ML20237C817

Ro:On 980521,discovered That argon-41 CAM Exceeded Argon CAM Alarm Level.Caused by Sudden Increase in Gaseous Level. Reactor Was Shutdown & Subsequent Analysis of Air Particulate Filter Determined Presence of Fission Products

ML20237C817
Person / Time
Site:	University of Texas at Austin
Issue date:	08/14/1998
From:	Wehring B TEXAS, UNIV. OF, AUSTIN, TX
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9808240186
Download: ML20237C817 (60)

Text

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NuclearEngineering TeachingLaboratory (512)471-5787 FAX (512)471-4589 NQ w, Dr. Bernard Wehring Nuclear Engineering Teaching Lab Pickle Research Center University of Texas at Austin 10100 Burnet Road, Building 159 Austin, Texas 78758 Document Control Desk U. S. Nuclear Regulatory Commission Washington, District of Columbia 20555 August 14,1998 Docket 50-602

Subject:

Fuel element leak (May 1998)

Dear Sir,

4 Please find enclosed a report of a fuel element leak and effluent release of gaseous fission products. This report is provided as a courtesy and is not a reporting requirement of the license, technical specifications, or regulations.

Leakage of fission products from TRIGA fuel elements are not a common occurrence.

Records indicate that two leakage events (both aluminum clad fucD have occurred during the 25 years of operation of the University of Texas license R-92 (Taylor Hall). The fuel burnup for  ;

the R-92 operation was 26.1 MW-days. The leak event in May was the first cladding failure of a stainless steel clad element at the University of Texas. Total burnup of stainless steel clad fuel elements in the licensce's inventory has been 37.6 MW-days. The leaking element fallnre was a weld defecting new element.

A design basis condition for the TRIGA-type reactor is a clad failure of a single fuel y element with peak fuel burnup conditions. This event is an infrequent fault. As an infrequent fault, this event can be expected to occur a few times during the operable lifetime of a facility.

Street Address: 10100 Burnet Road Austin, Tgas 78758 Mail Address:JJ Pickle Research Campus Bldg.159 Austin, Texas 78712 9808240186 990814 PDR ADOCK OSOOO602L 8 PDR

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Evaluation of the upper limit for the release demonstrates that the fission product gas release did not exceed the limits of 10CRr20 (1302 (2) ii) for effluents. As a consequence, the leak was judged to be a " minor" not a "significant" failure of the fuel clad.

Sincerely, 6 erat.o.ud Bernard Wehring Director Nuclear Engineering Teaching 1.ab CC: T. L. Bauer J. P. lamb D. E. Klein J. M. Sanchez J. C. White A. Adams B.Thurgood  !

T. Viviola i

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i Report of Fuel Element Failure May 21,1998 (IFE #10808) l Docket 50-602 NRC License R-129 l

! Event Summary The following report documents the activities of a series of unusual events that began Mr.y 4 and ended May 21, Identification and recovery from these unusual events was not complete until June 24, with installation, test and acceptance of a new Instmment Fuel Element (IFE #10708). A fission product leak event occurring on May 21,1998 from the installation of a new IFE (#10808) was found to be the cause of several unusual i

measurements by the argon-41 gas monitoring system. Two evaluations were made regarding reporting of the event to the US Nuclear Regulatory Commission.

The University of Texas TRIGA reactor experienced a fuel element failure event on Thursday, May 21,1998. Event detection was made just prior to the completion of.

l- three hours of operation at 500 kilowatts. At 1634 on May 21 a sudden increase in the gaseous activity (noble gases) at the argon-41 Continuous Air Monitor (CAM) exceeded the argon CAM alarm level. The reactor was shutdown and subsequent analysis of an air particulate filter determined the presence of several fission products. Two previous l similar but less significant events on May 7 and May 8 had been observed, but efforts to detect fission products had been unsuccessful. A conservative analysis of the argon CAM I

release data for the event on May 21, which was more significant then either of the

! preceding events, provided an upper bound to the release activity of 0.2 curie. Most all j the activity was the noble gas isotopes of krypton (88, 89 and 90) and xenon-138. The location of the leak was found to be a new Instrument Fuel Element (IFE #10808) first placed in the reactor core in May. Installation and operation of the instrument element was coincident with the first argon CAM alarm. The initial operation tests and results of argon CAM measurements on May 7 and May 8 were considered as possible fuel element I failure indications but no direct evidence was detectable. Following the May 21 event the reactor was shutdown and not operated. Subsequent physical inspection of the new

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instrument fuel element found the evidence of gas bubble releases from an area of the i

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upper weld seam. No additional operation of the reactor was necessary to comple the final determination of the element identification and location of the leak.

One evaluation was the classification of the event as an emergency. One classification of the site Emergency Response Plan (PLAN-E) and procedures is for a ,

Notification of Unusual Events if there is a " single fuel element failure" with " release of radionuclides into operations area" At the initiation of the event the guidance of the Emergency _ Response Plan steps Cl through ClI were taken, although the event at the initiation point did not release radionuclides into the operations area or exceed the radiological action limits of the plan. In fact the release of radionuclides into the

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operations area was after steps Cl through C9 of the plan had become effective. The remaining steps, Cl2 through C16, excluding the immediate notification requirements of C13, were done as necessary. An evaluation was made of the reporting requirements, procedure step Cl3. An analysis of reporting requirements in the license Technical Specifications 6.6.2.2 identifies as a reponable occurrence i

6.6.2.2e. Abnormal and significant degradation in reactor fuel, or

< cladding, or both, coolant boundary, or confinement boundary, (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both.

Prior to the May 21 a notification by telephone had been made to notify the UT NRC project manager of the possibility of a fuel failure condition. A follow-up notification to the project manager's office was made the day following the May 21 event with ,

l confirmation of the leak. A conclusion was made after evaluating records of the release

! data that the event did not exceed the conditions of TS6.6.2.2e since there was no i significant exposure to personnel and all constraints for exposures to the public (10CFR 1302(2) ii) were met. These conclusions allow classification of the event as a " minor leak". No report is necessary as a requirement of the license for reporting a minor leak.

I i Another evaluation of the event was made to determine the applicability of l 10CFR21. The possibility that failure of a new instrument fuel element was the source of .

1' unusual argon CAM measurements was a consideration at the first sign of unusual conditions at the CAM. Upon verification that a failure of the cladding of the new instrument fuel was the cause of the argon CAM alarms a review was made of the Page 2 of 20

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requirements of 10CFR21, Reporting of Defects and Noncompliance. The purpose of part 21 and a definition important to the evaluation states in part, 21.1 (b) that the facility, activity, or basic component supplied to such facility or activity contains defects, which could create a substantial safety hazard to immediately notify the Commission of such failure to comply or such defect, unless he has actual knowledge that the Commission has been adequately informed of such defect or failure to comply.

" Substantial safety hazard" means a loss of safety function to the extent that there is a major reduction in the degree of protection provided to public health and safety for any facility or activity licensed.

Notification by telephone of the discovery of a failure in the new instrument fuel element was made to the manufacturer on the day of the detection of the fission product release. Notification by telephone was also made to the UT NRC project manager's office the following day. A follow-up notification was made to the DOE who owns and purchases the fuel. Subsequent conversations with the fuel  !

manufacturer, regulator, and owner were made to identify the leak location with visual observations made during an inspection of the element prior to installation.

A determination was made from the analysis of the leak event that this event did not represent a " . major reduction in the degree of protection". The radiation exposures during the entire event did not exceed the part 20 limits. Further evaluation of the event in reference to fuel fabrication quality, however, is a significant concern.

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'O 4 floom 1.104 Ventilation Systems l Two ventilation systems (see Safety Analysis Repon section 7.2.2 Reactor Bay Ventilation Design, Appendix A) provide for the control of air in room 1.104, the reactor bay. Both systems are independent of other building ventilation systems.

l Main ventilation l Wntilation of room 1.104 that contains the reactor core and pool structure is controlled with two dedicated fans. The two fans, a supply fan and an exhaust fan, confine the air flow path through the room by the application of negative pressure control relative to adjacent areas. No special air filtration is provided in the flow path besides typical filters for equipment protection and personal comfort. Isolation dampers in the flow path are provided to isolate the flow path in the event of a release of radionuclides into the room. Other features such as weather seals at access doors serve to minimize leakage of room air in the event the fans are not in operation. Total air volume in the room is at least 4150 cubic meters. Recirculation of room air occurs when the reactor is not operating. During reactor operation, the ventilation system exhaust fan maintains an air exchange rate that exceeds 2.5 room air changes per hour without any air recirculation. The design ventilation rate is two air changes per hour. The typical ventilation rate is 6900 cfm (195 cubic meters per minute) which exceeds the design rate.

Argon ventilation l

Air sources adjacent to the reactor, including experiment areas such as beam L ports, ventilate by way of the argon ventilation system. The argon ventilation system controls the release of the radioactive isotope argon-41. High efficiency paniculate filters j prevent the release of other radioactive particulate material from any of these areas. A primary argon production area is the air in one or more of the five pons for reactor beams. A sixth argon ventilation pon is available to connect to experiments that are in the reactor pool. Most of the ventilation volume, however, is taken from the pool access area

_ (about half) and the reactor room (about half). The ventilation of the pool access area (approximately 110 cubic feet) controls the release of argon-41 from the pool water while the reactor room ventilation serves to dilute the moisture in the exhaust air. Argon ventilation exhaust flow rate at the argon vent stack is about 1365 cfm (9.9 cubic meters per minute).

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Aireffluennimquitors Two radioactivity monitors operate during reactor operation to detect release of radionuclides in the event of a fuel element failure or release of radioactivity from experiment facilities (see SAR section 9.5 Evaluation of Monitoring Systems, Appendix B). An air paniculate CAM with a flow rate of 60 liters / minute collects room air particulate matter on a filter paper to record the radioactivity accumulation on the filter.

An alert level at 4000 cpm warns of non-routine conditions with an alarm level set at 10000 cpm. At the alarm level a signal will automatically isolate the reactor room by shutdown of the ventilation fans and closure of special isolation dampers. Measurements q of the gaseous release of the radioactive noble gas argon-41 is made with a gas detector that monitors the stack ventilation efiluent prior to discharge at the roof stack. Sample flow rate at the CAM is 3.2 cfm. An alen level of 2000 cpm and an alarm level of 10000 cpm warn of unusual radioactivity levels at the gas CAM. No automatic shutdown occurs

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I for the alert or alarm set point of the argon gas momtor.

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Precursor Events Several unusual operating conditions were apparent prior to the confirmation of an efiluent release of gaseous fission products. Although the possibility of a leak was considered from the first unusual argon CAM measurements, there was not a positive conclusion until several days later.

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May 4 - FT#L Thermocouple fa.ihgg One instrument fuel element, FT#1 (IFE #5283) thermocouple, failed to return to a normal shutdown reading at the conclusion ofreactor operation on May 4,1998. An inspection of the reactor Instrument Control and Safety (ICS) system archive file was done prior to restart of the reactor. The inspection concluded that the thermocouple was no longer reliable arid that the other two thermocouple of the element were also not operable. Instrument fuel element, FT#2 (IFE #5982), was already operating with a single functional thermocouple. Replacement of the instrument fuel element, FT#1, would be necessary since no functional thermocouple were available in the present element. The instrument fuel element FT#1 (IFE #5283) was declared inoperable. Preparations were made to replace the element with a new instrument fuel element, IFE #10808.

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o s' l May 7 - Test ofinstrument Fuel Element (#10808) FT#1

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A new thermocouple element was placed in the reactor core grid position B-6 l

(FT#1). Fuel inspection measurements and surveillance tests were made of the new instrument element thermocouple (procedure MAIN-5 and SURV-1). Thermocouple resistance between the two leads was 85 - 90 ohms. Ambient instrument fuel temperature was recorded, and a test was made of the fuel temperature scram set points. An operation I i

test of the instrument fuel element was made to compare fuel temperatures. The  ;

instrument fuel element in core position B-2 (FT#2) increases from the ambient pool temperature of 20-22 *C to a power operating temperature of 380 *C at 950 kilowatts. At j initial operating conditions,950 kilowatts, the temperature of FT#1 (#10808) was 433 C and FT#2 (#5982) was at 370 *C. At one minute into the operation the temperature of FT#1 was decreasing to 403 C and FT#2 was at 380 C, a steady-state condition. At two minutes into the operation, the temperature of FT#1 was at 334 C and FT#2 was at 380 C, a steady-state condition. At the two minutes point of the test run the temperature of FT#1 had dropped approximately 100 C, but was at a steady-state condition.

An argon CAM alarm (>10000 cpm) was noted at the reactor console at about the  !

