ML18150A140
| ML18150A140 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 06/05/1987 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | NRC |
| References | |
| CON-#287-3701 2.206, DD-87-09, DD-87-9, NUDOCS 8706100177 | |
| Download: ML18150A140 (12) | |
Text
In the Matter of VIRGINIA ELECTRIC AND POWER COMPANY (Surry Power Station Units 1 and 2}
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- UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Thomas E. Murley DD-87-09
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Docket Nos. 50-280 and 50-281
( 10 CFR 2. 206)
DIRECTOR'S DECISION UNDER 10 CFR 2.206 INTRODUCTION By Western Union Mailgram dated December 11, 1986 (petition) and a supplement to that petition dated January 20, 1987, Mr. Thayer Cory and Ms. Judy Zwelling (the petitioners}, on behalf of Citizen Action for a Safe Environment (CASE), requested pursuant to 10 CFR 2~206 of the COl1l"Tlission's regulations that the Director of Nuclear Reactor Regulation require Virginia Electric and Power Company (the licensee) to show cause why reopening the Surry Nuclear Power Station would not endanger the health and safety of the co1T1T1Unity and that the Nuclear Regulatory Commission {NRC) issue an order directing that both reactors at Surry Nuclear.Power Station remain shut down until the licensee takes a *number of actions discussed in their petition. The petitioners requested that both reactors at the Surry Power Station remain shut down until the licensee fully inspected all pipes* and publicly issued a complete report on the December 9~
1986 pipe break accident at Surry Unit 2.
In addition, the petitioners requested that NRC order the licensee to keep the Surry Power Station* shut down until the licensee demonstrated that the plant complied with NRC regulations and could be operated with reasonable assurance of safety. The petitioners
e also requested that, before the Surry Power Station be allowed to resume operation, NRC condition the Surry operating license to require the licensee to inspect all piping handling condensate and feedwater during each refueling outage, including inspection of all check valves. The petitioners presented several bases for the requested actions, including the following:
- 1.
the December 9, 1986 pipe-break accident at Surry Unit 2;
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an alleged ongoing pattern of violations fn areas such as plant operations, surveillance, fire protection, radiological control, emergency preparedness, security and safeguards, quality assurance, and administrative control affecting quality;
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inadequacies fn the emergency alert system and evacuation plans for the Surry Power Station; and
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alleged falsification of welder verification and an allegedly poor quality assurance program.
Shortly after the pipe rupture event, the NRC fonned an Augmented Inspection Team (AIT) and dispatched ft to the Surry site. The purpose of the AIT was to augment the inspection efforts by the Senior Resident Inspector at Surry in identifying the cause of the event and in monitoring followup actions taken by the licensee. In addition to the AIT inspection activities, other inspectors knowledgeable in security, fire protection system, water chemistry, and check valve design were assigned to review specific concerns in these areas. The NRC staff issued an Augmented Inspection Team Report (AIT Report) on February 10, 1987. In the AIT Report, the staff reviewed the detailed infonnation on the December 9, 1986 event, including a recovery plan and corrective actions the licensee planned to take before restart. The staff agreed with the licensee's recovery plan and the corrective actions planned before restart.
_I By letter dated February 13, 1987 from Richard H. Vollmer to the petitioners, the NRC acknowledged receiving your petition, addressed a majority of the issues raised, and concluded that delaying restart of plant operation at_ the Surry Power Station until all the issues raised in the petition had been resolved was not warranted. Unit 1 resumed operation on February 23, 1987, and
-Unit 2 resumed operation on March.20, 1987.
On March 30, 1987, we sent you the AIT Report. For the reasons stated in this Decision, your requests are denied.
My Decision in this matter follows.
DISCUSSION.
On December 9, 1986, while Surry Power Station Unit 2 was operating at full power, an 18-inch suction line to main feedwater pump A failed catastrophi-cally. The break was located in an elbow in the 18-inch-diameter line about 1 foot from the 24-inch-diameter header. The unit was taken to cold shutdown by the morning following the accident. The pipe is believed to have failed in Surry Unit 2 because the pipe wall had become thin as a result of single-phase flow erosion/corrosion mechanisms.
