IR 05000482/1993009
| ML20036A566 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/04/1993 |
| From: | Howell A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20036A564 | List: |
| References | |
| 50-482-93-09, 50-482-93-9, NUDOCS 9305120096 | |
| Download: ML20036A566 (8) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-482/93-09 Operating License: NPF-42 Licensee: Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, Kansas 68839 Facility Name:
Wolf Creek Generating Station Inspection At:
Burlington, Kansas Inspection Conducted: April 12-16, with in-office review through April 20, 1993 Inspectors:
L. E. Ellershaw, Reactor Inspector, Maintenance Section, Division of Reactor Safety Dr. Dale A. Powers, Chief, Maintenance Section, Division of Reactor Sa y
Approved:
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6 -Y-N Arthu'r T. H ell,~ )eputy DTitctor, Division of Date Reactor fety Insoection Summary Areas Insoected: Routine, announced inspection of followup of licensee event reports, followup of corrective actions for violations, and other followup.
Results:
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The licensee's corrective actions taken in-response to Cycle 5 fuel fretting failures was broad scoped and comprehensive (Section'2).
The licensee's corrective actions taken in response to Cycle 6 fuel
fretting failures was narrowly focussed toward F fuel assembly performance, and did not validate (by appropriate ~ examination) the adequacy of suspect H fuel. assemblies for use in Cycle 7 (Section 2).
Summary of Insoection Findinos:
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Licensee Event Report 482/92-019 was closed (Section 2).
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f Licensee Event Report 482/93-004 was reviewed but not closed
(Section 2).
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Inspection Followup Item 482/9309-01 was opened (Section 2).
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Inspection Followup Item 482/9211-01 was closed (Section 3).
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i Violations 482/9230-II.A.1, 482/9230-II.A.2, and 482/9212-01 were closed
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(Section 4).
Attachment:
Attachment - Persons Contacted and Exit Meeting
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DETAILS 1 PLANT STATUS
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During this inspection period, the plant was in a refueling outage.
2 ONSITE REVIEW OF LICENSEE EVENT REPORTS (92700)
(Closed) Licensee Event Report 482/92-019 and (Ocen) Licensee Event Report 482/93-004 The licensee reported fuel cladding degradation by letter dated November 5, 1991. During Cycle 5 operation, the licensee's primary coolant radiochemistry
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data had indicated that fuel failure had occurred. Consequently, the licensee had entered into its applicable Administrative Procedure ADM-01-221, " Failed Fuel Action Plan," and maintained compliance with the Technical Specification limit for dose equivalent Iodine-131. During the core offloading operations, operators observed a broken fuel rod leaning out from a fuel assembly.
Fuel handling operations were then suspended until the fuel rod fragment was retrieved.
Subsequent ultrasonic and visual inspections of the Cycle 5 fuel assemblies in the spent fuel pool identified a total of three fuel assemblies that contained failed fuel. These assemblies were campaign F assemblies and were in their second cycle of operation when failures occurred. Their burnups at the end of Cycle 5 ranged between 38,100 and 39,000 MWD /MTU.
It was notable that the failed fuel rods were non-statistically distributed within the three.
assemblies.
Specifically, the failed fuel rods were predominately clustered on peripheral rows in the fifth-to-seventh grid cell locations (as counted from the assembly corners).
Failed fuel rods that were visible (on peripheral
rows) exhibited hydriding in the higher elevations of the core with steam cleaning plumes emanating from hydride blisters and other perforation. sites.
In addition, many of the total of 44 failed fuel rods had dropped and were in contact with the upper surface of the fuel assembly bottom nozzles. During examinations in the spent fuel pool, these dropped fuel rods were raised off the bottom nozzles to expose the cladding-to-grid contact areas.
It was then observed that the reason for the fuel failures was fretting that had occurred at the lower three grid locations where the cladding was in contact with grid spring clips and dimples.