1 same time that the steady-state condition of the FT#1 was apparent. A peak count rate of 45400 cpm was recorded. The operator (a Senior Reactor Operator) terminated reactor operation to evaluate the argon CAM alarm (procedure OPER-5) and the FT#1 temperature response. The argon CAM alarm condition is not a direct technical  ;

I specification limit or an Emergency Plan action level.

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The argon CAM alarm event was of short duration, with an initial count rate of 1400 cpm 30 seconds prior to the alarm. The peak value of 45400 cpm was down to 36800 cpm 15 seconds aner the alarm indication prior to any change in reactor power l

level. About ten seconds later the reactor was shutdown, approximately 45 seconds aAer l the initial argon CAM alarm. Following the argon CAM transient, there was no other

! indication of unusual response from the argon CAM.

l Evaluations made aAer the reactor was shutdown found no direct indication of the presence of fission products, which was the first concern. No increase of the particulate .

CAM count rate was apparent during reactor operation or aAer reactor shutdown. The room particulate CAM average reading prior to reactor operation was 800-900 cpm. A l t

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test was made of the air region at the pool surface by using a flexible tube to sample the air with the room particulate CAM. The test efTort did not increase the ambient count rate of the room particulate CAM.

Inspection of the recent argon CAM record was done without any specific conclusions. Several previous argon CAM transient events that were unresolved were a possible cause. These unknown events were thought to be rare (less than two per 6 months) but did not cause argon CAM alarm conditions to occur. Inspection of the argon CAM system found a cooling fan for the system electronics operating at partial speed and another fan not operating at all. A temporary external fan was set up to cool the system.

May 8 - Test ofinstrument Fuel Element (#10808) FT#1 Experiments scheduled for Friday May 8 were used to further evaluate the events of May 7. Two separate experiments each for two minutes at a reactor power of 100 kilowatts were done to measure thermal neutron flux by irradiation of foils. The experiments were an opportunity to check the fuel element temperatures and reactor

. conditions at a test power level prior to an increate to full power for a subsequent experiment. All conditions were normal during measurements of the excess reactivity at 50 watts before each two-minute foil irradiation and during each two-minut;: experiment irradiation at 100 kilowatts.

A third experiment with the Pneumatic Transfer System was scheduled to operate

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at full power (950 kilowatts) to irradiate short half-life materials. A check was made of I

the experiment system reactivity at 50 watts and the fuel temperature indications at the operating conditions of 950 kilowatts. Initial response of the argon CAM and other parameters, excluding the new thermocouple device, were within normal values. The new instrument element (FT#1) thermocouple reading was 40 - 50 C lower than the other instmment fuel element (FT#2). Approximately six minutes after reaching the operating power of 950 kilowatts the argon CAM count rate rose to 18000 cpm, exceeding the alarm set point. Inspection of the CAM record noted that the high level did not follow the same pattern as the previous event by appearing to be a transient event.

The count rate levels at the peak did not return as quickly to pre-event levels. ,

1 Instead, the level following the peak remained above normal operating values. A decision l was made to stop the experiment and change the reactor power level to 500 kilowatts to Page 7 of 20 L________________ _ _ _ _ _ _ _ _ -

examine the response of the argon CAM. At 500 kilowatts the argon CAM read about 4000 cpm. The count rate level of 4000 cpm, on a log scale, at 500 kilowatts was not considered abnormally high although the level may have been above normal values. Use of the alignment valves in the argon ventilation system was made to examine the possibility that the cause of the argon CAM count rate was a source in one of the beam ports. Different alignments of the purge system valves were examined, both increasing and decreasing the argon CAM levels. No unusual results were found. A check of the pool surface air was made by a test suction of air from the pool surface area with room particulate CAM. No change was apparent in the paniculate CAM count rate.

The reactor operation at 500 kilowatts did not identify the source of the argon CAM alarm at 950 kilowatts. No identifiable radioactivity sources were observable at 500 kilowatts. The test of the valve alignments and operation at 500 kilowatts took 50 minutes of operation. Valve alignment of the argon ventilation system was reset to the normal condition. Reactor power level was again increased to 950 kilowatts. At 950 kilowatts the argon CAM count rate was 5000 to 7000 cpm. This reading was above i normal but not excessive, and did not peak above the set point as in two previous occurrences at 950 kilowatts. An operation of the reactor at 950 kilowatts for 15 minutes was done to observe the argon CAM response. No sudden count rate increases were apparent on the argon CAM.

Following shutdown of the reactor, the argon CAM count rate once again exceeded the alarm level. The immediate operation history of the reactor at power was 15 1

minutes at 950 kilowatts,50 minutes at 500 kilowatts and 15 minutes at 950 kilowatts.

The argon CAM count rate rose to 12000 cpm then, within I to 2 minutes the count rate i was at 4000 cpm. At the conclusion of reactor operation an increase of 1500 cpm during a period of time of about 15 minutes was apparent on the room particulate CAM chart l record. The final particulate levels were 2500 cpm, which is below the alert level for the particulate CAM.

Two follow-up measurements were made to test for the presence of fission products. One test was an immediate check of a filter in the argon CAM detection line to 1

identify the presence of the shon half-life Rb-90 (2.7 minutes). A gamma spectroscopy measurement was made to detect the 831 kev energy gamma ray peak from Rb-90. This l Page 8 of 20 j

I test of the argon CAM filter was negative and should have been negative since the sample point and filter are located at the exit of the argon ventilation high efliciency air filter. A second test was made of the room particulate filter to locate fission products.

This test was a qualitative gamma spectroscopy measurement made of the room air particulate filter after removal from the particulate CAM approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> after the operation at 950 kilowatts. Gamma rays present in the sample at the time of measurement and after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay were those of naturally occurring radionuclides decay chains.

e May 12 Reactor Operation Prior to continuing operation of the reactor, an additional air particulate monitor was installed to supplement the detection and measurement of air particulate material at the pool water to air interface. The supplemental air particulate monitor was to provide immediate sample collection and analysis of the particulate material at the pool surface access point. This supplemental CAM would facilitate the detection of a fission product release from the instrument element extension tube as well as a release from the general pool surface region. A two inch fixed filter and an air flow rate of about 8.0 cfm provided particulate collection for measurement with a shielded thin window GM tube. Alert and alarm set points were as set previously,2000 and 10000 cpm. A test of the linearity of the CAM with count rate was done to assure proper response as a function of the source I radioactivity. No calibration was made of the efficiency, although it was probably 10 to 30 percent.

On May 12 the reactor was operated at 500 kilowatts for about 30 minutes without any unusual air monitor readings. The reactor was then operated at 950 kilowatts for 30 minutes without any unusual air monitor readings. Argon CAM count rate was 4800 - 5000 cpm, with peaks as high as 7500 cpm. These readings were not indicative of unusual conditions. Peak values of 7000 to 8000 cpm are sometimes present as air bubbles from the reactor core burst at the pool surface. Previous experiments with these bubbles had shown that they were argon-41 (and not nitrogen-16, for example). The rate of count rate increase of the argon CAM was consistent with the irradiation of the element argon. The room particulate monitor value was also normal, reading 800 - 900 cpm. Supplemental data from the spare particulate CAM at the pool surface access point was less than 500 cpm and did not indicate any significant change relative to reactor Page 9 of 20

I power level Reactor operation was terminated by the action of a software watchdog SCRAM.

On May 13 two reactor experiments were done at 100 kilowatts. Each experiment began with a reactivity check at 50 watts, and then the reactor was run for 30 minutes to measure neutron flux by activation of test foils. All operating parameters during the experiments were normal. At 100 kilowatts the fuel temperatures were 99 C for FT#1 and i16 C for FT#2. l

, instead of continuing operation at 950 kilowatts, a series of experiments at 500 kilowatts were planned to monitor the response of the room air particulate CAM, argon gas CAM, and supplemental particulate CAM. The pneumatic transfer system was put in place for neutron activation irradiation. A reactor operation on May 20 proceeded for 1

nearly three hours at 500 kilowatts without any unusual results. The following day another operation of three hours at 500 kilowatts was in progress when the argon CAM again exceeded the alarm set point.

Release Event - Fission Product Release May 21 An event timeline was prepared following the event ofMay 21 to summarize the precursor events, and the fission product release detection, including follow-up evaluations (see Appendix C).

Opsrating history Reactor operation on May 21 began at 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> (CDST). A Senior Reactor Operator was at the controls for all operations for the day. An excess reactivity measurement was made at 50 watts. Operation at 500 kilowatts began at 1343 hours0.0155 days <br />0.373 hours <br />0.00222 weeks <br />5.110115e-4 months <br />.

Normal operation with the pneumatic sample transfer system proceeded for the next two hours and fifty minutes. Two persons were in the reactor bay at beam port three to make test measurements with a prompt gamma activation analysis system.

Abnormal or unusual event

. At 171 minutes into the run, the argon CAM count rate began to increase, exceeding the alarm set point. In less than 30 seconds, the argon CAM count rate began an increase from 5000 cpm to nearly 3,000,000 cpm, then began to decrease, returning to less than 100,000 cpm one minute after the transient began. Analysis of the reactor i

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console archive recc rds found that the time dependence of the argon CAM transient had the approximate shape of the normal probability distribution, with a full width at half maximum of 21 secor.ds:  !

l The senior reactor operator at the controls requested assistance from the reactor supervisor. At this time the room air particulate CAM was continuing to read normal levels. In the following two to three minutes, the reactor supervisor noted the room air i particulate level as unchanged, inspected the argon CAM record for the transient peak value, and inspected the supplemental air paniculate monitor for indications of radioactivity at the pool access point. The radioactivity detection circuit of the supplemental particulate monitor was no longer functional. Evaluation of the record j indicated that the particulate count rate from the pool surface had rapidly risen to 10,000 cpm. The detection circuit exceeded the alert level and failed at the point the CAM reached the alarm level. Sample collection on the filter was still functional since the j supplemental particulate CAM pump was still operational. A ponable thin window I pancake style frisker was available nearby to check for the presence of radioactivity on 1

the filter. The reactor was immediately shutdown by manual SCRAM upon verification with the frisker that the supplemental CAM was accumulating radionuclides at the pool access point. This decision took less than three minutes from the start of the initial argon CAM alarm. Later analysis of the supplemental particulate CAM found a short circuit in I

the alarm bell circuit, which caused failure of the unit's electronic circuits at the alarm .

level. -

A review by the reactor supervisor and senior reactor operator was made of the events. Notification was made to the two experimenters in the reactor room to leave the room. Confinement of the ventilation was initiated by actuation of the manual system isolation switch on level one of the reactor bay. An arrangement was made to retrieve the supplemental CAM filter and analyze the filter for fission product activity. The room air L particulate CAM began to increase five to ten minutes after shutdown of the reactor and  ;

the two ventilation systems.

Evaluatien of air particulate material on the filter from the pool access point found fission product gamma rays of the rubidium isotopes 88-90 and cesium isotope 138. Noble gas activities that correspond to these particulate materials would be krypton Page i1 of 20 I

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isotopes 88-90 and xenon isotope 138. The respective half-lives of the krypton isotopes are 2.80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />,3.16 minutes and 32.3 seconds. Xenon isotope 138 has a half-life of14.2 1

mmutes. Data for selective fission products list the krypton and xenon isotopes and l respective decay isotopes (see Appendix D).

Ef11uent release Accident evaluations of the Safety Analysis Report (see Table 11-7 Doses from Fission Product Releases page Il-31) indicate that the design basis conditions of a fuel element leak in air will not exceed the regulatory limitations of 10CFR part 20. The conditions of the unusual event were a representation of the design basis condition in the case of shon half-life materials that were in radioactive equilibrium. Since the leak was an element at the maximum power density and the release was one or more bubbles that l rose to he pool surface, the release condition was similar to the design basis accident analysis case for short half-life radioisotopes. The release quantity, however, was certain to be a fraction of the available gaseous fission product inventory in the fuel element gap.

l Two estimates of an upper limit for the fission product release were made from

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1 the reactor archive records. The first estimate assumes a constant release at the peak l

l value for the duration of the " puff." A second calculation applies the integral data of the l

argon-41 counting channel to integrate the total release. Data from the argon CAM provided information about the fission product noble gases in the release. Although the argon CAM is not calibrated for fission product gases, the sensitivity of the CAM to make such a measurement is noted in the Safety Analysis Report (page 9-15). Table 1.0 and Table 2.0 list the physical data for the radioisotope argon-41.

