IR 05000186/1981003
| ML20010C313 | |
| Person / Time | |
|---|---|
| Site: | University of Missouri-Columbia |
| Issue date: | 07/28/1981 |
| From: | Boyd D, Ridgway K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20010C300 | List: |
| References | |
| 50-186-81-03, NUDOCS 8108190342 | |
| Download: ML20010C313 (11) | |
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~U.S. NUCLEAR REGULATORY COMMISSION
. OFFICE OF INSPECTION AND ENFORCEMENT nEGION III Report No'. 81-03 Docket No. 50-186 License No. R-103
. Licensee: University of Missouri Research Park Columbia, MO 65201 Facility Name: Research Reactor Inspection Conducte1: -May 18-22, ?981 N
Inspector:
K.
. Ri ay l'
44 Approved By:
D. C. Boyd, Chief-
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Reactor Projects Section 1A Inspection Summary
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Inspection on May 18-22, 1981 (Report No. 50-186/81-03)
Areas Inspected: -Routine unannounced inspection of records, logs and organi-zation, review and audit functions; procedures;. surveillance and maintenance; refueling; fuel shipping; experiments; radiation protection program; radwaste management program; transportation of radioactive materials; and followup action relative to IE Circulars, Open Inspection Items and Licensee Event Reports. This inspection involved a total of 30 inspector-hours onsite by one NRC inspector including 0 inspector-hours oasite during off-shifts.
Results: Of the ten areas inspected, no items of noncompliance were iden-tified in eight areas, three apparent items of noncompliance were identified in two areas. (Failure to provide high pressure safety circuit scram capability, failure to maintain contsinment leak rate within limits, failure to test containment truck-entry rod run-in on the require frequency.)
0100190342 010004 PDR-ADOCK 05000186, O
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DETAILS
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1.
Persons Contacted
- R. Brugger, Director, Research Reactor Facility
- P. Collins, Associate Vice-President for Academic Affairs
- J. - Tolan, University Radiation Safety Officer
- 0. Olson, Manager, Reactor Health Physics
- W. Meyer, Chairman, Reactor Jafety Subcommittee
- J. McKibben, Reactor Manager
- P. Keenan, Director, Research Program Service C. Edwards, Reactor Plant Engineer M. Vonk, Reactor Engineer T. Seeger, Chief Electronics Technician J. Litton, HP Technician S. Stewart, HP. Technician
- S. Growcock, HP Technician
- D. McGinty, Reactor Physicist The inspector also interviewed other licensee employees and facility researchers.
- Indicates ~those attending the exit interview.
2.
Organization, Logs and Records The facility organization was reviewed and verified to be consistent with the Technical Specifications and/or Hazards Summary Report.
The minimum staffing requirements were verified to be present during reactor operation', and fuel handling or refueling operations.
The reactor logs and records were reviewed to verify that:
a.
Required entries were made.
b.
Significant problems or incidents were documented.
The facility was being maintained properly.
c.
d.
Records were available for inspection.
Since the last operational safety inspection,1/ the Reactor Operations Engineer, two Senior Reactor Operators and one Reactor Operator have terminated. The licensee is in the process of recruiting replacements.
Two reactor operator trainees are scheduled to take license tests in June, 1981.
No items of noncompliance or deviations were identified.
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IE Inspection Report No. 50-186/80-05-2-
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3.
Reviews and Audits The licensee's review and audit program records were examined by the inspector to verify that:
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Reviews of facility changes, operating and maintenance procedures, design changes, and unreviewed experiments had been conducted by a safety review committee as required by Technical Specifications or Hazards Summary Report.
b.
That the review committee and/or subcommittees were composed of qualified members and that-quorum requirements and frequency of meetings had been met.
c.
Required safety audits had been conducted in accordance with Technical Specification requirements and that any identified problems were resolved.
