ML20070J522

Annual Operations Rept IA State Univ UTR-10 Reactor 930701- 940630

ML20070J522
Person / Time
Site:	University of Iowa
Issue date:	06/30/1994
From:	Adams J IOWA STATE UNIV., AMES, IA
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9407250157
Download: ML20070J522 (5)

Text

1 IOWA STATE UNIVERSITY ocparunenunteawucal Ensnenng 01- 5 C 1 L N C E A N D 1 E C 11 N O L 0 G Y Nutlear Enonernng l'rogram 107 N udcar Engmccring 1 ab Ames, Iowa yoi1-224 L 5 i 5 .,o4 ,84 0 l35 515 2Q4*7214 Docket No. 50-116 Ref: 10 CFR 50.71 (a)

July 15,1994 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington , DC 20555

Dear Sir:

Enclosed with this letter is the Annual Operations Report for the Io va State University UTR-10 reactor. The period covered by this report is from July 1,1993, to June 30,1994.

Sincerely,

, L b~ (huvS John T. Adams Reactor Manager

Enclosure c

American NuclearInsurers D. B. Bullen, Facility Director R. A. Jacobson, Chm., Radiation Safety Committee T. H. Okiishi, Chm., Mechanical Engineering Department E. E. Sobottka, Dir., Environmental Health and Safety Department T. L. Zimmerman, Chm., Reactor Use Committee U. S. NRC, Region Ill

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Annual Operations Report Iowa State University's UTR-10 Reactor Docket No 50-116 July 1,1993 to June 30,1994 This is a routine operations report to the Nuclear Regulatory Commission in accordance with the requirements of Section 6.6 of the Technical Specifications, Appendix A to Operating License R-59.

1. Summary of reactor operatine experience including the enerev nroduced by the reactor:

The reactor is operated in support of graduate teaching laboratories and graduate student research in the nuclear engineering program. One laboratory course was taught during this reporting period which provided hands on laboratory experience for students in the i graduate nuclear program. The reactor was also used to support the training of reactor operations personnel.

During the period July 1,1993, to June 30,1994, a total of 15.72 kw-brs of energy production and 78.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of operation were recorded. Last year's numbers were 154.31 kw-hrs and 173.13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Since the initial criticality of the LEU core in August of 1991, the cumulative kilowatt-hours are 170.09 kw-hrs and the cumulative hours of operation are 467.53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br />. The total energy produced during the life of the facility ( both HEU and LEU cores) is 7494.09 kw-hrs with a cumulative operation time of 9139.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. A percentage breakdown by operational categories for the years 92-93 and 93-94 is shown below.

Table 1. Allocation of energy production and operation time,in percent.

Year Research Teaching Teaching Maintenance Operator Service Undergrad. Graduate Training Energy %

93 94 0.0 0.0 4.8 91.8 3.4 0.0 92 93 92.7 <0 I <0.1 3.0 4.3 0.0 Time %

93-94 0.0 0.0 4.8 56.4 38.8 0.0 92-93 20.6 0.9 16.0 23.6 38.9 0.0 ,

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2. Unscheduled shutdowns includine, where applicable, corrective action taken to preclude recurrence:

There were no unscheduled shutdowns during this reporting period.

3. Maior oreventive and corrective mainterance operations havine safety sienificanee:

The process instmmentation test switch was replaced on July 29,1993, afler the failure of two of the five switch wafeis. The associated wiring was also replaced at the same time.

The process instrumentation associated with the test switch was tested following the switch installation.

The inner pillow block bearings were replaced on Safety Rod #1 (July 7,1993) and Safety Rod #2 (July 13,1993). Scram times on Safety Rod #1 and #2 were measured and found to be in compliance with the Technical Specifications. The replacement of these bearings was done in order to comply with a commitment made to the NRC to replace the inner pillow block bearings on all safety type rods after the Shim Safety Rod stuck 33%

withdrawn on November 13,1992.

A new 500 watt true sine wave static inverter was installed on November 11,1993, to provide an uninteruptable power source for the area radiation monitors. A battery capacity test was commenced, but failed prior to the completion of the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> test discharge.

Inspection of the battery revealed that cell number 3 had failed. Exide Battery was i contacted and it was decided that the battery should be replaced. On December 15,1993, a new battery was ordered. The new battery was received on January 5,1994. The battery was install and a freshening charge was performed on January 7,1994, in accordance with the Exide technical manual. The battery load test was completed on January 18,1994, and the emergency power system declared operable on January 20,1994, following the recharging of the battery.