1 Table 1.0 3 1

! Argon-41 Physical Data Element Ar A 41 Z l8 Decay mode B-AR1000 CAM Calibration 1

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Sensitivity 2.5 x 10E8 cpm / (microcurie / cubic centimeter)

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Table 2.0 Argon-41 Decay Data Ikcay mode T]/2(mtm) Tl/2(txt) Radiation Modifier IXrad) E(end pt) 1(rad) Ikse S 6.56N0E+03 109.34 M 12 B- 459.5 3 1198.6 7 99.10 0.970 B- 6.56040E+03 109.34 M 12 B- TOT 464.5 3 99.982 20 0.989

& 6.56040E+03 109.34 M 12 G 1294. 99.10 2.73 Data for the beta energy of the radioactive argon-41 isotope indicates an average beta energy of 464.5 kev. Argon CAM sensitivity to beta radiation is a function of the detector propenies, a 50 by 0.4 millimeter scintillation detector, and the electronic

~ discriminator settings. Sensitivity to fission product noble gases relative to argon-41 will change since the beta energies of the isotopes of Kr-88, Kr-89, Kr-90 and Xe-138 are different, with average total beta energies that are respectively 363,1312,1365 and 625 kev. Argon CAM calibration sensitivity can be expected to increase significantly with the substantially higher average beta energies of the fission product gases. The one beta energy that is less than the beta energy for argon-41 is also the least detectable of the four fission products ofinterest. This isotope also represents the longest half-life of the of the four precursor noble gas isotopes, by more than a factor of ten.

A calibration factor of 2.5 x 10E8 cpm / (microcurie / cubic centimeter) for argon-41 detection by the argon CAM will be a conservative estimate for the fission product gas concentrations. A review of reactor console records following the event found that most of the radionuclides release was in a period of 20 seconds. At the argon ventilation flow rate of 1365 cfm a total of 455 cubic feet (-13 cubic ineters) of air will exhaust from the room in 20 seconds.

L An initial estimate of the release was done by assuming a constant release for 20 seconds at the peak value observed by the argon CAM. Release time was 20 seconds, with a subsequent exhaust air volume of 13 cubic meters. By rounding the available argon console data to one significant figure, a peak release rate of 3,000,000 cpm was chosen. At the argon-41 conversion factor of 2.5x10E8 cpm /(microcurie per cubic Page 13 of 20

d centimeter), the puff release was 0.16 Curies. A second calculation of the release quantity was taken from the integral argon count channel of the argon CAM. The integral argon total of the argon CAM includes the total of argon-41 and the fission product gas releases. Total release of argon and fission products fiom the integral data was 2,068,118

. counts during reactor operation. Using an argon-41 count rate at 500 kilowatts, including background of 2500 cpm for 170 minutes, determines the argon-41 component of the integral to be 475000 counts. The estimate of the fission product integral is then 1.6 x 10E6 counts. Applying the same release time and volume exhaust, a release estimate of 0.26 Curie is made. Both calculations depend on the conservative assumption that the argon CAM sensitivity is the same for argon-41 and all the fission product gases. The calculations are shown in Appendix E.

Summary of SAR Doses 1 Page 11-31 of the Safety Analysis Report summarizes the possible onsite, room doses, and offsite, public doses that might occur in the event of a fuel element failure in 9

air The puff type release of May 21 was similar to the SAR calculation in several ways.

A fission product release of one or more bubbles of gaseous or volatile fission product elements was the cause of the release. The release source was a maximum power density fuel element. Short half-life isotopes with a half-life less than 30 minutes are near maximum concentration. The release scenario was a pufr(or gaseous release in air with minimal decay time) of mostly short half-life noble gases. Measurements of transit times for gas bubbles from the reactor core to reach the exhaust or monitoring point range from 38 to 50 seconds.

The effluent release was from a fuel element operating at the maximum reactor

. core power density sufficiently long to have the short half-life (less than 30 minutes) fission product elements in equilibrium. Therefore the short time exposures will be l similar to the accident analysis. In contrast, however, the puff release represents a reactor power level at 500 kilowatts which is one-half the accident analysis value and a fraction of the total available gaseous or volatile fission product inventory in the fuel element.

Some release delay occurs during the transit time from the fuel element to the pool surface, and then to the roof exhaust, and finally, to the street level. Decay, dispersion and elevation of the release all contribute to lowering the actual measurable dose offsite.

Page 14 of 20 l l

l i

I

r The SAR integral dose at 6 minutes is 2 millirem onsite and 0.2 mrem offsite.

l Onsite the dose assumes no removal of the gas. Offsite the dose assumes a rate of ten air changes per hour. The short duration of the event and the flow path for the release requires some further analysis. The argon ventilation system effectively removed the puff at the pool surface to the building exhaust stack with no accumulation in the reactor room i until the ventilation system was shutdown. The pool surface volume that exits the stack is 110 cubic feet (3.14 cubic meters). A flow rate of 650 cubic feet per minute (28,000 liters per minute), or half the total system flow rate, removes the air at the pool surface. This

! exhaust provides a pool surface ventilation rate of 350 air changes per hour. Dilution of I the exhaust flow mixes with a room dilution factor of two for the argon ventilation -

system, and 5.1 for the main ventilation system, for a total dilution factor of 10.2. No l provision is made for the effective exhaust height in decay and dispersion calculations.

Building Extemal doses l

Calculation of the offsite dose was done to determine whether the fission product release was a minor leak, not immediately reportable, or significant leak, requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification of the event. A calculation was made following the methods of page 7-  ;

27 of the SAR; Averaging the release concentration for a period of two and one-half minutes then finding the efTective one-hour dose, an estimate of the upper limit for an unrestricted area exposure was 0.16 millirem in an hour. The 10CFR part 20 limit for unrestricted areas is 2.0 millirem in an hour (10CFR 1302(2)ii). The estimate is well below the limit 2.0 millirem per hour without correcting for detection sensitivity, decay, gamma ray energy and other factors such as distance.

L The site health physicist provided further verification of potential offsite doses. A source term assumption was made of 0.2 curie for each of the isotopes Kr-85m and Xe-135. Respective fractions of the annual limits for the release were 0.018 and 0.027. The

calculations are included in Appendix F.

Building internal doses Following the release event, an estimate was made of the possible Kr-85 and Xe-135 doses prior to entry into reactor bay, room 1.104. A calculation was made using the

, release quantity of 0.2 curie and assuming release into the room instead of exhaust ventilation. The actual concentrations of the two isotopes of krypton and xenon would be Page 15 of 20 u________________________ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l substantially lower than the estimate since the 0.2 curie was a measurement of short half-i life radionuclides, half-life ofless than 30 minutes. Respective one hour doses for each case were found to be 2 and 3 millirem. Buildup ofgas in the room did not occur until at least 10 minutes after the release began No indication was found of the presence of radioactive gas in the room air the following day. A sample of the air was taken with the argon CAM. The argon CAM readings 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> after the release were at the normal background value of 25 cpm.

j Event Rfcovery Reentry into room 1.104 for routine work was done the following day after analysis of the room air paniculate filters. Two measurements were made prior to restart of the room ventilation system in the non-reactor operation mode (partial air recirculation). One measurement was a gamma spectrum of the room air particulate CAM filter from the previous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Another measurement was a replacement of the filter and verification that the room air particulate concentrations were typical of values preceding the release event. A grab particulate sample with a portable air sampler was f also taken to evaluate the airborne conditions.

Several follow-up measurements of filters and water samples were done to further document the presence or absence of radionuclides. Airborne radioactivities decayed and

' dispersed. Fission products in the pool water were accumulated in the ion exchange resin tank of the pool water purification system. One day after the release a gamma spectrum of one liter of pool water taken for 5 minutes found the fission product gases Kr-85m (4.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />), Xe-133 (5.3 day) and Xe-135 (9.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) as the primary radioactivity. A similar i I measurement taken for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> three weeks after the event found no obvious fission product isotopes as the sample primary radioactivity.

One liquid spill event during the inspection of the leaking fuel element on May 28 led to a surface contamination area. A vent ofliquid from the fuel element was created by changes in pressure with a liquid spill volume of one to two cubic centimeters occurring when the element was above the pool surface. A cleanup effort of approximately 9.5 man-days was necessary to restore clean conditions to pool area surfaces that came into

- contact with water leaking from the element. Radioisotopes in the spill were Zr-95, Mo-99, Tc-99, Ru-103, I-131, I-132, Te-132, Ba 140, La-140, and Ce-141.

Page 16 of 20 l

~

1 Argon-41 CAM Evaluation Normal operation Operation of the argon CAM detects radionuclides from reactor beam pons, the pool access area, and the level one area of room 1.104. Valve alignments can aid in the detection of a radioactive airborne source. In the case of changes to the valve alignment, the ambient release rate of argon-41 will change. Previous experiments had found that alignment of beam port I and 5 valves could produce substantial changes in the argon-41 exhaust rate. Another source of argon-41 is the release of bubbles of air at the pool surface. Air bubble releases were noted to increase the argon CAM count rate as much as a factor of two or more on some occasions, but the releases are all transient with a negligible effect on the time average release.

Transientjensitivity Several measurements were made to verify the presence of argon-41 in the air bubbles being seen at the pool surface. In particular a check was made of the approximate half-life to eliminate the presence of nitrogen-16 (7 second half-life). A few transient events on the CAM at conditions inconsistent with a radioactive gas release and with an unknown explanation were known to be present. Evaluation of all such events during 1996 and 1997 was done to evaluate the CAM operating performance. A complete review of all operating logs found a total of 35 events that were evaluated. Of the total of 35 transient type evems,19 were events with a known cause such as short operation or pulse operation. Eleven events were of unknown cause, and four were most likely fission product release events with the final event being that of May 21.

Unknown events were further categorized into three groups. The groups were four during reactor startup, five following reactor shutdown, and two during low power operation. Five of the argon CAM transient events had magnitudes ofless than 500 cpm.

Four transient event magnitudes were between 500 and 2000 cpm. The largest two events were 2500 cpm and 3000 cpm.

Four events that were present after installation of the new instmment fuel element were probably all indications of fission products. The magnitude of the four events in the Page 17 of 20 l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ ._ __ _ ._ N

order of occurrence was 300 cpm,45000 cpm,6000 cpm, and the fmal event of4 million epm (chart recorder value).

A review of the CAM record indicates that all eleven unknown transient events were between 6/20/97 and 5/21/98. A significant repair, including modification of the L

argon CAM electronic circuit, was done by the manufacturer 6/97. No similar events were found on chart records between 11/95 and 1/97.

[ Fission product vs. arnon noble nas l, . (Refer to Efiluent Release and data in Appendix G) l f

Instrument Fuel Elements (#10808)

Technical Specifications for the reactor require two protective action channels j

that measure reactor fuel element temperature. Fuel temperature sensors in instrument type fuel elements are to be located in the B or C ring of the reactor core. The Instrumentation, Control and Safety (ICS) System is set to provide a protective action or SCRAM if the fuel temperature reaches 550 *C. Technical Specification requirement (TS3.2.3) for the set point is less than or equal to 550 C.

Each fuel element contains three temperature sensors that are chromel-alumel, Type K thermocouple, one of which functions as the sensor for the protective action function. Two temperature sensors in one instrument fuel element can be used as the two protective action channels, although the preference is to use two separate instrument fuel elements for better channel to channel independence between the protective action functions.

The university has five instrument fuel elements in its fuel element inventory.

These fuel elements are element serial numbers 10808,5982,5981,5283 and 2990. Fuel element #10808 is a new element received by UT 11/96 and loaded in the core 5/7/98.

Elements 5982 and 5981 were new elements received by UT prior to 4/70. Element l numbers 5283 and 2990 are transfers with previous operation histories.

Element #5981 was loaded into a previous reactor core in 1970 and element #

l 5982 was loaded into the present reactor core on 10/95. The 5283 element was refurbished and operated in the UT reactor from 2/11/92 to 5/7/98. Element 2990 is an Page 18 of 20 L-_-______________________-_____-____

l . element with a small (0.5 inch diameter) extension tube. The conditions of the thermocouple in this element are not presently known.

Fuel element #10808 is a GA element with cast end fittings to maximize the cooling water flow at the element ends. Two such elements were requested to replace i

failing thermocouple sensors that each element contains for measurement ofreactor fuel temperatures. Fuel is made available through a federal University Fuel Assistance Program. Projections of fuel needs and fulfillment of requests is dor + as allowable by available funds. One instrument element was received in Nov.1996. A request for a second element is pending availability. A replacement element #10708 has been obtained from Kansas State University to return the UT reactor to an operable condition.

Acquisition of a replacement element was done by tl:e DOE with receipt at UT 6/17/98.