The failure in the December 9 accident occurred in piping that is not essential for shutting down the reactor or for preventing or mitigating the consequences of an accident. The NRC does not classify this piping as "safety-related" piping, and therefore, no specific re-quirements are* imposed on its licensees to inspect such piping.
In contrast, the NRC considers reactor coolant piping safety related and specifically requires its licensees to inspect such piping periodically as required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Reactor coolant piping is fabricated from stainless steel and is not considered to be susceptible to degradation by erosion/corrosion mechanisms such as occurred ;n the carbon steel feedwater piping at the Surry facility.
e -Experimental results and field experience have shown that stainless steel exposed to reactor coolant flow is highly resistant to erosion/corrosion. The loss-of-nonnal-feedwater events have been analyzed in the Final Safety Analysis Report for Surry and there is no reason to conclude that the recent Surry 2 event exceeded those analyses.
By a letter dated January 14, 1987 from W. L. Stewart to Dr. J. Nelson Grace, the licensee forwarded to NRC a report entitled "Surry Unit 2 Reactor Trip and I
Feedwater Pipe Failure Report, Revision 0, January 14, 1987." This report provided detailed information on the December 9, 1986 accident and the licensee's recovery plan and planned corrective actions. Before Surry began to operate aga*i n, the NRC reviewed the report and concurred with the 1 i censee 's actions.
After reviewing this report, we concluded that the licensee had taken appropr1ate steps to inspect those piping systems that are susceptible to erosion/corrosion mechanisms, and identified the exte~t of the degradation of piping 1n both units. The licensee has taken appropriate actions to provide reasonable assurance that both units can be operated without posing any undue risk to the health and safety of the public. The corrective actions include replacing some piping and modifying other piping, as well *as increasing inspections, as appropriate. Although the licensee did not inspect all piping at the Surry facilities as you requested in your petition, the licensee did inspect piping that is highly susceptible to erosion/corrosion mechanisms experienced *in feedwater system. In addition, the licensee inspected several safety-related piping systems (i.e., auxiliary feedwater and chemical and volume control systems) in which erosion/corrosion is not expected. The NRC found the scope of these inspections sufficient to pennit the facility to operate. Based on the review of infonnation provided by the licensee and generic evaluation discussed in the next paragraph, the staff concludes reasonable assurance exists that other safety-related piping at Surry is not subject to the type of erosion/
corrosion mechanisms that occurred in the feedwater piping.
On February 10, 1987, the NRC issued the AIT Report, documenting the staff's finding about the December 9, 1986 event and concurring on licensee's recovery plan and actions to take before restarting power production at the Surry Plant.
The NRC staff is continuing to evaluate the generic implications of the erosion/corrosion mechanisms that occurred at the Surry plant. On January 15, 1987, nuclear power industry and NRC experts from several engineering disciplines met to discuss the failure mechanism in feedwater piping at Surry Unit 2.
From the panel discussion, it was generally agreed that the important variables influencing the erosion/corrosion mechanisms are: material, local fluid velocity/turbulence, water chemistry, and operating temperature of the system.
It should be noted that the ASME standards for inservice inspection of piping only require inspection of the welds, not a more general inspection of the pipe wall for thinning. This is based on the experience that most failures result from cracks near welds, not from thinning of the sort that occurred at Surry.
The NRC staff is collecting additional information from various plants and will make appropriate generic recommendations after analyzing all pertinent data collected from the industry.
The February 10, 1987 AIT Report also evaluated the cause ~f the failure of the main steam trip valve which you mentioned.
The staff concluded that the valve was tested to verify that it was in compliance with Technical Specifications, and did not fail to perform its safety function.
However, the licensee did not provide adequate detailed instructions in maintenance procedures for corrective maintenance. The staff issued a Notice of Violation to the licensee. In
. accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR 2, Appendix C (1986), this violation was considered to be a Severity Level IV.violation. This level of violation did not warrant suspending plant operation.
The AIT Report also addressed in detail the issue of check valve-failure, concluding that the failure of the c~eck valve did not cause the pipe rupture.
However, the condition of the check valve would have contributed to the amount of feedwater that came out from the feedline break and, possibly, the extent of I
the pipewhip that followed.
The licensee has inspected other Unit 1 and Unit 2 feedwater pump check valves and modified the valve internals. Because these valves are not considered to be safety related, they do not have specific inspection requirements similar to-those required for safety-related components.
