ML20217A724
| ML20217A724 | |
| Person / Time | |
|---|---|
| Issue date: | 08/12/1997 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-3068, NUDOCS 9803250190 | |
| Download: ML20217A724 (7) | |
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CERTIFIED B,_X:
Mr. John Barton - 8/I2/97 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
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MINUTES OF THE JOINT MEETING OF THE ACRS SUBCOMMITTEES OM PLANT OPERATIONS AND ON FIRE PROTECTION JULY 18, 1997 ARLINGTON, TEXAS INTRODUCTION The Advisory Committee on Reactor Safeguards (ACRS) Subcommittees on Plant Operations and on Fire Protection held a joint meeting on July 18, 1997, at the Region IV office, 611 Ryan Plaza, Suite 400, Arlington, Texas.
The purpose of this meeting was to discuss negion IV activities and other items of mutual interest, including significant operating events and fire protection issues.
The entire meeting was open to the public.
Mr. Amarjit Singh was the cognizant ACRS staff engineer for this meeting.
The meeting was convened at 8:30 a.m.
and adjourned at 3:00 p.m.
ATTENDEES ACBS.
J. Barton, Chairman M.
Fontana, Member D.
Powers, Co-Chairman R.
Seale, Member D. Miller, Member J.
Larkins, ACRS Staff T. Kress, Member A.
Singh, ACRS Staff W.
Shack, Member J. Mitchell, EDO Staff Princinal NRC Recion IV Sceakers E. Merschoff, Regional Administrator J.
Dyer, Deputy Regional Administrator P.
Gwynn, Director, Division of Reactor Projects (DRP)
Ji. Howell, Director, Division of Reactor Safety (DRS)
D.
Chamberlain, Deputy Director of DRS J.
Shackelford, Senior Reactor Analyst (SRA)
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W.
Jones, SRA D.
Powers, Chief, Maintenance Branch, DRS h b>U /
C. Vandenburgh, Chief, Engineering Branch, DRS D.
Spitzberg, Acting Chief, Nuclear Materials Inspection and Decommissioning Branch, Division of Nuclear Materials Safety (DNMS)
No written comments or requests for time to make oral statements were received from members of the public.
A complete list of meeting attendees is kept in the ACRS Office File and will be made available upon request.
T1.e presentation slides and handouts used during the meeting are attached to the office copy of these minutes.
Chairman's Ooenina Remarka Mr. John J. Barton, Chairman of the Plant operations Subcommittee, convened the meeting at 8:30 a.m.
He stated that the purpose of 9003250190 970812 h]( j)} f)()
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Pl' ant Ops. & Fire Prot. July 18, 1997 Joint Subete. Minutes the meeting was to discuss Region IV activities and other items of mutual interest, including significant operating events and fire protection issues.
i Egalon IV Presentations Introduction Mr.
Ellis W.
Merschoff, Regional Administrator of Region IV, discussed his philosophy for directing the Region IV activities.
One 'of his primary goals was to establish good communications between Region IV and NRC Headquarters, and between Regional Inspectors and the licensees. He indicated that Region IV includes 11 contiguous States, plus Alaska, Hawaii, and Samoa.
He has emphasized the significance of having frequent -and vigorous communication between regional and Headquarters counterparts.
He holds weekly discussions with the resident inspectors and arranges quarterly meetings with the licensees to discuss topics of interest.
These communications are intended to obtain timely feedback on areas of concern before big problems arise.
Realon IV Orcanization Mr. James E.
Dyer, Deputy Regional Administrator of Region IV, addressed organization, allocation of resources, and supervisory and management development within the region and activities for maintaining uniformity among the regions.
He noted that Region IV has the responsibility for overseeing 21 operating reactors including Grand Gulf from Region II and Callaway from Regien III.
Responsibility was assigned to Region IV in October 1995, 20 test and research reactors.
Region IV was the first region to transfer all of the non-power reactor regulation activities back to NRC Headquarters.
In addition to the operating reactors, Region IV also oversees te uranium fuel fabrication facilities and 1,241 byproduct materials licensees.
In April 1994, Region V was merged with Region IV and Walnut Creek Field Office was created as a temporary satellite office.
Overall, Region IV is organized into four divisions: the Division of Reactor Projects (DRP), the Division of Reactor Safety (DRS),
the Division of Nuclear Material Safety (DNMS) and the Division of Resources Management and Administration (DRMA) Region IV has four principal responsibilities:
inspection, enforcement, licensing activities, ar,'d incident response.
The Walnut Creek Field Office has no incident response center; all the previous incident response activities for Region V sites are managed through the Arlington office, In regard to maintaining uniformity, Mr. Dyer spoke of counterpart meetings, joint inspections involving inspectors from
.other regions, rotational assignments in and out of the region, and frequent coordination and contact throughout the NRC.
He also
Plant Ops. & Fire Prot. July 18, 1997 Joint Subete. Minutes I
discussed various phases of the inspection programs; the training and development programs for new and qualified inspectors; J
organization, composition, and planning considerations; and current
' topics for team inspections.