Acquisition of another spare instrument fuel element remains a goal for future department fuel acquisitions.

Visual inspection of element #10808 during December of 1997 found one significant weld artifact. The inspection found several dark pits in the weld zone where the upper cast fitting joins the fuel cladding can. The most pronounced pit was inspected with a binocular microscope. Pit appearance showed a dark colored deposit in the bottom of the pit. A digital photograph of the weld feature was sent to the element manufacturer for evaluation. Review of test records and review of the photograph by the manufacturer did mt identify a fuel element quality problem. The manufacturer regarded the pits as art'Jucts of the Weld process. Examination of the instrument fuel element (#10808) after the fuel leak event of May 21 found a source of air bubbles (fission product gas) at the approximate location of the weld artifact that was found prior to installation of the element in the reactor core. Movement of the fuel element from a depth of approximately i

24 feet to a depth of 6 feet led to the release ofgas bubbles from the element 5 days and 7 i

' days after removal of the element from the reactor core. Total operation history of the reactor core with element #10808 in the reactor was 4866 kilowatt-hours. Estimating that l an element in the B-ring produces 1.8 percent of the total power of the core, the power production of the instrument fuel element was 86.8 kilowatt-hours. This energy production is equivalent to a loss of about 4 milligrams of uranium-235. The U-235 fissile inventory for the instrument element is 39 grams.

Page 19 of 20 u_ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - . - - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Table 3.0 IFE #10808 Operating liistory Date liours Kilowatt-hours' 5/7/98 1,40 58

! 5/8/98 2.20 976 i

5/12/98 1.50 822

, 5/13/98 1.52 51 5/20/98 3.76 1503 5/21/98 3.13 1456

Total 13.61 4866 k '

. Note: Burn-up estimate of #10808 is found by multiplying the total operating history by the power production factor 1.6 for the B-ring and dividing by the total number of elements in the reactor core. (#10808: 0.018 4866= 86.8 kwhr) i I

l l

I l

Page 20 of 20 l

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .. __ _ - u

SAR 5/91

~

Appendix A 7.2.1. Reactor Pool and Shield Structure Pool water system and shield structure desig,n combine to control the effective radiation levels from the operation of the reactor. One goal of the design is a radiological exposure constraint of 1 mrem / hour for accessible areas of the pool and shield system. Dose levels assume a full power operation level of 1.500 megawatts (thermal). Radiation doses above the pool and at specific penetrations into or through the shield may exceed the design goal. Figure 7-1 displays the basic design dimensions of the pool and shield system with some of it's features.

Representative 1 arem/ hour dose curves for a reference case design are shown in Figure 7 2. The reference case design is a solid structure without any system penetrations. Design of the reactor pool was of 1/2 inch (1.27 cm) base plate and 1/4 inch (9.64 cm) wall plate of 6061 aluminum alloy. Tank assembly is by shop fabrication. A protective layer of epoxy paint and bitumen coal tar pitch with paper provide a barrier between the aluminum pool tank and the reactor shield concrete.

A four foot (1.22 meter) thick foundation pad supports the reactor pool and shield structure. Standard weight concrate, 150 lb/ft3 (2.33 g/cm3), comprises the foundation pad. High densf.j concrete, 180 lb/ft)

(2.89 g/cm3), with a magnetite aggregate is the shield material of the first level of the shield structure. A transition from his density to standard density concrete is present about 4.5 feet (1.4 m) above the mid level platform of the shield. The top part of the shield stem and the top level platform are standard density concrete. The total shield weight is 2.03 x 10' lbs. (920 metric tons). Approximately 24,400 lbs.

(11,100 kgs.) of structural steel, 56 conduits for signal and alectrical lines with diameters of 1/2 inch to 3 inch, three central junction boxes I and numerous local junction boxes are part of the shield system.

Five beam tubes at the level of the reactor provide experimental access to reactor neutron and gamma radiations. Two of the tubes combine to penetrate the complete reactor pool and shield structure from one side to the other side. <

I Special design features of the beam tubes are beam plugs, sliding lead shutters, bolted cover plates, and gasket seal for protection -

against reactor radiation and coolant leakage when the tubes are not in (

use.

{

l 7.2.2. Reactor Bay Ventilation Design Ventilation design is specified to control air confinement and to isolate the reactor bay in the event a radioactive release is detected in the reactor areas. The ventilation system is designed to maintain a negative pressure within the reactor bay with respect to the building exterior and other building areas. Confinement and isolation is achieved by air control dampers and leakage prevention material at doors and other room penetration points. A separate system is designed to exhaust air from several locations within the reactor ' bay that could contain airborne radionuclides such as argon-41. Manual operation of start /stop controls of both main and purge air systems will be available in the reactor control room.

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SAR 5/91 Ventilation of the reactor bay is provided by two modes of system operation. One mode is for standard operation with recirculation of air. The other mode is an exhaust operation with high volume flow that has no air recirculation. Design during exhaust mode operation is a rate of air exchan6e in excess of two per hour. Total volume for the room exceeds 4120 cubic meters. Normal operation of the ventilation system uses a roof stack for the exhaust of air from the reactor bay.

Air filtration in the ventilation system is to be of typical design for normal HVAC operation with no special provisions. Schematics of the ventilation system for the reactor bay area and a logic diagram of the ventilation control system sensors and controls are provided in Figure 7-3.

Control of air confinement within the reactor bay is provided by di f fe rential pressure control between the reactor bay and a representative ambient external measurement point. Additional measurement points in ventilation zones adj acent to the reactor bay maintain the differential pressure between the reactor bay and adjacent access areas. The differential pressure control is intended to function in both standard and exhaust operation modes of the ventilation system.

Isolation of the reactor bay is provided by ventilation dampers.

These dampers will shut in response to either manual or automatic signal actuation. An automatic signal will initiate shutdown of the ventilation system by closure of the dampers if a set point for airborne particulate radioactivity exceeds a setpoint. Protective switches within the ventilation system will cause the air fans to respond to the position change of the dampers. Damper design is for fail-safe operation so that loss of control power will isolate the reactor bay.

Dampers locations are in the vicinity of the duct penetrations into the reactor bay. An isolation damper is in each of two supply air ducts.

One return air duct with two sections contains two isolation dampers, one in each section. A pair of return air ducts also contain a damper in each duct.

The separate air purge system is designed to exhaust air that may contain radionuclides products by a low volume system. The primary nuclide of interest is argon 41. Figure 7 4 shows a schematic of the argon purge system and its control logic. Air from potential sources of neutron activation such as beam tubes, sample t rans fer systems, and ,

releases from exchanges at the pool water surface are subj ec t to confinement and isolation by the system. Flit rat inn of air in the l system will include prefilter and high efficiency particulate filter, j Design provisions allow for the addition of charcoal 111ters if j experiment conditions should require the additional protection. Sample -

ports in the turbulent flow stream of the purge system exhaust provide for measurement of exhaust activities.

Actuation of the isolation damper in the argon purge system is by manual operation of the fan l

[ control switch.

A schematic of each ventilation system is shown in Figure 7-5.

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SAR 5/91 9

both the Anmain exhaust and stack purgeon the roof combines the ventilation exhaust of systems. Mixing occurs at the exhaust exit l point. The exhaust stack will extend 14 feet (4 roof level. The effective release point above the,.24 meters) above the calculated from the equation exhaust stack can be {

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I 4 - height of plume rise above release point, m, D - diameter of stack, m,

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- mean wind speed at stack height, m/s, Vs - effluent vertical efflux velocity, m/s.

The following values were used for this calculation D - 0.46 m, y - 4 m/s, l

V, - 12.2 m/s, 1 then (12.2 m/s)

Ah - 0.46  !

4 m/s i Ah - 2.2 m.

Ground elevation in the area is 794 feet. Roof elevation at the stack is 843 feet. Therefore, the effective release point is at least 60 feet above the maximum ground elevation at the building.

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SAR 5/91 7.4. 2. Evaluation of Arron 41 Release The release of argon-41 from the facility is diluted by the ventilation exhaust rate, assumed to represent two air changes per hour, and averaged for a 5 day, 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> operation schedule at full power. The release concentration from the pool averaged for one year is,

.24 (2.12 x 10-8) - 5.1 x 10*9 pCi/cm 3 Only 20% of the experiment facility argon-41 is assumed to exhaust since experiments will replace some or most of the exposed air,

, 6 (.20) (.24)/2.29 x 10 -6 1.3 x 10*7 pC1/cm3 Total estimated release is 1.3 x 10'7 pCi/cm 3 The whole body gamma ray dose rate to a person immersed in a semi-infinite cloud of radioactive gases can be approximated by D - 900 EAD (29) where E -

the photon energy, 1.3 Mev AD - effective exposure concentration, Ci/m 3 I The concentration downwind from the point at which the activity is l discharged from the building is i AD - Aq p(x), (30) where 6 -

the dilution factor at the distance x, (sec/m 3),

A -

activity concentration in the discharge (C1/m3 ), l q - the building exhaust rates (m 3/sec).

If it is assumed that the discharge is at the roof line, the dilution factor in the lee of the building (x - 0), is given [6] by:

p(0) - 1/csu ,

(31) where c - a constant (0,5),

s-buildingcrogs-sectionalareanormalto direction (m ),

the wind I

u - wind velocity (m/sec).

A minimum cross-sectional area is assumed of 234 m 2 (60 x 42 f t) and, for a wind velocity of 4 m/sec, p(0) - 1/(0.5 x 4. x 234) - 2.1 x 10'3 sec/m 3 (32) l 7-27 l

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• SAR 5/91

, , o The averaged dose rate at the exhaust stack is D - 900 x 1.3 (1.3 x 10~7) - 1.5 x 10~4 rads /hr, an average of 0.15 mrad /hr in the stack or D - 1.5 x 10*4 (2.3 x 2.1 x 10*3) - 7.2 x 10~7 rads /hr, an average of 0.72 prad/hr at ground level.

l At the limiting exhaust rate, 640 cm3/sec, to ignore the argon 41 decay, the source term is 27 pCi/sce for the beam ports which would

!. 4 Increase the dose rate 4.5 times to 3.2 prad/hr. If core experiment facilities, such~as,the center tube and rotary specimen rack, are vented at the - same exhaust rate the releases would increase by 102 pCi/sec.

L However, these exhaust rate conditions represent limiting conditions, not actual release rates. ' For the exhaust manifold rate of 3.15 cm3/see the release rate would be 50 pCi/sec for the two core experiment facilities. The rotary specimen rack is the ' primary source of the -

activity.

l .

Venting of experiment facilities, especially the rotary specimen rack, will require monitoring of the release rate or replacement of the air with gases such as nitrogen or carbon dioxide to control release j: concentrations of air activity. However, normal operating conditions do not vent the rotary specimen rack, although this is an optional operatin5 condition. The pneumatic transfer system, by comparison, routinely contains nitrogen or carbon dioxide gas to limit the releases within the room that contains the access terminal.

Actual dose values for the argon-41 release may vary. The beam port release estimate is less than 4.8 prad/hr which is equivalent to 10 mrad /yr, Lower neutron fluxes, smaller air volumes, shorter operation times and larger dilution factors will assure that releases do not exceed annual release constraints. Monitoring the exhaust will verify that 'other release points such as the core experiment facilities do not cause the total to exceed preset limits.

i i

l 7-28 i

__ _ - - - - _ - - - - - i

I

• SAR 5/91

,, ., Appendix B 9.5. EVALUATION OF MONITORING SYSTEMS l

The radiation monitors provide information to operating personnel of impending or existing hazards from radiation so that there will be sufficient time to take the necessary steps to control the exposure of personnel and the release of radioactivity or to evacuate the facility.

Three types of radiation monitors are used: a continuous air particulate monitor for determining radiation levels due to particulate radioisotopes suspended in the reactor room air, a continuous air gaseous monitor for determining radiation levels due to argon-41 in the

' room air, and area radiation monitors for determining the gamma field at several locations in the facility.

Each type of radiation monitor has a specific radiological purpose. The particulate air monitor is used to detect radioisotopes released due to fuel element failure (a design basis accident). The gaseous air monitor is used to determine the effluent radiation release of argon-41. Argon, a component of air (.04% by volume) may be activated to produce argon-41 in potentially significant quantities.

Finally, the area radiation monitors are used to minimize personnel radiation exposures. The radiation monitors in section 9.5.1., 9.5.2.

and 9.5.3. are typical instruments at the time of original installation.

Replacements may have slightly different characteristics.