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The licensee investigated potential root causes for the fretting failure
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mechanism. Among the potential causes investigated was the possibility of excessive flow through the subject fuel assemblies due to flow blockage under the. lower core plate. A search under the core plate, however, did not identify any debris capable of such significant flow diversion.
In addition,
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there was no direct evidence of significant cross-flow, baffle jetting, or other flow anomalies. The licensee explored the possibility of inadequate grid spring clip force by conducting fuel rod pull tests. These pull tests
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were inconclusive due to large variability in the data.
Spacer grid sizing tests on some individual spacer grid cells indicated that the cell dimensions were within the bounds expected. Other root causes that the licensee discounted were the possibility of adverse primary coolant system chemical conditions, inauspicious operating power history, and fuel handling damage.
The licensee's investigation also included the assistance of a consultant for a team audit of the fuel manufacturer's facility. This effort focussed on reviews of the manufacturing and fabrication records. From this effort, the licensee identified a plausibly significant commonalty:
the three F fuel assemblies were all assembled on a particular loading device, Fixture 1.
There was no compelling evidence that Fixture 1 was misaligned or that fuel
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assembly on Fixture I had been conducted improperly. However, the licensee, having reasonably exhausted potential theories for the root cause, redesigned the Cycle 6 core to preclude further fuel failures by excluding all F fuel assemblies that were assembled on Fixture 1.
Other F fuel assemblies that were assembled on another of the manufacturer's fixtures were reloaded into the Cycle 6 core.
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During the ensuing Cycle 6 operation, the licensee determined that fuel failures recurred.
Licensee Event Report 482/93-004, dated April 15, 1993, documented this experience, which was a virtual repeat of the prior occurrence in Cycle 5 operation including the subsequent core offloading experience with a broken fuel rod. As a result of the Cycle 6 F fuel assembly failure experience, the root cause~of the Cycle 5 fuel failures is now known not to be related to the use of Fixture 1.
The new licensee event report hypothesized that high cross and axial flow, resulting from the use of mixed fuel designs, reached a critical state of fluid elastic instability with concomitant fuel rod instability. This hypothesis was the result of the fuel manufacturer's computer' sensitivity study and is now the suspected root cause for the excessive fretting. The inspectors agreed that this hypothesis may'be correct. However, the inspectors also learned that the licensee had observed during the outage that 26 H fuel assemblies each had 1 to 10 dropped fuel rods that were resting on bottom nozzles. Although fuel rods in contact with the bottom nozzles is not
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in itself a problem, this observance brings into question the adequacy of the designed grid spring clip force in the F and H fuel assemblies. Moreover, the licensee's representative explained that the H fuel assemblies utilize a i
weaker top grid in order to allow the fuel rods to grow more easily in the axial direction as a function of burnup. The licensee's representative stated that.the fuel vendor had pcedicted that fuel rods in H fuel assemblies might contact bottom nozzles at sometime in their life. The inspectors agreed that the use of a weaker top grid was beneficial for enabling axial growth (and thereby reducing the extent of rod bowing), it is, nevertheless,.
counterproductive in that decreased grid spring clip force results in a decreased fuel rod mechanical stability.
(There are no Technical Specification thermal-hydraulic penalties for fuel rod bowing occurrences at
the Wolf Creek Generating Station. Consequently, the design change to a weaker top grid would have not improved an operational limitation.)
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Of concern to the inspectors was that the licensee did not perform any
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examinations of the dropped fuel rods in the H fuel assemblies in order to determine whether excessive fretting had taken place in Cycle 6, their first cycle of operation. As stated above, such examinations (i.e., fuel rod lifts)
were performed during the prior outage on F fuel assemblies and would have required little additional handling of the H fuel assemblies during the latest outage. During a meeting with personnel from the Office of Nuclear Reactor Regulation on April 20, 1993, the licensee's fuel vendor representative stated that he would have strongly disagreed with the licensee performing fuel rod lift examinations on fuel to be reinserted into the core. The representative, however, responded to an inspector's question that there was no operational experience that validated his concern. The inspector noted that the risks associated with such an examination were inconsequential in comparison with the risks from fuel handling that would result from possible fuel failures in Cycle 7.