The failure of the area radiation monitor emergency power system caused delays in the performance of 12-M-4, Core Flow Rate Calibration; 12-M-6, Thermal Power Measurement; 12-M-8, Multi-range Linear Power Channel Calibration; 12-M-12, Picoameter Calibration; 12-M-11, Nuclear Instrument Calibration; and 12-M-7 Rod Worth Measurements. During this time their were no reactor operations except those required for the completion of the previously listed surveillance procedures.

Throughout the reporting period, monitoring of the "in core" cladding samples has continued. Two of the three samples in the core have continued to darken. The remaining F sample has evenly darkened and does not appear to be changing. Alm, fission product l analysis using the high purity germanium detector (HPGe) has continued. The cause of the j darkening has still not been determined. Dr.Gerard Hofman, a senior metallurgist from I

Argonne National Lab (East), was provided a small piece of the cladding sample from the

core in July,1993, for analysis. Sample analysis results have not yet been received from Dr. Ilofman.

4. Maior channes in the reactor facility, procedures, and new tests or experiments, or both, that are significantly different from those performed previousiv and are not I described in the Safety Analysis Report,includine conclusions that no unreviewed safety auestions were involved:

The UTR-10 Emergency Plan was reviewed during August of 1993. The only major change made was the reassignment of responsibility for the transport ofinjured personnel.

Under the plan approved in 1991, this responsibility was assigned to Environmental Health

& Safety Department. The changes made in August,1993, now assigns this responsibility to the Mary Greeley Medical Center Ambulance Service.

In June of 1993, a license amendment was submitted to the NRC to add to the definition section of the UTR-10 Technical Specifications a definition for "Confmement Secured",

This request was made to clarify any ambiguity concerning the passage of personnel through the confinement boundary during fuel handling operations since T.S. 3.4.3.B requires the confinement boundary be secured during fuel handling operations. In November of 1993, notification from the NRC was received that the 10th amendment to the UTR-10 license had been approved adding the definition for " Confinement Secured" to the UTR-10 Technical Specifications. The addition of the definition required minor changes to the fuel handling procedure. Those changes were made, reviewed and approved by the ' Reactor Use Committee, and issued for use in March of 1994.

Following the failure and replacement of the emergency power supply for the area radiation monitors, the sun /cillance procedures were reviewed for applicability with the new equipment. Se,eral changes were required in the battery's quarterly maintenance procedure ( 3 M 2) due to the differences with t'ne new battery charging equipment. Also, a section was added to 3-M-2 to provide instructions for the performance of a equalizing battery charge. The procedure for the equalizing battery charge was obtained from the technical documentation provided by then Exide Corporation. The procedure for the triennial battery capacity test (36 M-1) was also changed to include the procedures for a equalizing battery charge.

In addition, a new monthly maintenance procedure (1-M-2) was developed to reflect the inspection and maintenance activities recommended by the Exide technical documentation.

This procedure provides direction to persoanel for the inspection of the battery and charging equipment, and the criteria to indicate the need for an equalizing battery charge.

Ali changes to the battery maintenance procedures were reviewed and approved by the Reactor Use Committee. The updated and new procedures were issued for use in February,1994.

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.- l A change was made to the thermal power measurement procedure (12-M-6) that reduced the indicated power at which the procedure is performed from 100% down to 95%. This action was taken to prevent exceeding the maximum licensed power level (10kW) in the ,

event that non-conservitive instrument drin has occurred since the last calibration of the nuclear instrumentation. The change was presented to the Reactor Use Committee and ,

was approved on March,1994.

5. Summary of the nature and amount of radioactive emuents released or discharned to the environs beyond the effective control of the University as determined at or before the noint of such r#3p or discharge. (included, to the extent ornetical, are estimates ofindividual radionuclides present in the emuent. If the estimate averane release after dilution or difTusion is less than 25 percent of the concentration allowed or recommended, a statement to this effect is usedh Argon-41: The operating records show that less than 25% of the concentration allowed was released to the environs.

Others: No measurable amounts of other radioactive effluents were released to the  ;

environs.

6. Summarized results of any environmental surveys performed outside the facilit n No environmental surveys outside the facility were required to be performed since the trigger level, based on surveys inside the facility, was not exceeded.

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7. Summary of exposure received by facility personnel and visitors where such l exposures are creater than 25 percent of that allowed or recommended:

No facility personnel or visitors had exposures greater than 25% of that allowed or recommended.