9.5.1. Particulate Air Monitor Set points for the particulate continuous air monitor warn of the presence of particulate fission product nuclides. Since gaseous and volatile elements such as krypton, xenon, bromine, and iodine have particulate decay products, the presence of some of their radioisotopes should also be detected. An alarm set point at 2000 picoeuries/ milliliter detects particulate activity concentrations at the occupational values of 10 CFR 20 for 70% of the relevant isotopes in the ranges84-105 and 129-149. These ranges of isotopes represent the one percent yield for fission products of uranium-235. Significant fission products as a percent of total release are shown in Table 9 1.

The air monitor in use is a Ludlum Model 333-2 beta air monitor, configured for continuous sampling of airborne beta-emitters. It uses i two standard pancake G-M tubes, each having a 1.75 inch effective j diameter. The two detectors are arranged in line so that gamma j background subtraction is performed. This increases the accuracy of the j beta count. The 333-2 will accept air flow rates ranging from 10-100 liters per minute.

Particle accumulation on a fixed filter continues at a constant l

rate. Activity on the filter, however, is a function of the air flow l rate, filter collection efficiency, and the decay rates for nuclides l j that are present on the filter. If one assumes that the source in the j room is constant the activity at the filter will be the accumulation  !

term minus the decay term and will have the same functional form as the activation equation, 9-10 l

... SAR 5/91

,, ,a' dNg -q.

- (N r.

(1) dt V with the solution, q <N r Ng(t) - (1 e*AE), (2)

V A l

lwhere,.

. Ng - number of atoms on the filter, Nr - number of atoms within the room, q/V, c, A - facility and material constants.

Figure 9-3 represents a plot of equation 2 for the particulate monitor with typical background conditions and the following assumptions:

0 6.5 x 10 cm 3/ min q/V - - 1.35 x 10 6/ min .

4.83 x 109 cm 3

A - (In2)/(2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) .346/hr .

e - 98 percent .

At equilibrium, the ' saturation condition determines the number of atoms. Fora filter count rate .of 2000 cpm with a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> half-life isotope the results are 2000 cpm 2hr 60 min 5 m,

.35 in(2) hr on filter i.

And -'

l q (N r

l. Ng ( t- =) - -

v A Nr./c 9.9 x 105/.98 ( 346/hr)

' Nr " A~ >

(q/v) (1.35 x 10-6/ min) (60 min /hr)

[ - 4.32 x 10 9atoms or .89 atoms /cm3 .

i At this concentration t filtegcount rate of 2000 cpm corresponds to an activity of 2.5 x 10"ge pCi/cm for a two hour half-life isotope.

9-11 L________-_-__-_______-_________-__________-_____-_--_-_ . _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ -

SAR 5/91 Table 9-1 I Significant Fission Products Contribution to Total Activity, Percent l

Element Is:* ope 1 Day 10 Days 30 Days 90 Days 1 Year B*enerny (MeV)

Strontium Sr 89 2.8 6.7 10.5 2.7 1.49 Sr 90 1.8 .546 Yttrium 90 Y

2.29 91 Y 3.4 7.6 12.5 3.9 1.55 Strontium Sr 91 6.7 1.1,1.38,2.68 i 92 Yttrium Y 4.2 3.62 93 Y 7.6 2.89 Zirconium Zr 95 3.7 8.2 14.7 7.3 .366,.398,.888 Niobium Nb 95 4.1 18.2 15 .159 Zirconium Zr 97 9.0 1.93 Niobium 97 Nb 9.6 1.28 Molybdenum Mo 99 4.6 6.8 1.23 Rhodium Rh 103m 2.55 5.5 7.0 Ruthenium Ru l03 2.65 5.7 7.2 .225 Rhodium Rh 105 1.35 .566,.25,.26 Rh 106 3.54 Ruthenium Ru l06 2.4 .0394 131 Iodine 1 6.8 3.7 .606,.25,.81 132 1 2.7 5.3 .80 133 1 7.3 1.27 Tellurium Te 132 2.6 5.1 .23 Xenon Xe 133 1.23 11.4 2.6 .346 Iodine 135 1 4.7 1.0,.5,1.5 Xe 135 I Xenon 12.5 .91 f Barium Ba 137" Cesium Cs 137 1.5 .512,1.173 I Barium 140 Ba 1.25 10.6 10.8 1.6 1.01,.47 ,

Lanthanum La 140 12.0 12.5 2.4 1.36,1.25,1.68 Cerium Cel '1 6.3 11.2 8.5 436,.581 Lanthanum La 141 1.4 2.43 Praseodymium Pr 143 10.0 11.2 1.9 .932 Cerium 143 6.8 Ce 1.09 1.39 lM Ce 2.0 6.0 26.5 .316,.182 Praseodymium Pr 144 3.00 Neodymium Nd 107 4.8 4.1 .804,.364 Promethium Pm 109 1.45 1.072 9-12 L______________________ 1

l SAR 5/91 I

I

' Activity Release (althorne concentration) 1 R.9 e

1.5 i -~ ~ _ _ _ _

a t.7

\ 'f

.7

'~

1 v.5 '\ ', /

. \/

C f X l.

i a

c.2

/ \ N I t / N l.2

N

~

l .

R.1 .

/ ~

P t

l' l, 3: ll o I, 1, l' gi Time Cometant Retto, (ventilation rate)/(decay rate)

Activity Accumulation (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> half-life) 1-R . 9-

[

,. /

e ,

l 1.5 /

l e j t . 7< /

1 /

l v.& /

l e 8

. 5-t r.+

a c.>

t 1 3<

e ,

n .1< /

. I e 1E 2E 3E 4E 55 6E Time (half-lives) l l

ACTIVITY ACCUMULATION ON PARTICULATE FILTER Figure 9-3 9-13

SAR 5/91 If a sudden appearance of an additional nuclide or group of nuclides of similar half life occurs, an additional curve similar to the background condition will occur. Cases are shown in Figure 9-3 for 2 hr, 8 hr, 2 dy and 8 dy half lives.

To determine the count rate set point, a, on the particulate air monitor, the following equation is used:

a - aVtn, (3) where, a = alarm level, cpm a a reference activity concentration, dis / min-cm3 q = air sampling rate, cm /3 min t a particulate accumulation time, min n a ' efficiency of the detector, cpm /(dis / min)

The reference activity concentration may be expressed as follows:

a - (2 x 10~9 pCi/cm3 ) (3.7 x 100 ) (60 sec/ min) ,

pCi dis / min

- 4.44 x 10'3 .

3 cm The following data pertains to the Ludium Model 333-2 particulate air monitor, 0

q -

6.5 x 10 cm /3 min ,

n - 0.3 .

An electronic circuit monitors the detector chamber flow rate to assure adequate flow rates. Particulate in the air flow path accumulate on a filter with 1% efficiency for 0.25 micron particles.

The value of the detector efficiency is a conservative estimate based on ,

the beta energies of interest. Using a Tc-99 source, the detector efficiency is rated at 36% for a beta spectrum with 0.3 MeV energy range. Since most of the isotopes of interest have higher beta energies than Tc-99, values of detector efficiency will exceed 30% in most cases.

In fact, efficiencies of 50% may be applicable for many isotopes, thus assuring a conservative detection limit for all except a few fission product isotopes.

ll l

9-14

(- _ _ _ _ _ _ _ _- __ _ _ _ - _ _ _ ______-_______-_ _ ___ - ____ _ - _ .

SAR 5/91 l

Monitor position to sample reactor room air is within 5 meters of the pool at the pool access level. The location will sample air activity in the vicinity of the reactor pool. Leakage - of fission l products from the fuel into the water then into the room would occur at l the room air to pool water interface. Background measurements of air particulate activities between Sept. 90 and Sept. 91 provide a record of the naturally occurring count rate levels for the Ludlum Model 333 2 in room 1.104 of the NETL facility. These data indicate that count rates of 4000 cpm to 6000 cpm will occur several times each year as a result of weather conditions that effect vertical air stability such as, frontal lines, temperature inversions and storm systems. A set point at 5000- cpm will provide an alert level with an occasional alert for a natural occurrin6 condition.

The count rate alarm set point will assume beta energies of 0.3 and detector efficiency of 30%. The set point may now be calculated:

dis / min 0 3 a - (4.44 x 10'3 3

) (6.5 x 10 cm / min) (120 min) (0.3) {

cm

= 10,400 counts / min.

Refer to Table 9 1. Most of the isotopes listed in the 1 day column have beta energies greater than 0.5 MeV. Since the detector efficiency increases with the incident beta energy, a more representative estimate for the detector efficiency may be 50%. The set point may be calculated as follows:

dis / min 0 3 a -

(4.44 x 10'3 ) (6.5 x 10 cm / min) (120 min) (0.5) 3 em e 17,300 counts / min.

A particle' accumulation time, t, of two hours tay be considered, as shown in Figure 9 3.

9.5.2. Arzon-41 Monitor Set points for the argon 41 continuous air monitor should warn of excessive radiation levels for effluent release and occupational exposure. This radiation monitor will operate whenever the reactor system and the auxiliary air purge system are operating.

As specified in 10 CFR 20, the reference concentration of argon-41 is 1 x 10"g pCi/cm . Dividing this number by the purge exhaust system flow rate and by the building wake dilution factor yields the averagg annua 1 concentration limit for release at the stack, which is 2

• 10 pC1/cm3 . An alarm set point at ten times this level, 2 x 10'5 pCi/cm3 '

will warn of an excessive daily release. In the event of a gaseous fission product release in interference will occur in the argon 41 count due to betas emitted by isotopes of krypton and xenon, refer to Table

]

9-2. '

9 15

SAR 5/91 0

Table 9 2 l

Beta-Emitting, Caseous Radionuclides of Interest i

i Reference Level Concentrations

1. isotopes in Air Isotope Yield (%) of mass N Half-Life Beta Energies  % (uci/ml) 41 Ar 0 1 109m 2.49 1% 1 x 10-8 1.20 99%

85 Kr 1.33 5 10.7y 0.687 99.57% 7 x 10~7 85mKr 1.33 5 4.48h 0.841 78.8% 1 x 10~7 87 Kr 2.56 5 76m 3.889 31% 2 x 10-8 3.486 40%

1.335 9.2%

3.044 7.1%

1.475 5.7%

88 Kr 3.7 5 2.8h 0.521 67% 9 x 10'9 2.913 14%

0.681 9.1%

89 Kr 4.8 5 3.16m 4.93 23% -

2.33 15%

3.24 10%

2.53 5.6%

133 Xe 6.77 6 5.29d 0.346 99.2% 5 x 10*7 135 Xe 6.7 6 9.17h 0.909 96% 7 x 10-8 138 Xe 6.7 4 14.2m 0.803 34% 2 x 10-8 2.82 22%

2.38 18%

2.418 12%

0.567 10%

9-16

SAR 5/91 The radioactive gas monitor in use ' is the P.R.M. Model AR 1000 l argon 41 monitor. It uses a 50mm X 0.4mm CaF1 scintillator to detect t, the betas emitted by argon-41. Detection chamber volume is 11.4 liters, i

and it accepts a nominal gas flow rate of 30 liters per minute. The

!. system automatically performs background subtraction.

f To determine the count rate set point on the argon-41 air monitor, the following equation is used:

on a -

(4) l WVf where l j

a a alert level on the AR-1000, cpm i

o a reference concentration level, pCi/cm 3 n a detector response, cpm /(pCi/cm3) p a building wake dilution factor, sec/cm 3 V = argon purge system flow rate, cm /3 min f a fraction of 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> day the reactor actually operates Refgrence egncentration level, a, at ground level outside the building is 1 x 10~ pCi/cm For'the AR-1000, the detector response, n, has been determined to be n - 1.5 x 10 8 count / min .

pCi/cm 3 The building wake dilution, p, may be calculated from the following equation:

1 p -

(5) 0.5Av where 2

A = building cross-sectional area, m v = wind speed, m/sec ,

l l

l l

1 I

l 9-17 I I I

SAR 5/91 The building cross-sectional area, A, is conservatively determined from the smallest side of the reactor building:

1 A -

234 m 2

- 2.34 x 10 6 c,2 ,

The wind speed, v, is assumed to be 1 meter per second, also a conservative value.

The building wake dilution may now be calculated:

1 6

. 0.5.(2.34 x 10 cm) (100 cm/sec) - 8.55 x 10~9 sec/cm3 The argon purge system flow rate, V, is 6.14x105 cm3 /sec (1300 cfa).

Assuming the reactor operates for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> each day, calculate f:

f -

8/24 - 0.333 .

The count . rate set point on the AR-1000 for an alert may now be calculated:

(1 x 10 8 pCi/cm3 ) (1.5 x 10 8 counts / min )

pCi/cm 3 a-3 5 3 (8.55 x 10~9 sec/cm ) (6.14 x 10 cm /sec) (0.333)

- 858 counts / min.