Nevertheless, an appropriate examination would have been prudent corrective action in order to validate the adequacy of the H fuel assemblies for continued use.
The licensee's representative stated that it was expected that the Cycle 7 core will experience less cross flow with reduced propensity for fuel rod fretting as a result of the similarity of fuel assembly bottom nozzles used in the Cycle 7 core. The licensee's assertion was reasonable, but not quantified inasmuch as the analytical determination of the magnitude of the difference in cross flow has inherently large uncertainties. The inspectors also noted that the fuel batches to be used in the Cycle 7 core will not be exactly alike and -
some continuing cross flow will remain. The fuel design differences for Cycle 7 include differences in fuel rod diameter and grid designs.
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Licensee Event Report 482/91-019 is closed. The licensee's corrective actions taken in response to the Cycle 5 fuel failures was broad scoped and comprehensive.
However, NRC will conduct further followup on Licensee Event Report 482/93-004
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to determine the adequacy of the licensee's corrective actions.
The NRC review of the adequacy of the licensee's corrective actions to preclude fuel
failures in Cycle 7 is considered an inspection followup item (482/9309-01).
3 FOLLOWP (92701)
i (Closed) Insoection Followuo Item 482/9211-01: Review the New Technical Staff and Manaaer Trainina Proorg
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At the time this inspection followup item was identified, the licensee was in
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the process of developing a new technical staff training program to amplify the training specified in Section 13.2.2.8 of the Updated Safety Analysis l
Report. The licensee had not yet identified the functions, positions, or
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individuals that would be subject to training provided by this program.
The new technical staff training program was issued as Revision 4 to i
Procedure KGP-1851, " Training and Qualification Program for Engineering
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Support Personnel and Supervisory Personnel," dated March 11, 1993. The
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procedure defined engineering support personnel as Wolf Creek Nuclear i
Operating Corporation (WCNOC) personnel and contractor personnel working under the WCNOC quality assurance program and performing activities that were
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important to the safe and reliable operation of Wolf Creek Generating Station.
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The procedure provided detailed discussion regarding qualifications, orientation training, position-specific training, continuing training, and engineering support personnel supervisory training.
The procedure also referred to Revision 0 to Manual WCNOC-13A, " Engineering Support Personnel
Qualification Manual," dated March 12, 1993. The manual described the
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training and qualification requirements for engineering support personnel in both narrative and table format. The use of documentation necessary to support the accomplishment of the above activities had been clearly established.
The licensee appeared to have developed a well planned training program for the technical staff.
4 FOLLOWUP ON CORRECTIVE ACTIONS FOR VIOLATI WS (92702)
4.1 (Closed) Violation 482/9230-II.A.1:
Failure to Incorporate a Procedure Chance A procedure change form was not incorporated into Revision 17 to Procedure CKL EF-120, " Essential Service Water Valve, Breaker, and Switch Lineup," thus, the procedure did not delineate the properly locked throttle position for Valve EF V058.
Upon identification of this problem, the licensee initiated Procedure Change Form M192-725 to reflect the correctly locked valve position. The change was incorporated into Revision 18 to Procedure CKL EF-120 on October 5, 1992. The licensee determined that the administrative procedures which controlled this activity did not provide adequate controls to ensure that procedure change forms would be incorporated under all conditions. The licensee revised Administrative Procedures ADM 02-011. " Shift Clerk Qualifications and Responsibility," ADM 02-106, " Procedure Update and 2-Year Review," and ADM 07-100, " Preparation, Review, Approval, and Distribution of Wolf Creek Generating Station Procedures," to provide clear guidance and responsibilities, and to establish additional controls to ensure that all procedure change forms would be incorporated into the appropriate procedures.
After review of the above changes, the inspectors determined the licensee's.
evaluation and disposition of this issue were acceptable.