He presented examples of recent l
management meeting tcpics, systematic assessment of licensee
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performance (SALP) board evaluation techniques, and inspection I
findings.
He also noted Region IV's interaction with Institute of the Nuclear Power Operations (INPO) and discussed briefly INPO's activities.
Current Issugg Fort Calhoun Extraction Steamline Break Event Mr. Jeffery L. Shackelford, Senior Reactor Analyst (SRA) discussed the extraction steamline break event that occurred on April 21, 1997 at.the Fort Calhoun nuclear plant along with the results of
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the investigation of the event by Region IV personnel.
On April j
21, 1997, at approximately 8:20 p.m., when the plant was operating at 100 percent power, the plant operators heard a loud noise coming from the turbine building.
The plant operators opened the control I
room access to the turbine building and noted steam emanating from j
the grating, which separates the main turbine deck from the areas i
below. The plant operators determined that a steamline rupture was in progress and manually tripped the reactor.
The reactor trip resulted in a turbine trip.
Following the reactor trip, emergency boration was initiated as a precautionary measure.
i Upon examination the licensee determined that the piping which had failed was the fourth-stage turbine extraction steamline, and that the most likely failure mode was flow-induced accelerated erosion.
The design conditions of the fourth-stage extraction system were i
300 psig/425 'F and the system was composed primarily of 12-inch-diameter piping fabricated from A-106 carbon steel with a nominal wall thickness of.375 inch.
The " fishmouth" break that occurred I
was approximately 4-feet long and 1-foot
- wide, and it was postulated that approximately 2-to 4-inch-wide by 4-foot-long section of pipe was below minimum wall thickness before the i
rupture.
The failure occurred on what is known as a "large radius elbow."
The as-found readings on the failed pipe revealed a minimum wall thickness of the rupture seam of.054 inch, whereas the code minimum allowable thickness for this piping is.126 inch.
The failure location was modeled in the licensee's erosion /
l corrosion program, but the actual wall thickness had never been measured by nondestructive examina, tion techniques.
The licensee had relied on a predictive methodology (CHECWORKS) to monitor the condition of the large radius elbows in the extraction steam system.
The CHECWORKS methodology had predicted a lower wear rate i
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Plant Ops. & Fire Prot.
-4 July 18, 1997 Joint Subcte. Minutes on the large radius elbows relative to other potential wear locations within the fourth-stage extraction steam system.
Finally, it was determined that the licensee's analytical model for predicting wear rates on the affected system components had not accurately predicted the actual wear rates and that the licensee over relied on the model's predictions for the extraction steam system.
These deficiencies led to significant degradation (i.e.,
eroded below minimum wall thickness requirements) in six separate piping locations ~in three separate plant systems.
One of these areas of degradation resulted in a catastrophic failure of the piping, which caused plant transient, created a significant hazard for personnel and contributed to a reduction in the station's fire protection capabilities.
The NRC staff is in the process of issuing an information notice on this event.
Fire Protection Issues Mr.
Chris A.
Vandenburgh,
- Chief, Engineering Branch, briefly discussed the results of the inspection conducted by his staff after a fire at the Arkansas Nuclear One (ANO) Unit 1 reactor 1
building on October 17, 1996.
The inspectors identified several problems in the licensee's implementation of the fire protection requirements of Appendix R to 10 CFR Part 50.
Most significantly, a recent ANO Unit 1
modification to replace Reactor Coolant Pump B failed to install an adequate lube oil leakage collection system, which resulted in an oil fire inside the reactor building.
The Region IV concern regarding this failure was heightened because of several missed opportunities to identify and correct the problem before the fire occurred.
Specifically, the design change that replaced the Reactor Coolant Pump motor did not undergo an adequate fire protection review. The oil leakage observed during maintenance was i
not thoroughly investigated.
Repeated observations of oil leakage by radiation protection personnel were not investigated.
Most importantly, the reports of oil-soaked insulation and excessive smoking during the plant heatup were not adequately communicated and reviewed. These instances indicated that the corrective action programs may not have adequately addressed fire hazards and fire protection deficiencies as significant conditions adverse to the quality and the safe operation of the plant.
The containment fire was caused by the ignition of oil that had accumulated in fibrous insulation on Steam Generator B.
The oil
- originated from a previously uncontrolled leakage from a crack in the discharge piping of the high-pressure lift oil pump of the lube oil system for the Reactor Coolant Pump.
The oil ignited below its i
normal auto-ignition temperature because of the " wicking ef fect" in i
fibrous insulation.
The event resulted in an enforcement action
' Plant' Ops. & Fire Prot. July la, 1997 Joint Subete. Minutes against the licensee involving its failure to notify the NRC that it had declared an unusual event during the fire.
Senior Reactor Procram Mr. William B. Jones, SRA, discussed the objectives of the senior reactor analyst. program and the responsibilities of the Senior Reactor Analysts.
He stated that Region IV has two SRAs.