Since the alarm concentration level is ten times the value of the alert concentration level, the count rate set point for an alarm would be at 8580 cpm.

l 9.5.3. Area Radiation Monitors Several area radiation monitors which observe the gamma field are

! part of the permanent installation. Some locations are experiment areas in which shield configurations determine the levels of radiation during

! reactor operation. When possible, alarm set points for all area j

radiation monitors will be at either 2 mr/hr or 5 mr/hr. The first l number is obtained by dividing the maximum desired dose each week by the l

number of working hours each week. The second number is obtained from the definition of a radiation area in 10 CFR 20, 9-18

_, . . . j

SAR 5/91 A high radiation area, defined in 10 CFR 20 as having a radiation level > 100 mr/hr, may exist above the pool access area during some operations. The area radiation monitor located above the pool access area vill have an alarm . set point of 100 mr/hr. Although the doses within the pool protection railings may exceed doses of 100 mrem /hr, the dose exists 'only in the immediate area of the pool surface. At other locations of the pool shield platform level, the doses are significantly less than 100 mrem /hr, but may exceed the 2 to 5 mrem /hr range. While the reactor is operating, one area radiation monitor will operate above the pool access area, as well as at least two additional area radiation monitors located at other positions around the reactor shield and at the beam port facilities. The radiation monitor system consists of six units with GM tubes that detect dose rates from 1 mr/hr to 10 kr/hr.

9.5.4 Monitor Availability Conditions Several factors apply to the requirements for availability of the reactor radiation monitoring systems. Among these factors are the types of conditions each monitoring system detects. If one of the continuous air monitors (CAM) is out of service, reactor operations may continue for a limited period of time, provided the other CAM is operating.

Reactor operation is permitted for up to one week when the particulate air monitor is inoperable, provided a filter evaluation is performed daily, and a signal from the argon-41 air monitor is available to provide information for manual shutdown of the HVAC. This is necessary to detect the design basis fuel element leak. When the argon 41 monitor is not available, operating the reactor with auxiliary air purge system shall be limited to a period of ten days. This constraint restricts any effluent release from the reactor building.

The particulate air monitor may be inoperable due to either an electronics failure or a pump failure. If an electronics failure occurs, the filter will accumulate particles as usual. Since the flow rate is known, the radioactive particles in the filter may be evaluated daily using a portable detector. The expected drop in detector efficiency may be offset by the eight hour accumulation time. If a pump failure occurs, particles will not accumulate on the filter of the particulate CAM. Instead, a daily evaluation is performed of the radioactive particles accumulated on the filter of the argon CAM.

Evaluation of the radioactive particles on the CAM filters should occur near the end of daily reactor operation. This technique will detect the presence of those particulate fission products having half lives greater than a few hours.

While daily evaluations of the CAM filter will detect a minor, persistent fuel element leak, the signal from the AR-1000 must be monitored for changes resulting from a maj or sudden fuel element failure. This method detects gaseous fission products, such as xenon and krypton, which have rather short half lives, as shown in Table 2.

The reference level concentrations for most of the xenon and krypton isotopesofintgrestmeetorexceedthereferencelevelconcentrationof argon-41 (1x10* mci /ml); powever, the argon-41 alert concentration at the detector is 2x10- mci /ml. Because occupational level concentrations for the isotopes of interest are typically 300 times 9-19 L_____________-- __ _ - _ - _ - . - - _ - - - - - - - - - - - - - - - - -

,.4 SAR 5/91 their re ference values and because their betas should be readily detected by the argon continuous air monitor, the argon CAM alert set point can identify concentrations below the occupational levels for these xenon and krypton isotopes.

When the argon 41 monitor is inoperable, argon production and release may be calculated. Provided the shielding configuration (including beam ports) is not altered, argon production and release should not change. The release rates can ba. calculated from previous measurement data and design flow rates for air through the stack. A 10 day limit is set for the inoperable period to limit the amount of release without direct measurement. This period represents a release averaged over a year. The effluent release during a 10 day period would be about 4% of the average annua 1' concentration limit.

If the reactor is operating, at least half of the six area radiation monitors must be operating, one of which must be located above the pool access area. This number of monitoring points including the pool area detector is sufficient to warn of unusual operating conditions.

t However, some consideration would be made to assure that monitors are operable within areas of experiment and personnel activity.

l i

9-20 -

L_.m__ . __m____ ___m.... _ _ . - _ _ 1

l

• 1 l .

l ,, ., Appendix C Precursor Event Timeline 5/21/98 5/7/98 Fuel element TC #5283 fails. Temperature sporadically drops low.

(ICS System FT#1 requires replacement, Tech. Spec. 3.x.x) 5/7/98 Prepare and install new TC instrument fuel element, number 10808.

Test element ambient temperature response prior to operation.

5/7/98 Test new TC element at power operation conditions.

16.49 initial condition (0 minutes):

(ICS System at 950kW FT#1:433 FT#2: 370) 16.50 Initial condition (1 minutes):

(ICS System at 950kW Ff#1 : 403 FT#2: 380) 16.51 Final condition:(2 minutes):

(ICS System at 950kW FT#1 : 334 FT#2: 380) 16.51 Terminate run to evaluate argon CAM alarm.

5/7/98 Transient event on argon CAM.

16.51 Peak at 45400 CPM sets alarm condition.

16.51 CAM at 36900 CPM duration of peak less than 30 seconds.

16.51 Shutdown reactorby run down of rods.

16.51 CAM count rate decreases to 31200 CPM.

16.52 CAM count rate decreases to 10300 CPM.

Noincrease on room particulate CAM.

No increase of room particulate CAM by sampling pool surface area.

Inspect argon CAM, identify fan failure (s) at argon-41 CAM.

5/8/98 11.% Operate at 100 kW,2 minutes Foil measurement experiment,(tiermal flux).

5/8/98 12.01 Operate at 100 kW,2 minutes Foil measurement experiment, (thermal flux).

5/8/98 15.24 Operate at 950 kW, ~12 minutes (1524 1536), argon CAM 18000 CPM Pneumatic Transfer S31cm experiment (1 sample tien abort exp.)

5/8/98 15.39 Operate at 500 kW, ~50 minutes (1539-1629) argon CAM 4000 CPM Evaluate beam port sources by isolation of argon manifold vahrs.

5/8/98 16.32 Operate at 950 kW, ~15 minutes (1632 1647) argon CAM at 5000-7000 CPM.

Argon CAM count rates within normal range, but indicate high values.

5/8/98 16.47 Shutdown reactor, all rods down 16.48 argon CAM increases to 12000 CPM.

16.50 argon CAM decreases to 4000 CPM.

1500 CPM increase on room particulate CAM.

Evaluate previous argon CAM records for suspicious transients.

Previous transients noted 15 in 1997 and 16 in 4.5 months of 1998.

5/11/98 Install NMC particulate CAM with suction at pool surface.

5/12/98 Operate at 500 kW,30 minutes for test (1331-1401)

Operate at 950 kW,30 minutes for test (1411 1442)

SCRAM - unknown CSC WD event, history archive fails.

Also subsequent NM1000 message length error was found prior to recovery.

Argon gaseous CAM 4800-5000 CPM Room particulate CAM 800-900 CPM NMC particulate CAM 250-400 CPM 4

44~

! 5/13/98 Operate at 100 kW,30 minutes (1523-1553)

Foil measurement experiment,(fast flux).

5/13h8 Operate at 100 kW,30 mmutes (1612-1632)

Foil measurement experiment, (fast flux).

5/20/98 Operate at 500 kW,178 nunutes (1334 1632)

{ Pneumatic Transfer System experiments (25 samples).

i 5/21/98 Operate at 500 kW,174 minutes (1342-1637) l%w.4 Transfer System experiments (18 samples).

-> Argon CAM alarm... Fuel leak cat prior to shukiown!

i I

(

4 I

i I

l >

Event Timeline 5/21/98 16.30 Pre existingconditions:

SRO (MGK) at the reactor controls.

l Reactor at 500 kilowatts steady-state.

Tire escady-state power was 171 minutes.

l i Pneumatic samples were being run in core position G34.

No sample was in the core at time of the event initiabon.

l A gamma ray measurement c@ mcit was in progress at BPW3.

l. Three air activity monitoring system were in operation.

T}pical room particulate CAM count rate is 500-1000 CPM.

Typical argon (nobic gas) CAM count rate is 4500 CPM.

l NMC nominal count rate at full power is 400-500 CIM 16.34.32 Argon CAM exceeds alarm set point at 10000 CPM.

16.34.32 Argon CAM increases to 56000 CPM Room particulate CAM is -800 CPM.

16.34.45 Argon CAM increases to 2660000 CPM Room particulate CAM stays at 400 CPM.

MGK calls TLB for assistance. Two persons are in the reactor bay.

B. Wehring and K. Unlu are at BP3 area on first level of reactor boy, 16.35.24 Argon CAM dropped to 94000 CPM Room particulate CAM reads 650 CPM.

]

i 16.36.30TLB arrives to assist evaluation of conditions. j Check made of argon CAM recorder trace;  :

PulTtype transient event recorded.

Check made of Particulate CAM recorder trace; l No detectable change is apparent.

Check made of NMC particulate CAM trace; I

CAM inoperabic, trace termmates at 10k CPM.

NMC CAM shows increase to 10000 CPM, fadure at high alarm set point.

(fuse and loss of count channel power supply, probably short in alarm circuit).

NMC CAM motor and pump continues to collect air paruculases on fiber.

Frisk of NMC CAM filter indicates high actisity, >2000 CPM.

Check of room particulate CAM, icading less than 1000 CPM.

Request immediate reactor shutdown (TLB).

l 16.37.03 Manual reactor SCRAM (MGK).

16.38 TLB and MGK discuss situation, resiew events

, 16.40 Notify BW and KU to lean the room, (MGK).

I 16.42 Manual isolation of the reactor bay main ventilation system, j (includes shutdown of argon exhaust s entilation system),  !

! (all persons out of the room).(MGK) 16.44 Re-enter room to retrieve NMC CAM filter (TLB, MGK, & FYl),

}.

I 4

l

r l-I Room particulate CAM reads less than 1000 CPM.

16.50 Room Particulate CAM count rate increanng.

16.55 Positive identifxation of fissior product gamma rays on NMC CAM Alter.

(Primary gamma ray indicator was 831 kev Rb-90 4 min half-life, two secondary gamma rays also present 1032 and 1248 kev Ri>49.)

Post room as airborne radioactive material area.

Turn off argon CAM pump and integral argon cous channel.

17.14 BW enters room to secure equipmes at BP3 (~ 6 minutes).

Room particulate CAM reads 3500 CIW, still incressag, Safety office notification (recording), will make notification tomorrow.

17.15 MK enters room to shut down heat exchanger (1.5 minutes).

Electronic equipmem burning smell at NMC CAM, '

Turn off NMC electronics (power supply transforma overheats). I 18.00 Room peruculate CAM reaches 4000 6000 CPM.

Call GA to discuss fuel leakage and TC element expenence.

19.00 Room particulate CAM reaches 10000 CPM (reactor console),

Count rate at CAM remains below alarm set point (10000 CPM).

20.00 No reactor roorn ventdation to the building exterior.

Upper bound estimate made of release from console arson CAM data.

Release duration was effectively 20 seconds with a peak value of 3x10^6 CPM.

Assume a constant release at the peak value for 20 seconds.

Assume fission prochsct detection efficiency same as argon 41.

An upper bound estimate was made of the total release,0.15 Curie.

Assumes no decay, conservative efficiency and total gaseous release.

20.30 Particulate CAM stabilizes at 10000-12000 CPM.

24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to NRC rosy be necessary for puff release? (evaluate AM).

Fuel element leak ir notification of unusual event.

Dose releases do not exceed emergency action levels.

20.30 TLB leans.

21.xx Notification to health physicist (ATT).

l 21.20 MGK leaves.

l l

1 l

l I

I r + Appendix D l NUCLEAR ENGINEERING TEACHING LABORATORY NUCLEAR ANALYTICAL SERYICES

.............................. - 33s,tE ,ARxxEyExS o..o..m... ..............

