 Hg-203 1779.9 4.20E-12 0.42 0.00105 I-123 18222.5 4.30E-11 0.21 0.00054 I-125 6312.4 9.63E-12 3.21 0.00802 I-131 69510.1 4.06E-10 53.02 0.13254 Gross Particulate 375.9 5.87E-13 58.75 0.15000 i

TOI'AL 1150.51 2.88295 Equivalent Radiation Dose (mrem) 1.44 I The total airborne effluent releases are well within the allowed release concentrations when the conservative dilution factor of 400 is applied.

The equivalent total dose fr,m all airborne effluent releases is well below the 10 mrtm per year constraint described in NRC Information Notice 97-04," Implementation of a New Constraint on Radioactive Air Effluents."

3. Liquid Efflueras No radioactive liquid effluents were released from the reactor and the contiguous laboratory facility in 1998:

f. If levels of radioactive materials in environmental media, as determined by an environmental monitoring program, indicate the likelihood of public intake in excess of 1% of those that could result from continuous exposure to the concentration values listed in Appendix B, Table 2,10CFR20, 24 Revised May 1999

Report of Reactor Operations Ford Nuclear Reactor January 1 - December 31,1998 estimate the likely resultant exposure to individuals and to population groups and the assumptions upon which those estimates are based.

Exposure of the general public to 1 ERL would result in a whole body dose of 50 mrem. The maximum public dose based on airborne and liquid effluent releases of 2.88% ERL is 1.44 mrem. This dose is based on a member of the public being continuously present at the point of minimum dilution near the reactor building.

6.6 Occupational Personnel Radiation Exposures Two hundred and five facility personnel were provided personal radiation dosimeters. Individuals for whom extremity monitoring was provided received TLD ring dosimeters for each hand. No radiation exposures greater than 50 mrem were received at the facility by individuals under the age of 18. There was one declared pregnant female at the facility from June 26, 1998 through October 2, 1998.

A summary of whole body exposures for the year is as follows:

Estimated Whole Body (DDE) Number ofIndividuals Exposure Rance (rem) in Each Rance No measurable exposure 157 Measurable exposure:

less than 0.10 28 0.10 - 0.25 10 0.25 - 0.50 6 0.50 - 0.75 3 0.75 - 1.00 1 1.00 - 1.25 0 Greater than 1.25 0 i

Total 205 i Maximum individual whole body exposure: 0.75 rem Total Effective Dose Equivalent Facility total " deep" whole body exposure: 7.29 rem Total Effective Dose Equivalent Mean " deep" whole body exposure: 36 mrem Total Effective Dose Equivalent I l l

25 Revised May 1999