ML19341B206
| ML19341B206 | |
| Person / Time | |
|---|---|
| Site: | University of Virginia |
| Issue date: | 01/27/1981 |
| From: | Shriver B VIRGINIA, UNIV. OF, CHARLOTTESVILLE, VA |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8101300443 | |
| Download: ML19341B206 (6) | |
Text
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. UNIVERSITY OF VIRGINIA SCHOOL OF ENGINEERING AND APPLIED SCIENCE
- , v CH AR LOTT ESVILLE. 22909 DEPARTMENT OF NUCLEAR EN GIN EERING AND ENGIN EERING PHYSICS TELEPHON E: 304 924 7136 ICEACTOR F ACILITY January 27, 1981 Director Division of Reactor Licensing U. S. Nuclear Regulatory Comission Washington, D. C. 20545
Dear Sir:
This letter provides a writtu report of a reportable occurrence under license R-66 which covers the University of Virginia Reactor. (UVAR).
The reportable occurrence involved an uncontrolled or unanticipated change in reactivity which is defined as an abnormal occurrence in our technical specifications and is reportable.
The reportable occurrence occurred at approximately 0700 on Monday, January 19, 1981. A report of the occurrence was made to Mr. Dance, U. S.
NRC Region II at about noon on January 19, 1981. A written report was submitted to the University of Virginia Reactor Safety Committee on January 19, 1981. The Safety Committee met on January 23, 1981 to review the incident.
This report completes the actions defined by our technical specifications in the event of a reportable (abnormal) occurrence.
The reportable occurrence did not result in any safety limits or limiting safety system setting being exceeded. The reactor was not adversely affected and there was no hazard to the operating personnel or to the public.
REVIEW 0F UVAR SYSTEMS The University of Virginia Reactor (UVAR) is a two megawatt swimming pool reactor. It uses MTR fuel elements which are approximately 3 inches by 3 inches in cross section by about 34 inches long. They are held in the reactor grid plate by gravity and by the downward flow of the reactor coolant.
The reactor has three safety control rods which drop into the core if a reactor trip occurs. A fourth rod having a lower reactivity worth is used as a regulating rod to provide control of the reactor power when in steady state through an automatic control system. The control system is automatically snitched ,from automatic control to manual control if a limit between the actual p'ower and set point power is exceeded. The regulating rod does not drop into A the core during a reactor trip. [f0 M The reactor also uses a small (~one inch diameter) antimony beryllium source to provide neutrons during a reactor startup. The neutron source is fg normally located in a special holder adjacent to the core but can be moved by using a wire which connects the source to the reactor bridge located approxi-mately twenty one feet above the core. There are more'than nineteen feet of water between the top of the core and the reactor bridge.
810130C/N $.
Letter to Director Division of Reactor Licensing U.S. Nuclear Regulatory Commission January 27, 1981 Page 2 As noted below, small air bubbles are occasionally observed on the top of fuel elements. The bubbles are typically less than 0.5 inch in diameter and most often occur during initial startups when the pool is cold. The source of the bubbles appears to be the deaeration of the reactor pool due to the decreasing solubility of air as the pool temperature is increased.
The bubbles collect at the top of the fuel because much of the air is re-moved in the fuel elements when the water is heated. They remain above the fuel elements because they are balanced by the downward coolant flow against the bubbles buoyancy.
DISCUSSION OF THE INCIDENT During the routine core visual examination at 0700 on Monday, January 19, 1981, one of the two reactor operators on duty (operator 1) noted a small air bubble on the top of the fuel element in grid plate position 43.
The reactor was operating at a steady state power of 1.3 megawatts at that time.
Reactor operator 1 lifted the neutron source from its normal location (grid plate position 62) and lowered it above the fuel element to break up the air bubble. While removing the source, it became lodged under the fuel element handling bar and the fuel element was lifted from its normal location in the grid plate. According to a written statement provided by the operator, the fuel element may have been pulled as far as four inches above its normal position before she realized what was happening and lowered the source and fuel element back into the core.