Since the last operational safety inspection,2/ the Reactor Safety Subcommittee has established a minimum meeting frequency of every six months. The Reactor Advisory Committee (RAC) at a meeting March 26, 1981 established a formal audit program whereby operat tonal areas will be audited. The first two audits are scheduled for June, 1981.
No items of noncompliance were identified.
4.
Procedures The inspector reviewed the licensee's procedures to determine if procedures were issued, reviewed, changed or updated, and approved in accordance with Technical Specifications and HSR requirements.
This review also verified:
That procedure content was adequatt to safely operate, refuel a.
<rd maintain the facility.
b.
That responsibilities were clearly defined.
c.
That required checklists and forms were used.
The inspector determined that the required procedures were available and the contents of the procedures were adequate.
No items of noncompliance were identified.
5.
Surveillance The inspector reviewed procedures, surveillance test schedules and test records and discussed the surveillance program with responsible personnel to verify:
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a.
That when necessary, procedures were available and adequate to perform.the tests.
b.
That tests were completed within the required time schedule.
.c.
-Test records were available.
During the review of surveillance test records, the inspector found that Compliance Test, CP-15 Truck Door Rod Run In, a semiannual test required by Technical specification 3.4.c and 5.4.a had not been conducted since July 20, 1980. This is considered to be a noncompliance item. The test had been scheduled during a mainten-ance outage in February 1981, but had been overlooked. The licensee scheduled and performed the test during the next maintenance outage, June 1, 1981, and has instituted a double check of surveillance records to assure that required tests will be conducted en schedule.
No other items of noncompliance were identified.
6.
Experirt;nts The inspector verified by reviewing experiment records and other reactor logs that:
Experiments were conducted using approved procedures and under a.
approved reactor conditions.
b.
New experiments or changes in experiments were properly reviewed and approved, The experiments did not involve an unreviewed safety question, c.
i.e., 10 CFR 50.59.
d.
Experiments involving potential hazards or reactivity change were identified in procedures.
Reactivity limits were not or could not have been exceeded e.
during the experiment.
No items of noncompliance were identified.
7.
Refueling The facility refueling (fuel handling) program was reviewed by the inspector. The review included the verification of approved procedures for fuel handling and the tachnical adequacy of them in the areas of radiation protection, criticality safety, Technical Specification and security plan requirements. The inspector determined by records review and discussions with personnel that fuel handling operations and startup tests were carried out in conformance to the licensee's procedures.
No items of noncompliance were identified.
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8.
Fuel Shipping The inspector reviewed records of.the irradiated fuel shipments made since the last inspection to determine that conditions of the Certi-ficate cf Compliance for the shipping cask and DOT regulations were followed.
Three shipments had been made without incident in November and December, 1980. The inspector reviewed records of these shipments to determine that the licensee's approved Quality Assurance Program
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for shipping containers and the handling of them was being followed during the cask receipt, loading and shipping.
No items of noncompliance were identified.
9.
Licensee Event Reports Followup Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifi-cations.
The licensee reported on October 20, 1980, that on September 22, a.
1980, while performing a routine semiannual compliance check of the reactor coolant inlet high temperature scram unit 980A, the meter relay control unit scram did not actuate in response to a high temperature si -
1.
The failure was attributed to the de-t crease in effectivene rs of a capacitor in the meter control cir-cuit. There had been no previous problem with this thannel since its installation in July, 1973. The meter relay control unit was replaced with a new unit and the compliance check was satis-factorily performed. The capacitors in six similar control units
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were scheduled to be tested when their respective semiannual compliance tests are performed.
b.
The licensee reported on February 5, 1981, that on January 9-10, 1981, the exhaust stack radiation monitor sampler blower was inadvertently left shutdown following the calibration of the iodine instrument for about 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> of startup and operations at full power (10MW)..The particulate, gas and iodine indication and alarm instrumentation were operating but low air flow in the sampler (7.5 l'.ters/ min, compared to 235 liters / min. with the blower in operation) would have made the particulate and iodine sampler results unreliable and probably would have degraded the systems chility to promptly respond to particulate and iodine activity. However, the gaseous monitor was not affected and the alarm function of this monitor would have responded before maximum permissible concentrations were reached in the stack gases.