Project Title sample ID  : NMC CAM FILTER Welqht 1.00000E+00 0 Acquisition Time  : 21-MAY-1998 16:43:04.16 Sample Geometry  : ALL DOWN Decay Corrected to : 21-MAY-1998 16:30:00.00 Decay Time  :

Preset Live Time : 0 00:05:00.00 Preset Real Time  : 0 00:00:00.00 Elapsed Live Time : 0 00:02:17.02 Elapsed Real Time : 0 00:02:17.09 Detector Dead Time : 8 Config File Name

! ***************************** ANALYSIS PARAMETERS * * * " * * * * * ' * * * * * * * * * * * * * * *

• Sensitivity  : 5.00000 ' Start Channel 1 Half Life Ratio  : 9.00000 Stop Channel  : 8192

..............*************** DETECTOR FARAMETERS * * * * * * * "" * " * * * """ * * * "

, Detector Name : TENNELEC40 Ave Efficiency : 40.00000 Entg Calib Dates MAR-1998 14:05:36.12 kev / Channel  : 2.68232E-01 Energy Tolerance: 2.00000 calib Offset  : -1.70342E-01

' Energy (KEV)= (-1 .70342E-01 ) + [2.68232E-01 )* Channel + (-6.33200E-08)* Channel **2

.....................................................y)

FWHM (KEV) = (6.91085E-01)+(4.91493E-02)

• SQRT(Energ Pk It Energy Area Bkqnd IMHM Channel Left Pw Cts /sec 4 Err Fit 1 0 10.62 40 69 0.66 40.21 36 9 2. 95E-01 43.6 ???

! 2 0 238.47 92 120 0.92 889.87 881 18 6.74E-01 30.6 ???

l 3 0 462.35 341 111 1.63 1725.03 1713 28 2.49E+00 9.9 Cs-138 4 0 546.39 89 67 1.28 2038.63 2033 19 6.47E-01 23.6 Cs-138

-5 0 657.40 187 52 1.79 2452.93 2443 21 1.36E+00 11.2 Cs-138 6 0 831.23 105 82 - 1.47 3101.81 3089 26 7.69E-01 22.5 Rb-90 7 0 897.63 90 34 2.05 3349.74 3340 18 6.58E-01 16.3 Rb-88 i

9 0 1009.05 243 31 1,47 3765.83 3751 27 1.77E+00 8.2 Cs-138 9 0 1031 .6 551 76 2.26 3850.26 3834 30 4.02E+00 5.5 Rb-89 l 10 0 1247.88 -331 67 2.65 4657.98 4644 33 2.42E+00 8.5 Rb-89

! 11 3 1433.33 79 15 1.64 -5351.00 5337 42 5.74E-01 22.0 1.0E+00 l- 11 3 1433.33 79 15 1.64- 5351.00 5337 42 5.74E-01 22.0 Cs-138 12 3 1435 00 157 24 2.18 5357.27 5337 42 1.15E+00 13.9 Cs-138 13 3 1436.03 143 26 2.15 5361.11 5337 42 1.04E+00 15.0 Cs-138 i

i

'" ~ ' *

- _ _ _ _ - _ _ _ _ _ _ _ _ - _ - - _ _ _ - - - _ - _ - - - - - - _ _ - - - - - - - - - _ - - - - - _ - - - - - - - - - - - - - - - - - - ----- _ '_2_'_[_~ L __ _-__ .

a Table of Fission Product Gamma Raye NMC Supplemental CAM Particulate Filter Pool Surf ace Area Sample Fuel Leak May 21, 1998 Radiation End-point Radiation Decay Rad. Energy Energy Intensity Dose A CLEMENT Z Made Half-Life Type (kev) (kev) (t) (G-RAD /UCI-H) 88 RB 37 B- 17.78 M 0.01 G 898.03 0.04 14.0 0.8 0.269 88 RB 37 B- 17.78 M 0.01 0 1836.00 0.05 21.4 1.3 0.837 89 RB 37 B- 15.15 M 0.12 G 657.77 0.06 10.0 1.0 0.140 89 RB 37 B- 15.15 M 0.12 0 1031.92 0.06 58, 6. 1.27

89 RB 37 B- 15.15 M 0.12 G 1248.14 0.06 43, 5. 1.13 l 89 RB 37 B- 15.15 M 0.12 G 2195.92 0.11 13.3 1.5 0.624

. 90 RB 37 B- 158 3 5 G 831.69 0.05 40. 3. 0.707 90 RB 37 B- 258 5 4 G 831.69 0.05 94. 6. 1.67 90 RD 37 B- 258 3 4 G 1375.36 0.03 16.7 1.0 0.488 90 RB 37 B- 258 S 4 G 2752.68 0.00 11.5 0.6 0.673 90 RB 37 B- 258 S 4 G 3317.00 0.12 14.3 0.7 1.01 A Element f. Decay mode T1/2(num) T1/2(txt) Radiation E(rad) Elend pt) I(rad) Dose 138 CS $5 IT 1.74000E+02 2.90 M 10 G X KA2 30.6251 3 11.5 4 0.0075 130 CS 55 IT 1.74000Et02 2.90 M 10 G X KA1 30.9720 3 21.3 0 0.0141 130 CS 55 B- 1.74000E+02 2.90 M 10 G 191.70 20 15.4 18 0.0628 138 CS 55 B- 1.74000E+02 2.90 M 10 G 463.00 20 18.6 22 0.184 138 CS 55 B- 1.74000E+02 2.90 M 10 G 1436.00 20 19.0 20 0.581 138 CS 55 B- 2.00460E+03 33.41 M 18 G 462.796 5 30.7 7 0.303 138 CS 55 B- 2.00460E+03 33.41 M 18 G 547.001 5 10.76 24 0.125 130 CS 55 B- 2.00460E+03 33. 41 M 18 G 1009.78 8 29.8 1 0.642 130 CS 55 B- 2.00460E+03 33. 41 M 18 G 1435.86 9 76.3 16 2.33 130 CS 55 B- 2.0046CE+03 33.41 M 18 G 2218.00 10 15.2 4 0.717 Data taken from: R. R. Kinsey, et. A1., The NUDAT/PCNUDAT Program for Nuclear Data, 9'"

International Symposium of Capture Gamma-Ray Spectroscopy and Related Tepics, Budapest, Hungary, October 1996. Data extracted from the NUDAT database, version March 27, 1999.

i I

r l

l l

1 1

a s

. g .7 l

l Radiation End-point ' Radiation Decay Rad. Energy Energy Intensity Dose A- ELEMENT Z Mode Half-Life Type (kev) (kev) (8) (G-RAD /UCI-H) 88 KR 36 B- 2.84 H 0.03 B- 109. 365.

j= 5, 14. 2.65 0.16 0.0062 88- KR 36 B- 2.84 H 0.03 B- 165. 6. 521. 14. 67 4. 0.235 88 KR 36 B- 2.84 H 0.03 B- 227 6. 681. 14. 9.1 0.5 0.0440 88 KR 36 B- 2.84 H 0.03 B- TOT 363. 11. 101. 6. 0.780 88- KR 36 B- 2.84 H 0.03 B- 441. 6. 1190. 14. 1.95 0.11 l.

0.0183 88 KR 36 B- 2.84 H 0.03 B- 678. 7 1731. 14. 1.02 0.06 0.0147 l

88 KR 36. 9- 2.84 H 0.03 B- 825. 7 2051, 14. 1.3 0.3 0.0228

88 KR 36 B- 2.84 H 0.03 B- 1136. 7. 2717 14. 2.1 0.3 0.0508

( 88 KR 36 B- 2.84 H 0.03 B. 1233. 7 2913. 14. 14. 4. 0.368 88< KR 36.B- 2.84 H 0.03 E AU L 1.680 15.2 l 1.4 0.0005 88' KR 36 B- 2.84 H 0.03 E AU K 11.40 4.0 0.4 0.0010 I . 88 KR 36 B- 2.84 H 0.03~ E CE K 12.313 0.014 10.7 1.2 0.0028 88 KR. 36 B- 2.84 H 0.03 E CE L 25.448 0.014 1.5 0.0008 0.4 80 KR 36 B- 2.84 H 0.03 E CE K 101.101 0.010 1.14 0.15 0.0044 l

l 89 KR 36 B- 3.15 M 0.04 B- 473. 22. 1273. 50. 2.37 0.18 0.0239 l 89 KR 36 B- 3.15 M 0.04 B- 555. 23, 1457 50. 1.53 0.13 0.0181 l 89 KR 36 B- 3.15 M 0.04 B- 627, 23. 1619. 50. 12.00 0.15 0.0267

, 89 KR 36 B- 3.15 M 0.04 B- 632, 23. 1629. 50. 1.56 0.13 0.0210 l 89 KR 36 B- 3.15 M 0.04~ B- 647, 23. 1662. 50. 1.99 0.16 0.0274 89 KR 36 B- 3.15 M 0.04 B- 859. 24. 2124. 50. 4.0 0.3 0.0732 89 KR 36 B- 3.15 M 0.04 B- 898 24. 2208. 50. 1.45 0.13 0.0277 89 KR .36 B- 3.15 M 0.04 B- 983. 24. 2392. 50. 14.5 1.0 0.304 89 KR 36 B- 3.15 M 0.04 B- 1076. 24. 2589. 50. 5.6 0.5 0.128 09 KR 36 B- 3.15 M 0.04 B- 1190. 24. 2030. 50. 2.99 0.23 0.0758 89 KR 36 B- 3.15 M 0.04 B- 1266. 24. 2991. 50. 2.23 0.20 0.0601 89 KR 36 B- 3.15 M 0.04 B- TOT 1375. 34. 101. 5. 2.96 l 89 KR 36 B- 3.15 M 0.04 B- 1411, 24, 3296. 50. 9.7 1.0 0.292 l' 89 ' KR 36 B- ~3.15 M 0.04 B- 1489. 24. 3460 50, 2.8 0.3 0.0888 09 KR 36 B- 3.15 M 0.04 B- 1588. 24. 3665. 50. 3.7 0.5 0.125 89 KR 36 B- 3.15 M 0.04 B- 1649, 24, 3794. 50, 1.5 0.3 0.0527 89 KR 36 B- 3.15 M 0.04 B- 1942. 24. 4404. 50. 2.2 1.0 0.0910 89' KR 36 B- 3.15 M 0.04 B- 1946. 24. 4413. 50. 4.2 0.5 0.174 89 KR 36 B- 3.15 M 0.04 B- 1985. 24, 4492. 50. 1.5 0.6 0.0634 89 KR 36 B- 3.15 M 0.04 B- 2225. 25. 4990. 50. 23. 4. 1.09 90 ' KR 36 B- 32.32 S 0.09 B- 488. 14. 1309. 17. 2.07 0.19 0.0215 90 - KR 36 B- 32.32 3 0.09 B- 923. 14. 2264. 17. 2.29 0.19 0.0450 90 KR 36 B- 32.32 S 0.09 B- 1086. 15. 2612. 17. 65. 6. 1.50

} 90 KR 36 B- 32.32 3 0.09 B- TOT 1312. 17. 103. 8. 2.89 90 KR 36 B- 32.32 8 0.09 B- 1935. 15. 4392. 17 29. 4. 1.20 l

90 KF 36 B- 32.32 S 0.09 E AU L 1.680 4.5 0.4 0.0002 90 KR 36 5- 32.32 S 0.09 E AU K 11.40 1.20 0.12 0.0003 90 KR 36 B- 32.32 S 0.09 E CE K 106.62 0.03 2.65 0.25 0.0060 A Element Z Decay mode T1/2(num) T1/2 (txt) Radiation Modifier E(rad) E(end pt) I(rad) 138 XE 54 B- 8.44800E+02 14.08 H 8 B- 129. 14 432. 40 3.07 13 0.0004 I 138 XE 54 B- 8.44800E+02 14.08 M 8 B- 155. 15 507 40 9.5 4 0.0314 l 138 XE 54 B- 8.44800E+02 14.08 M 8 B- 243. 16 743. 40 32.1 13 0.169 138 XE 54 B- 8.44800E+02 14.08 M 8 B- TOT 625. 29 100. 9 1.33 138 XE 54 B- 8.44800E+02 14.08 M 8 B- 922. 19 2320. 4020.17 0.395

' 138 XE 54 B- 8.44800E+02 14.00 M 8 B- 939. 19 2358, 40 13.4 5 0.268 138 XF 54 B- 8.44800E+02 14.00 M 8 B- 1009, 19 2512, 40 5.1 21 0.110 138 XE 54 B- 8.44800E+02 14.00 M 8 B- 1112. 18 2759. 40 5. 5 0.118 138 XE 54 B- 8.44800E+02 14.08 M 8 B- 1121. 19 2754. 40 9. 6 0.215 138 XE 54 B- 8.44800E+02 14.08 M 8 E AU L 3.550 48. 3 0.0036 130 XE 54 B- 8.4 000E+02 14.00 M 8 E CE M 3.63 5 32. 7 0.0025 130 XE 54 B- 8.44800E+02 14.08 M 8 E CE L 5.14 5 51, 3 0.0056 138 XE 54 B- 8.44800E+02 14.08 M 8 E CE M 9.63 5 10.4 6 0.0021 138 XE $4 B- 8.44800E+02 14.00 M 8 E CE K 117.873 3 1.6 3 0.0040 138 XE 54' B- 8.44800E+02 14.08 M 8 E CE K 222.426 20 1.80 10 0.0085 l-

4 SAR 5/91 O

. .. Appendix E l

Table 11 7 DOSES FROM FISSION PRODUCT RELEASES 1

MN-O enn dowmn ehm

' e.4 O'O

'O 'O O 'O 'O 'O 'O 'O 'O 'O 'O 'O 'O 'O

--- ----- --- g C.