The second reactor operator (operator 2) was at the reactor control console completing the hourly data sheet during the incident. During this time the alarm indicating the reactor power control had switched from automatic to manual sounded and very shortly thereafter a reactor tr'ip occurred due to an indicated short reactor period. The period trip is normally set at a period of 3.5 seconds.
The linear power recorder indicated that the reactor power was at 1.3 megawatts prior to the incident. The reactor power dropped to approximately.
0.24 megawatts, apparently as a result of the fuel element being removed from the core driving the reactor subcritical. This power deviation was much larger than necessary to switch the regulating rod from automatic cor. trol to manual control as observed by operator 2.
The linear power recorder then shows that the reactor power increased to approximately 0.65 megawatts where the reactor trip occurred. This apparently resulted from the lowering of the fuel element back to its normal location.
It is important to note that none of the safety limits or limiting safety system settings were exceeded. All equipment and safety systems performed as designed.
REVIEW 0F THE INCIDENT '
There were four aspects of the reportable occurrence which were reviewed
Letter to Director Division of Reactor Licensing U. S. Nuclear Regulatory Comission January 27, 1981 Page 3 by the reactor staff and by the Reactor Safety Comittee. These four aspects are discussed in the following sections:
- 1. Cause of the Incident It was concluded that the incident was primarily a result of a personnel error. A secondary factor was the use of an unwritten procedure which had not been fully reviewed.
Two separate areas of personnel error were identified. These are:
- a. The operator should not have taken action to clear the bubble from the core without the cognizance or supervision of a licensed senior operator. While the procesure had been used before, it had normally only beer performed by a senior operator. Thus we consider that the removal of bubbles was a non-routine event which should not be attempted without the approval of a senior operator. The fact that there was no written procedure for this evolution and that it involved handling an object near the fuel should have been sufficient reasons for the operator to notify a senior operator prior to attempting to break up the bubbles.
- b. The operator should have noticed the difference in weight between the neutron source and the neutron source plus the fuel element.
This difference in weight should have provided a quick indication that there was an unusual condition. At that point a careful visual check would have confirmed the source of the problem prior to moving the fuel element. At that point.the reactor should have been shutdown and the problem corrected by a senior operator.
Because both of these items indicate poor judgement on the part of the operator, her previous performance was evaluated to determine whether there were any previous indications of operating deficiencies which would indicate -
that she should be removed from the operators list. Both a review of the training records and discussions with the reactor staff indicated that the operator had performed well in the past and that there were no indications of unacceptable or questionable performance.
The second contributing factor to the incident was the use of an unreviewed procedure. This procedure had been used on occasion in the past, but as far as we know only by licensed senior operators. -The presence of r
bubbles was not considered significant to reactor safety by the reactor staff and therefore was judged to not require a written procedure as defined in the technical specifications. In-hindsight, it is clear that the procedure used-did have possible safety concerns and should not be used.
- 2. Reactivity Insertion During the Transient-Since the reactor was initially at steady state the effective multiplication l factor must have been one. When the fuel element was removed the reactor would
Letter to Director Division of Reactor Licensing U. S. Nuclear Regulatory Conmission January 27, 1981 Page 4 be made subcritical and the reactor power would drop. This is consistent with the observation.
If no other reactivity changes occurred, the r? actor should have returned to its original reactivity condition when the fuel element was reinserted.
However, there would be an initial positive reactor period due to the reactivity insertion and reactivity feedback (e.g. temperature coefficient, etc.). It would be expected that the reactor power would level off at its original power or lower. Thus, no large power transient would be expected.
However, the initial increase in reactor power would be expected and the initial period could be very high. For example, if only the prompt term of the Inhour equation is used a reactivity insertion of less than 2 x 10-5 AK/K would cause an initial period of three seconds. A reactivity insertion of approximately 4 x 10-3 AK/K would cause a stable period of three seconds which would trip the reactor.