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The stack monitor has no automatic functions but gives indication and annunciation of high stack activity in the control room.
During this event both automatic containment isolation systems were still operable to prevent release of any activity to the environment.
The stack monitor charts were reviewed for any positive indica-tions and the filters were counted. A charcoal filter from an experimental iodine monitor was also counted. All analyses indicated the stack gas activities to be normal and within Technical Specification limits during the time the blower was not operating. Although the stack sampler system had a high/ low flow alarm it's power came from the same circuit as the blower and when the blower was shut down the alarm was also inoperative.
The alarm is now powered from an independent power supply and a blower run light has been installed in the control room. The pre-startup check list has been revised to include an operability check of the stack monitor.
c.
The licensee reported on March 13, 1981, that on February 13, 1981, during a reactor startup Nuclear Instrument (NI) Channel failed to respond positively to changes in reactor subcritical multiplication and after checking the compensation voltage for Channel 3, the reactor was promptly shut down by a manual scram.
NI Channel 3 and also Channel 2 provide intermediate power range and reactor period indication. The reactor had beca shut down earlier in the day when Channel 3 failed causing a reactor scram.
The failure was determined te be caused by a leak in the detector dry well which shorted the detector. The leak was repaired and new detector cab?es installed. During the trouble shooting to determine the Channel 3 failure cause, three cables from the Channel 3 detector were interchanged with cable s from Channel 2.
Following the check out and when the cables were reinstalled, the positive high voltage cable and the compensation negative high voltage cables for Channel 3 were instc11ed reversed. The required front panel checks and bugging of tcf instrument were carried out prior to startup and indicated tF: channel was functioning properly. To prevent reversing.he high voltage cables in the future, the detector cable connectors were made dissimilar, d.
The licensre. reported on March 17, 1981, that on February 23, 1981, one of two reactor convective cooling loop isolation valves (V546B) failed to fully open. The valve was rebuilt and tested satisfactorily prior to returning the reactor to operation. Valves V546A/B will be operationally checked routinely during preventive maintenance.
The licensee reported on March 20, 1981, that on February 23, e.
1981, while the reactor was shutdown for a regular maintenance / day a 3/4 inch isolation valve in the containment building leak rate determination system was found open. The open valve was discovered by a Health Physics Technician when he heard the abnormal noise of air in-leakage while making his rounds in the basement cutside the-6-
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containment bnilding. The 3/4-inch valve is in the system used to pressurize the containment building for required leak testing.
Technical Specifications 3.5.a requires " Containment integrity shall be maintained at all times except when:
(1) the reactor is secured, and-(2)
irradiated i,21 with a decay time of less than sixty days is not being handled."
Technical Specification 1.15 defines Reactor Containment Integrity as: "For containment integrity to exist, the following conditions
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must be satisfied:
(1) The truck entry door closed and sealed.
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(2) The utility seal trench filled eith water to the depth required to maintain a minimum water seal of 4.25 feet.
(3) All containment building ventilation system automatically-closing doors and automatically-closing valves are operable or placed in the closed position.
(4) The reactor mechanical equipcent room exhaust system, including the particulate anJ halogen filters is operable."
(5) The personnel airlock door operable.
(6) The most recent building leak test was satisfactory.
Therefore, containment integrity, as defined, was not violated.
However, Technical Specification 4.2 c requires, "The containment
leakage rate to be less than 16.3 f t / min. at an overpressure of one pound per square inch gauge (psig) or 10% of contained volume over a 24-hour period from an initial pressure of two psig."
On May 4, 1978 and June 26, 1978, the licensee applied for and on July 3, 1978 received License Ammendment No. 10 whic permitted leak rate testing at a constant one psig over pressure.gj In their submital was an evaluation to determine the effective orifice size that represented a leakage of ten percent of contained volume over a 24-hour period from an initial overpressure of 2 psig.