6 M M MM M MMM

~

• MMMMM MMM u

' . @~N l . O. O. n. a. m. . . m. a. <. ~. . m. e. m. A i - se e e se W M .= ** P. W N @ WNN O .O E te ,

l

.D E h 0 .e d N u. S H

e .e >= = h m,N

==.=== O

@m e@e

-O -'O.O b. N p

= = . =n.=e. M.

O .m4W O. O. O.

OOO O. O. O. O. O. MMM OC.

wp g

?

en OOOOO 0 .

.% . e. e6 o

-NN

1. g gob

-e8 uoA U-

. p6

.a

, -O-Nm O nI.

• • N M 4 4 9 ed 5 0 0+0 0+0 .

w 0 0-..=

== 0 .O* *O* ** ee == ee == @ .* en O.%w

.S MMMMM 500

• =

• MMMMN mee . N

. m@@ O. O. O* -wW o

OOO . .

a. W C .%. e. N. se. @.

Pm* .@ N. @. m.

Q .- e e6 .e se e e e W *e .= g.

G .g w E. 6.

F .

.h E .% so e W

. d. .$. 0m

m. .=. m. e. .e. **M4m o w

== N N m N O .o ====.O . 4

. O N m .m=M == 0. O. O. .O.==.N.*. 4 ee ODO C0000 MMM .t.

66.

,O .

3 e.==..=. .6.e N .e W t. o.

a m.

8e e4 -Wwe.

3 e.4

.=

.O.008 O.O

. O. O.

O-.

. O. O. O. O. . O. O.

• S X O-..O O-. . .

I--

N N. NN N

he 9 *.

e.

F ...6.

.t

s. 98 .

U

-o M.

. .o N.. m.. N. .,, a. >, .. .

a .

e

. .. .S ...O.

.u. ~

o o . ..

'E D b .N u

c e.-

ee u

o .o. .o N N. o..o

a. c. -N N.

-N

e. ..6 5 5 u.

. u.

2 .. .~

o

. .E we e

c ,

- - M - -Nm l

l l

11-31 l

Dose Estimate May 21,1998 (Calculation see Page 7-27 UT-SAR)

The w hole body gamma dose for a semi-infinite cloud of radioacthe gas is D = 900 E Ao D = rads per hour E = photon energy Assume 1.0 MeV(conservathe) l Ao = cffective exposure concentration Assume argon calibration (conservative)

Ao = A q y

. A = activity concentration in the discharge (Ci/m^3) 2.6 x 10^-3 Ci/m^3 q = the building exhaust rates (m^3/sec.) 6.4 x 10^-1 m^3 / sec.

y = the dilution factor at the distance x4) 2.1 x 10^-3 secim^3 y=l/csu e = a constant (0.5) s = building cross section area normal to the wind (m^2) 234 m^2 u = wind velocity (m/sec.) 4.0 m/sec.

y = 2.I x 10^-3 sec. / m^3 Three minute release concentration is (2,068,i18 - 425,000) counts

= 2.62 x 10^-3 Ci/m^3 2.5 min. * (2.5 x 10^8 cpm / (microcurie /cm^3))

A = 2.62 x 1P-3 Ci/m^3 Results:

l D = 900 E A q y i

= 900 (1.0) (2.6 x 10^-3 Ci/m^3) (6.4 x 10^-1 m^3 / sec.)(2.1 x 10^-3 sec. / m^3) l

= ~3.14 millirem / hour is the rate for 3 minutes (0.05 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />)

= -4.16 millirem effective dose in one hour l

t

.1- l NETL E%% Dose Equivaled) ce%lalion inm lhe rdease of Thus. 21 hh 1998 l

I 1 Knnwn; A_ min vo\utne. (VJ - A675m ]'

Ps ~

m b%soof hes h+ (ll{) 13 'm

.. . ?X hlf. cross 'orel avea 2 34m' i

. 84 iMX>n venfilahon (en-hou/ role E (400(. s,26j

=

W'& Speed N rn/s Assumc fSvom Ar Ch1 & h d,0 sudes) Wi-dhe: l 3rimary isokf s are. ' <r- 5s%8<. Xdd35fL l

N. east anthnf Uro/rr & CA/71 cita) ~. 0<-f 6 ~9 stc.

Mo) = 7,4 X fo 62. 1

~

. . . f Y - $5 * "Vz. = 4.48 hr = 4. 3 x 10~ " s ' ~

~

Ne..- 135 Tu - 9 14 hr A = 2.11 K 10 's^

..DA{ OF Kr- 85* =

S K 10 ' (6t. lmo

~

~

of Ke - 1Ss =

suo' /3/ /m s l DR 8

De.

Dec = 2. 5 me/hv .

Ocupa9. } me ; l nr or 36 00 3ecmds msed Y_irdykt vo ume sauce lot a nohle c,es L_______ _

8 usino the. ?og 2.s mm/hr kr / dad

( echnic'$l1 y incorrect lor a linile source volukeJ one bnen can OMein a ecnversion fechv

._ ... lox _ e i44,ev @ $ the dose epuiveled rabt_ pee _

unl- c.oncenira dion (Id/Mc) dor

. equi ve lery+ per und expo.sure' (H/E or bhe @b

._ h = h _ (f.5 m Pem/hr)(&psurc hime in sec) b}c E DA C. a (a. By/yn ) .

!Thts for Xr- 86'1, we Cassumig_1 b 3ef lhr-

- . .. _ .. . ..=. 2:5 Y[MCDNY . 8 ID mrem

_E 3 S

._?x.fo? g/m .. J /m /.s M

~ ,4 h -: {2. 6 h  : j.S9 XJO m dO3 5

E 5 x 10 3

$/m $/m'b

, -u._ ___ _ - _ _ _ _

L_.____...._.___.__ - ---- -----

-3' h canAebevJmLne a renwvel rale cons % & J

(.1)__fo cescribe & reanoval of krism andioy Ye-155 froon khe. reetkr volume (V -

, d5 kilotos ; _

~'

l. 3r00in ven4i 8 ion rabj_4 A'd/udice caute 6 l

roorn volume /

M

=3,,, _

6 26 45 45M rn'

+ 4.5x10-5 s' -- 7. s c rrrf

~

4' P! .

! = _9

+ 2.11 % 10 2s': 7. 34x lo

.. .. ..?.A _-'

+ 5 ~& % *

..?bL9: St exposur e E curira a h e t Cin_..hhis case- Moos) can d cel co lelecf

_ _..As k 11quls :

HI kr %r-Ts 5 * >_assv>n;rw J

Ro)= 7 4 x 10 ' 4 ' 8 ek \ / \ i t E = {_Mo) \( h \/l \

A __ A )4(1- e ft 1 AV)

N//$0ebudk u . u~ ~s s l Em f co)(1 - e

~

)h a k m-t 3 '

l 1ii .

Av i

.- f. -

he base.

ep(i v6l.e n t N can khen be .

ob h ed assoning Y . 4 0 o ' /S gom o.z c' 4 =

h-0A bhl E { _. ._\

l f . . . . -- . . . . - - . - . . . . . . . .

I y ..

~

_.71l?X_.10~.'._n9*n Y 1JAmwin

_ ._z.co e.. x..io j' 5 E

. p hn %> .... _ . .

C '

_*g h b_h ._

-9

, . _ .1, .'b 9.X IO ; a ------- - -

. , .. . ...  ? =...h9?_K 30. ir i?Tf5l5%_1D(m@n.:- --2. -85 inw>n

(

_ . _ .pmM.j.___.___

6 mp.D

_ pg m. p .

y ,

l . , _ . _

lI i

.e.-oom om.- -

_-4.. . , e ....,......,w, - . . ..-.%-.--- ==._-w..- = = = * = = - * * " ' " ' * ' * * " * * * ' " " * '

$01

ll l . ..

/n unreskic+ec_ area exists a pund kl __ j l 6{ Une .hi_lcljne _exlerioy . 1} ye, calculek S yound:_et/k redese. In hym3 of fhe dachoi)

..d % annd unresluhsd area eH%+ 'limW,_

' vit shu) k yeVV CDnserv&fi vs.

I (DUnenrveulea_.Ior bf weKe!

AssuA.]he insbanbanecas caewershion in he_eXluent is idenhcol k #1e certedvanm in -fne reec4cr oas., The fofal eunsure-U E3 _Ls_cBl.tul6aci 'ie6nkC&k as before , hf

_ .. _ win 4ht bmc inkrval 6 oc/nc &M6^4LT or 3,.15 x 1o

• s econcLs (Assenim v 7. 4ho% /dA

. . . . - - ,,~ #W '

.. E( =__Mo)O-e ~ h ~~

ACO) U & /n~3 s

- - . - . . - Jev 1V l

i

• ' I0 _h) 3 -

S'AONI0 15 g Ar-85m ge.y [

~

The pennissihIe eipwre G is w % ined* k se eH uen+ eencenirskon yC & O Li a ( a '~&

and 2n- annual ccnknuous eqx.sare- of

! b = -315_x 10* se_. _

l 1

..- .i i . .._

f" 89 94 'e y er

f W (r- 85,n, Eg 6 j Ep = (E C)(k ) = [5900e r ) 5.isn10s - 1.1 m o"

< )

pf rn 3) ,

jo For Xe - 35 , 6p l

=

~ {

. .. f2 5 9 0 . Q._.\f._$ 15 x.1.0..=s.._.

. . . . . _ . _ S... Ib 10 ..,

j

)

fn* ')%

~ ~

F <

_ 11 then m mafic 2 ate Ehe idosgfik i

% cwa ha as-  ;

a

.. .. . . . ..._...__._._.,__=__,Ld.,__.. . . _ _ . _ . _ . . . _ . _ --

._. b( . . .

bT . _ . _ _ -Kr4 5",_we.--. nave _

. . . F=  ; 2.14 x 10 (3 [ h = ,, g g j o-G( ). 1 T if10 " 6 m*3 5

/n; for Xe 'S5, we yr ,

I

'E 4 s ._ 2 9 x 10 ' 6g v3 3  ; 0,_02 7 l l

q v.

$f 9Ibx10*b(i's (

i

-=**i

,--.4 ., . e e ....i. ..m. ..--.....w.. .--.=.--*==e==- = = - * " " " " ' * ' " * * " * " " "

1 q- l

! @ ua wskkcanaiAuvs (ddeA lo caludux cw1 F 7 &..u M kLhmdL,_w_cw fabiu J.

___.. _y. nwMiq l

bYWd g 0Wmohosd ah.finsdd  ?&_)_

• Q gg , 1 l

~

., h --

I -

c

-bw _

_ l I l , sy, \ l

.._:._Ebw, > l Z 34 m') ( + m/s ) =

936 m% .

I s

1 1

,,.n..-

_ . . ....rcsfacFed area caske de ds (and &w.s L&  !

-% _eppw Q ) s%Id hJrehceA ako ate ;

Value s e 4W n W & rdio ok -Mc  ;

b _eK%F Iou)' rde o4 F[of

~

! dc Be__olda, wNe (J;lution)

. - befor F sw of 9  %...

l

~

Of -

-3

F

-p s. 2 Wc' _ 3,4s x f o py 9% m%

9 2JdKJg

% E (5.48K10~b h b 0 %)O s\ - 7.45x 10 ' & gh Kr-935 /T). 2 h

Egf+ 4sx109(2.2cx 70' g_) - 7. M x 70 '4.->;

w -- L : :mL

e -g-Thgs owvpagieu p q iga. wc;4 WL A F= Eg , - 7 . 4 s x t o ' q m -'s = L s w6' scis* Ep 1 17-x to Q,d's F = +.cc x ,o c g ,,, ss _

p.39x,o-s fe45 -

3 S.IQXto'* W 3 i

r

~ ~ ~ ~ ~ ~

~Z~

E -- _ - - - - _ _ _ _ _ _ _ - _ _ - - _ _ _ _ _