The staff is considering this issue on a generic basis and will make appropriate changes, ff necessary.
In addition, you alleged that an ongoing pattern of violations existed in the areas of plant operation, surveillance, fire protection, radiological controls, emergency preparedness, security and safeguards, quality assurance, and administrative control, and that these affected quality.
You stated that you believed the licensee should have been required to take action before restarting the units. It should be noted that the NRC staff identified these violations earlier in the Systematic Assessment of Licensee Performance (SALP) reports. The staff uses the reports to:
(1) improve the NRC regulatory program and pennit sound decisions regarding NRC resource allocations, (2) improve licensee perfonnance, and (3) collect available observations on a periodic basis and evaluate licensee performance based on those observations through an integrated NRC staff effort. Positive and negative attributes of licensee perfonnance are considered. The SALP process is oriented to improve NRC's understanding of the manner in which:
(1) licensee management directs,
guides, and provides resources for ensuring plant safety and (2) such resources are used and applied. The integrated SALP assessment is intended to provide enough diagnosis to offer a rational basis for allocating NRC resources and to provide meaningful guidance to licensee management. All of the violations you detailed had already been factored into NRC's evaluation of the licensee's performance. These violations are important considerations in assessing the licensee's perfonnance, and all of these violations have been reviewed and I
evaluated by the staff on a case-by-case basis. These violations are categorized (in accordance with NRC regulation in 10 CFR 2, Appendix C) in terms of five levels of severity to show their relative importance. Severity Level I is assigned to violations that are the most significant in terms of public safety; Severity Level V violations are the least significant. The severity levels assigned to the violations you wrote to us about did not warrant suspension of the plant operation. In addition, as indicated in a recent SALP report, the staff found the licensee's overall performance to be satisfactory in all areas.
Therefore, it is not appropriate to suspend plant operation on the basis of these violations.
You also raised the issue of plant aging by referring to a report from the National Research Council of the National Academy of Sciences, entitled "Revitali-zing Nuclear Safety Research," and requested that before restarting the plant, the licensee should be required to work with the NRC and the Department of Energy
{DOE) to implement an extensive research program on plant aging. It should be noted that the report you referenced discussed the research activities required to extend the lifetimes of plants beyond current license periods. Moreover, the licensee already actively participates in a research program for extending the lifetimes of existing facilities in cooperation with NRC and DOE.
In addition, the surveillance and maintenance practices that are implemented in accordance with the ASME Code and the facility Technical Specifications provide reasonable assurance that any unexpected degradation fn safety-related components in the plant will be identified and corrected during current license periods.
Therefore, I do not believe ft fs appropriate to suspend the plant operation on the basis of the generic aging considerations you asserted.
You also raised the issue of allegations brought in 1985 against the licensee by members of th~ International Brotherhood of Boilennakers that raised the iss_ues of falsification of welder certifications, poor workmanship, and health and safety concerns at the Surry Power Station. Because the NRC Office of Investigations is currently reviewing these allegations, and because no significant safety concerns have been identified at this time by the NRC staff, ft is not now appropriate to suspend the operation of the Surry Power Station on the basis of this issue.
You also expressed concerns about high radia~ion.exposure to the workers in the Surry Power Station and the potential deficiencies fn the quality assurance program.
We addressed these issues in detail fn our SALP report dated December 11, 1986. The NRC staff found the licensee's overall perfonnance in these two areas satisfactory and rated the licensee's perfonnance in these areas at SALP Level II. indicating that the licensee perfonned at the industry average and better than minimally satisfactory. The licensee currently meets all applicable radiation protection and quality assurance requirements.
. Consequently, there is no basis to ta"ke any actions based on licensee's perfor-mance in the radiation protection and quality assurance areas.
In the area of emergency planning, you asserted that the siren system for Surry Power Station and the surrounding corrmunities is inadequate because it is unreliable and in some cases fs inaudible to certain residents.
e Furthermore, you expressed concerns about the feasibility of existing evacuation plans.
10 CFR 50.54(q) requires the licensee to have a satisfactory emergency plan that meets certain criteria established in 10 CFR 50.47(b) and Appendix E to 10 CFR 50.
The licensee's emergency plan. has been upgraded to meet the requirements of these rules, including the installation and testing of a public alert and notification system.