One other source of reactivity which was considered was possible movement of the regulating rod. With the present UVAR core the regulating rod is worth less than 4 x 10-3 AK/ K. Thus this is an upper bound on the reactivity insertion which could have occurred if the regulating rod was fully inserted and was fully withdrawn during a transient. This did not occur.
The hourly data sheet taken immediately prior to the incident shows that the regulating rod was withdrawn 18.1 inch reactivity worth was approximately 4 x 10~gs. At that AK/ K. Thispoint its remaining information along with the fact that the regulating rod did trip from automatic to manual control shows that regulating rod movement did not contribute significant reactivity during the incident.
Thus, we believe that there is a very small probability that any of the reactor power limits would have been violated during the incident even if the reactor safety system would have failed.
- 3. Effect of the Incident on the Reactor As discussed above, none of the safety limits or limiting safety system settings were exceeded, thus, there is no basis for believing that the reactor was damaged by the incident. This was confirmed by visual examination of the fuel elements from the bridge and observation that none of the area radiation monitors showed an increase in the background level. It was also noted that there was no apparent change in reactor conditions when reactor operation was resumed.
- 4. Radiation Exposure During the Incident Since the reactor power remained below the normally allowed limits at all times during the incident and since the source and fuel element remained under over 18.6 feet of water (vs 19 feet minimum normally) there is no ' reason to believe that the incident contributed to radiation exposures to the operators
Letter to Director Division of Reactor Licensing U. S. Nuclear Regulatory Comission January 27, 1981 Page 5 or public during the incident. None of the area radiation monitors alarmed during the incident and no increase in radiation level was observed. The area monitors include one located on the bridge near operator 1.
CORRECTIVE ACTIONS The following corrective actions have ,3en or will be taken to ensure that this incident was adequately reviewe' and future, similar incidents avoided.
- 1. The procedure for removing air bubbles from fuel elements was revised on January 19, 1981. The new procedure involves the
, following steps.
- a. Notification of the senior reactor operator
- b. Shutdown the reactor
- c. turn off the reactor coolant pump
- d. observe that the bubbles have been removed 1
- e. resume reactor operation using normal startup procedures This procedure has been approved by the reactor safety comittee and will be included in UVAR S0P 11 " Procedures for Abnormal Conditions."
i This procedure may be revised in the future, but any revision will be reviewed by and approved by the Reactor Safety Comittee.
- 2. All rear. tor operators were informed of the incident and imediate corrective actions taken. This was accomplished by having operators on duty read a copy of the memorandum sent to the Reactor Safety Comittee and by discussions at the requalification meeting held on January 23, 1981. In addition to the details of the incident, the need for imediately informing a senior operator of any unusual conditions was discussed. The operators were also encouraged to' obtain approval from a senior operator prict to " attempting corrective actions" for unusual conditions unless they were considered necessary to protect the reactor or personnel. The major exception to this is shutting down the reactor, which all operators are authorized to do if their judgement indicates that it is the safe, conservative action to take.'
- 3. The reactor operator involved in the incident will be given an oral ar.d written reprimand for her actions. This will include a discussion of the specific concerns with her performance during the incident and methods for improving.
Letter to Director Division of Reactor Licensing V. S. Nuclear Regulatory Comission January 27, 1981 Page 6
- 4. The reactor safety comittee met on January 23, 1981 to review the incident and recommend corrective actions. The comittee agreed with the conclusions of this report and approved corrective actions identified above.
Sincerely, 8.
uh w
. Shriver, Director Reactor Facility BLS:ph cc: Mr. James P. O'Reilly, Director U. S. Nuclear Regulatory Commission-Region II 101 Marietta Street, N.W. , Suite 3100 Atlanta, GA 30303 Mr. James R. Miller, Chief-Standardization and Special Projects Branch Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 U. Va. Reactor Safety Committee i Mr. T. G. Williamson-U.Va.
Mr. J. R. Farrar-U.Va.
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