The SRAs are responsible for reviewing NUREG-1560,
" Individual Plant Examination Program:
Perspectives on Reactor Safety and Plant Performance," (IPEs) The SRAs also are responsible for gathering developing comprehensive information for the Region IV staff that will include the licensee's probabilistic risk assessments (PRAs),
Individual Plant Examinations (IPEs), Individual Plant Examination of External Events (IPEEEs), technical staff evaluation reports, risk--informed pilot programs., and background information. The SRAs have the capability to perform independent risk assessments, as
- needed, using the NRC risk assessment
- tool, the integrated reliability and risk analysis system (IRRAS).
The SRAs will also provide the operator licensing examiners with risk-significant operator actions that wcald be important in mitigating an accident or in minimizing the failure of components and systems.
Online Maintenance Dr.
Dale A.
- Powers, Chief, Maintenance Branch, discussed the reasons for conducting online maintenance and the risk assessment tools for online maintenance.
He stated that the reasons for performing online maintenance include the following:
(1) it balances the availability and reliability of the
- systems, structures, and components (SSCs), and (2) it also has economic benefits for shortened refueling and maintenance outages.
Many regulations were developed during the period when industry's maintenance philosophy ' was to conduct significant maintenance activities during long outages.
Dr.
Powers stated that risk assessments should be performed to determine the potential risk associated with online maintenance activities that affect the
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reliability and uvailability of SSCs.
Recion IV Inspection Procram Mr.
Thomas P.
- Gwynn, Director, Director of Reactor Projects, discussed the Region IV inspection program, the status of plants, the resident inspection program, the SALP ratings of the Region IV
. plants and plant performance reviews.
The report on the status of the plants is provided to Region IV by the NRC Operations Center every, day.
Mr.
Gwynn provided the following examples of j
information in the plant status report:
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Plant Ops. & Fire Prot. July 18, 1997
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Joint Subete. Minutes
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Callaway is operating at 95-percent power because of an. axial-offset anomaly in its core.
This issue is being handled by NRC Headquarters.
I San Onofre Units 2 and 3 were coming back up to power when they found a problem with some check valves that had potential implications for Unit 2.
Both units have been shut down for a maintenance outage to resolve the check valve problem.
Additionally, Mr. Gwynn stated that the plant status report is now put on the agency',s World Wide Web Site on a daily basis.
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The inspection program is conducted in accordance with Manual j
Chapter 2515 and the regions are committed to perform the core j
inspection program at every plant.
In addition, the inspections performed by the region's on its own initiative, go beyond the core program and typically focus on areas in which problems existed in the past.
The resident inspection program was begun in 1977, before the j
accident at Three Mile Island.
The purpose of the resident inspection program is to provide continuing NRC onsite coverage at each nuclear power plant in the country, to provide for a rapid NRC response to plant events, and to increase inspection time and direct observation of the licencee's activities at the plant. This program gives the regions an enhanced knowledge of the conditions at licensed f acilities and a better basis for regulatory decisions.
i It also provides NRC additional assurance that management of station operations by each licensee is effective and that licensee performance is acceptable. The resident inspector requirements are divided into six inspection categories: daily, biweekly, monthly, tri-monthly, semiannually, and during outage.
The region-based inspection is conducted by region n inspectors who typically have a
higher level of specialized expertise than the resident inspectors.
The Plant Issues Matrix program is a new addition to the region's inspection program.
This program compiles a chronological listing i
of all of the key issues for each plant in the region.
The i
chronological listing of issues is collected from inspection reports, licensee event reports, and event notifications that have been provided by the licensees to NRC Headquarters Operations Center.
This program is a valuable tool that is used as part of both the performance review and as input to the senior management meetings.
This program also identifies each of the items as either a strength of the licensee's program or a weakness.
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Plant Ops. & Fire Prot. July 18, 1997
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Decommissionina and Dry Cask Storace Dr. D. Blair Spitzherg, Acting Chief, Nuclear Materials Inspection and Decommissioning Branch, briefly discussed the. decommissioning process at Region IV sites.
He stated that Region IV has a higher number of reactors being decommissioned than the other regions.
The major decommissioning activities are currently taking place at.
the Trojan Nuclear Plant which was shut down in 1992.
The licensee for Trojan has removed the steam generators, the reactor coolant pumps, and the pressurizer.
The reactor vessel and the internals remain on site.
Trojan is actively undergoing dismantlement and decontamination which are expected to be completed by the year 2002.
i Region IV has two facilities with dry cask storage: Arkansas
! Nuclear one Unit 1 and Fort St. Vrain.
Some problems related to j
- the dry fuel storage systems have been experienced at Arkansas Unit i
1.
The licensee has not completed its investigation to determine the nature of weld cracking phenomenon which has been observed at Arkansas Unit 1.
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on behalf of the subcommittees, Mr. Barton thanked Mr. Merschoff and members of his staff for the very informat.ive briefing on Region IV activities and the professional manner in which it was presented.
Uuture ACRS Action j
i The Subcommittees did not specify any further action in regard to i
this meeting.
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Additional details of this meeting can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, N.W.,
Washington, D.C.
20006, (202) 634-3274, ox, can be purchased from Neal R.
Gross & Co., Inc., Court Reporters and Transcribers,1323 Rhode Island Avenue, N.W., Washington, D.C. 20005, (202) j 234-4433.
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