It then determined that an equivalent orifice size of 0.106 square inches would give the same leakage rate at a constant one psig overpressure or 16.3 cubic feet per minute.
The cross section of a 3/4-inch schedule 40 pipe is 0.53 square inches. Therefore, the Technical Specification 4.2.c containment leakage rate would not have been met during the period of time the valve was open.
3f Letters from NURR to NRR dated 5/4/78 and 6/26/78-7-I
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Inthesameappliegion/andinNRR'sSafetyEvaluationReport
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for this amendment and the Hazards Summary Report for the MLER6/7/
it has been dett.imined tt'at the design basis accident or any other credible accident will not increase the pressure inside containment and-that the-2 psig pressure overpressurization requirement was a building design and construction specification.- Since there could be no pressurization of containment, leakage from the 3/4-inch valve would be-only by. pressure differences between the containment building and the laboratory basement area at atmospheric pressure.
In addition, any escaping activity would be confined to the laboratory building which surrounds the containment building. The presence of activity in the laboratory basement would be detected -
by hot cell-radiation monitors or by manual monitoring-and the open line could be closed as it is outside of containment. Therefore, the rpea 3/4-inch valve is not considered to be a significant degregation of the reactor safety.
The containment had last passed the leak rate test on April 12,.
1980 and early in January, 1981, a valve lineup was conducted as a training exercise and found to be correct. The licensee was unable to determine why or when the valve was opened. A white "Do Not Operate" tag was attached to the valve throughout the interval.
The licensee immediately conducted a valve lineup check of the containment leak rate test system. No other valves were found out of position. A pipe cap has been installed on the open end of the test line inside containment and leakrate test procedures have been revised to include installing this cap at the conclusion of the leak rate test.
"Do Not Operate" tags will continue to be placed on all test valves. The licensee also reviewed all other containment penetrations and found no other similar problem on the other systems.
l-f.
The licensee reported on May 18, 1981, that on April 20, 1981, while performing a routine semiannual compliance test on the reactor low pressure scram unit 943, one of four low pressure reactor scram units, it failed to operate.
The failure was attributed to a degraded espacitor. There had l
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been no other failures of this channel since its installation in 1974, however,. a similar scram unit had failed because of a degraded capacitor (see Paragraph 9a) on September 22, 1980.
Corrective action to be taken following this failure was to test the capacitor in the other six similar meter relay control units during their semiannual compliance. test. This was the first compliance test for unit 943, however, the' capacitor in unit 917 was not tested during'its semiannual test in December, 1980.
Letter from MURR to NRR dated 6/26/78 j
.gf Safety Evaluation Supporting License Amendement 10, 7/13/78-77 Hazards Summary Report, Addendum 3, p. 87, August 1972 Hazards Summary Report, Addendum 4, PC-3 & 4, October 1973
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The licensee has replaced all seven meter relay control units with bench tested spares. Capacitors in the removed units were tested and found to be within design tolerances. Each compliance procedure was revised to include a check to verify that electronic preventive maintenance items have been completed. The licensee also plans to remove all seven units after a year's operation and measur e the capacitance values.
If no abnormalities occur, the capitors will be tested every five' years.
g.
The licensee reported on May 18 and May 26, 1981, that on May 18, 1981, while inspecting primary mechanical equipment in preparation for a reactor startup following a scram from loss of electric power, the pressurizer high pressure scram switch, 939, was found to be isolated. The valve (V599B) had been closed on April 17, 1981, for a compliance check wLich required the attachment and removal of a test Heise gauge. The valve was evidently left in the closed position following the Heise gauge removal. This is in noncompliance with Technical Specification 3.3.a. which requires an operable Pressurizer High Pressure scram whenever the reactor is operated.