On May 13, 1983, the NRC reviewed the onsite emergency plans and level of onsite preparedness and found them acceptable.
L Pursuant to 10 CFR 50.47(a)(2), the NRC bases its finding on the adequacy of offsite emergency plans on the findings and determinations of the Federal
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Emergency Management Agency (FEMA). The FEMA evaluation of the licensee's level of offsite preparedness included a review of State and local plans and the observation of full-scale exercises. Specifically, FEMA approved the Conmonwealth of Virginia's State and local emergency plans and level of preparedness for the Surry Power Station in February 1983, under the 44 CFR 350
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rule. This approval was conditioned with successful demonstration of adequacy of the public alerting and notification system in accordance with the standards set forth in Appendix 3 of the NRC/FEMA criteria of Revision 1 of NUREG-0654/
FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plan and Preparedness in Support of Nuclear Power Plants," and the subsequently published'standards in FEMA-43, "Standard Guide for the Evaluation of Alert and Notification Systems for Nuclear Plants." The last full-participa-tion exercise of the offsite preparedness plans for Surry Power Station was conducted on October 4, 1985, and State and local authorities participated fully.
The evaluation prepared by FEMA did not identify any significant deficiencies in the exercise. Moreover, on the basis of the results of this exercise, FEMA.
~oncluded that offsite radiological emergency preparedness is adequate to provide reasonable assurance that appropriate measures can be taken offsite to
e e protect the heatlh and safety of people living near the site in the event of a radiological emergency. Therefore, the 44 CFR 350 approval granted in February 1983 remains in effect.
In June 1986, the licensee conducted a demonstration of the adequacy of the public alerting and notification system as required by FEMA's conditional approval in 1983. The Federal Emergency Management Agency is currently evaluating the results from that exerci,se. The licensee has stated that the system was designed to meet the objectives for area coverage in the times prescribed by Appendix 3, NUREG-0654/FEMA-REP-l.
The NRC will consider FEMA's evaluation when it is received.
At this time, we have no basis for taking any action based on the emergency planning concerns raised in your petition.
CONCLUSION For the reasons discussed above, no basis exists for initiating a show cause proceeding pursuant to 10 CFR 2.202 and no basis exists to suspend operation of the Surry Nuclear Power Plant. Consequently, your petition is denied.
A copy of this decision will be filed with the Secretary for the Commission's review in accordance with 10 CFR 2.206(c).
Dated at Bethesda, Maryland this 5th day of June, 1987
- ~'i Thomas E. Murley, Director Office of Nuclear Reactor Regulation
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e UNITED STATES NUCLEAR REGULATORY COMMISSION VIRGINIA ELECTRIC AND POWER CO~PANY SURRY POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-280 AND 50-281 NOTICE OF ISSUANCE OF DIRECTOR'S DECISION Notice is hereby given.that the Director, Office of Nuclear Reactor Regulation, has issued a decision concerning a request filed pursuant to 7590-01 10 CFR 2~206 by Mr. Thayer Cory and Ms. Judy Zwelling on behalf of Citizen Action for a Safe Environment which requested that both reactors at the Surry Power Station remain shut down until all pipes had been fully inspected, until a complete report on the December 9, 1986 accident had been issued publicly by Virginia Electric and Powe~_ Company, and until all issues had been resolved.
The Director of the Office of Nuclear ReactorJegulation has detennined that the Petition should be denied. The reasons for this decision are explain~d 1n the "Oirector's Decision Under 10 CFR 2.206, 0 DD-87-09, which is available for public inspection in the Commission's Public Document Room, 1717 H Street, N.W., Washington, DC and at the Local Public Document Room at the Swem Library, College of William and Mary, Williamsburg, Virginia 23185.
A copy of the Decision will be filed with the Secretary for the Conmission's review in accordance with 10 CFR 2.206{c). As provided in this regulation, the Decision will constitute the final action of the Conmission twenty-five
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(25) days after issuance, unless the Cormrissfon, on its own motion, institutes review of the Decision within that time period.
Dated at Bethesda, Maryland, this 5th day of June, 1987.
FOR THE NUCLEAR REGULATORY COMMISSION Clu ~L.t P P a_t..J.
Chandu P. Patel, Project Manager Project Directorate II-2 Division of Reactor Projects-I/II