4.2 (Closed) Violation 482/9230-II.A.2:
Failure to Provide Adecuate Procedural Guidance Revision 0 to Temporary Procedure TP-TS-II5, "A-Train Essential Service Water Flow Verification to Component Cooling Water Heat Exchanger," provided an expected flow value and acceptable range of flow.
However, there was no
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-7-explicit guidance to indicate that the system would be inoperable if the measured flow was greater than the acceptance criterion. On August 28, 1992, the test engineer informed the shift supervisor that the completed flow verification test had been satisfactority completed, even though the measured flow exceeded the acceptance criterion. As a result, the licensee prematurely exited the Action Statement for Technical Specification 3.7.3, which is applicable to the component cooling water system.
During subsequent review of test results on August 28, 1992, the supervising engineer discovered that the flowrate exceeded the acceptance criterion of Temporary Procedure TP-TS-115, and the engineer notified the control room staff. The component cooling water A-Train was declared inoperable. After establishing the correct flowrate through the heat exchanger by repositioning the component cooling water heat exchanger bypass valve, the A-Train was declared operable.
Procedure ENG 09-506, "Results Engineering Pre-Job Check List," was revised (Revision 3) on October 22, 1992, to require Results Engineering personnel to review a check list of items during the pre-job briefing for any surveillance of technical or non-technical specifications, work requests, or temporary procedures.
Emphasis was placed on verification that acceptance criteria were clearly specified. The procedure further required performance of these activities any time the surveillances, work requests, or temporary procedures underwent an unanticipated change in plant status, significant delay in work, or as directed by the supervisor of Results Engineering.
The actions taken by the licensee should preclude a recurrence of this problem.
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4.3 (Closed) Violation 482/9212-01:
Inadvertent Dilution of the Soray Additive Tank During the performance of a surveillance test, an inadvertent dilution of the spray additive tank occurred when a misalignment of valves took place.
This occurrence was a result of personnel failing to assure that-surveillance i
procedure steps were performed in their proper sequence.. Subsequent to restoration of the spray additive tank, the licensee counselled the personnel involved.
Performance Improvement Request OP92-0430 dated June 3, 1992, was initiated to evaluate this problem and was included in " Operations Required
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Reading" for licensed operators and nuclear station operators. Wolf Creek
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Generating Station Standing Order No. 7, " Discussion of General Operating Philosophy Regarding Plant Evolutions," was revised (Revision 2) on July 6, 1992, to clarify the control of evolutions in progress.
In addition, Lesson Plans LR 10108 50, " Licensed Operator Requalification Training," and
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NO 16 360 09, " Nuclear Station Operator," were revised on November 17, 1992, and October 29, 1992, respectively, to include the circumstances surrounding this event. The inspectors verified that the requalification training presented subsequent to this event had included this information.
The inspectors considered the actions taken by the licensee to be appropriat,. -
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E ATTACHMENT
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1 PERSONS CONTACTED
.l.1 Licensee Personnel P> Adam, Supervisor, Reactor Engineering S. Ferguson, Supervisor, Fuel Engineering
- J. Lutz, Engineer, Regulatory Compliance
- 0. Maynard, Vice President, Plant Operations R. Meister, Senior Engineering Specialist, Regulatory Compliance G. Rathbun, Manager, Nuclear Analysis
- T. Reilly, Supervisor, Regulatory Compliance
- F. Rhodes, Vice President, Engineering R. Schneider, Shift Supervisor J. Tarr, Engineer, Regulatory Compliance 1.2 NRC Personnel
- G. Pick, Resident Inspector In additbn to the personnel listed above, the inspectors contacted other licensee oployees during.this inspection period.
- Denotes personnel attending the exit meeting.
2 EXIT MEETING An exit meeting was conducted on April 16, 1993. During tnis meeting, the inspectors reviewed the scope and findings of the report. The results of the--
inspection were also discussed during an April 20, 1993, meeting with licensee and Office of Nuclear Reactor Regulation personnel. The licensee did not identify as proprietary, any information provided to, or reviewed by the inspectors.
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