The reactor is normally operated at a pressure of 70 psig with inlet 'and outlet primary coolant temperatures of 140* F and 158* F respectively.
Pressure is controlled in a 300 gallon pressurizer (PZR) by bleeding in nitrogen at 66.5 psig or venting the PZR at 73.5 psig.
PZR pressure is indicated in the control room and has a high pressure annunciator at 77 psig.
In addition, the primary system is protected by three pressure relief valves set at 100 psig. PZR water level is automatically controlled by adding water through valve 527B with a 55 gpm positive displace-ment pump or bleeding water out to the drain collection system via valve 527A.
The Technical Specification pressure limit is 110 psig, the design pressure of the primary system is 125 psig and it is hydrostatic tested at 150 psig.
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The Hazards Summary Report describes the pressure control system and provides a High Pressure Transient Analysis. This report analyzes three possible pressure transients caused by PZR mechanical or instrument control problems.
Previous HSR analyses had proved that under the most unfavorable conditi will not experience boiling of the primary coolantgys the reactor and the design basis accident which is not considered to be redible or any other credible accident would not cause a high pressure transient. The three possible means of high pressure transients would require multiple equipment or instrument failure to exceed the Technical Specification high pressure limit as well as no operator action upon receipt of the high pressure annunciator,
"ff Hazards Summary Report Addendum 5, January 1974 Hazards Summary Repert Addendum 4, p. 8/, October 1973-9-
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The Hazards' Summary also ctates that the primary protection from overpressurization is the pressure relief valves, as an precaution,thehighpressurescramwasalsoinstalled.ggpditional
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During the time the high pressure scram switch was valved out and would not have performed its designed function, all other pressure control and high pressure protections were operable, i.e.,
PZR pressure indication and high pressure annunciation and all three pressure relief valves. Therefore, it is not probable that the Technical Specification high pressure limit of 110 psig, which can be considered a limiting condition of operation, would ever be exceeded.
For corrective action the valve 599B was opened prior to startup.
The operators who had made the compliance check in April were made aware of the importance of returning tested systems to normal following maintenance work or tests. The startup check list was revised to include checking that both the pressurizer high pressure and low pressure scram switch isolation valves are open prior to startup.
In addition a review of all safety system sensors was made to assure that they could not be isolated without indication in the control room. All compliance test procedures have been reviewed and revised to provide checks that valving on systems tested are placed in normal condition following tests.
10.
IE Circular Followup For the IE Circulars listed below, the inspector verified that the Circular was received by the licensee management, that a review for applicability was performed, and that if the circular was applicable to the facility, appropriate corrective actions were taken or were scheduled to be taken.
IEC 81-02 - Performance of NRC Licensed Individuals While on Duty.
11.
Review of Periodic and Special Reports The inspector reviewed the following reports for timeliness of submittal and adequacy of information submitted:
Monthly Operating Reports, September 1980 through April 1981.
12.
Radiation Protection The inspector reviewed records and irterviewed personnel to determine that the surveillance requirements in the area of radiation protection had been carried out and no adverse conditions were noted.
No items of noncompliance or desiations were identified.
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Hazards Summary Report, Addendum 4, p. A-12, October 1973
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1.- Ridioactive Effluents
The ~ inspector reviewed records of gaseous, liquid and byproduct releases -
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'since the:last inspection.
P Nofitems of-noncompliance were ident'ified.
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14. Exit' Interview The inspector met with licensee represent'atives (dencted in' Paragraph 1)
at the conclusion of the inspection on May 22, 1981, and summarized the-
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scope and findings of the; inspection. The following ma:.ters were dis-cussed:
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The fact that already in 1981 the number of reportable events have exceeded those in 1980 (5) and that four of the events were caused
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b.
LThe noncompliance items in'the Appendix A to the inspection report letter.
c.
. Corrective actions taken by the licensee on open items from the previous inspection.concerning Reactor Advisory Connittee business.
These matters are considered to be closed (Paragraph 3).
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