ML20204J362
| ML20204J362 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/20/1987 |
| From: | Zech L NRC COMMISSION (OCM) |
| To: | Markey E HOUSE OF REP., ENERGY & COMMERCE |
| Shared Package | |
| ML18150A013 | List: |
| References | |
| NUDOCS 8703270224 | |
| Download: ML20204J362 (11) | |
Text
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79 UNITED STATES '
y j 3,//c-(g h/ g NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555
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/g t h. 4,g March 20, 1987 CHAIRMAN The Honorable Edward J. Markey Committee on Energy and Commerce
.U.S.
House of Representatives Washington, DC 20515
Dear Congressman Markey:
I am responding to your letter of Jaruary 30, 1987, concernino the NRC's Office of Analysis and Evaluation of Operational Data (AE0Di 1984 Engineering Evaluation Report on Erosion in Nuclear Power.
Plants.
As mantioned in your letter, the AEOD report identified 140 events related to erosion of various components.
The report is an AE0D engineering evaluation and, like other engineering evaluations, was a general review to assess whether or not a more detailed case study was needed.
Since the report is an engineering evaluation, it contains only suggestions for review ard consideration.
The report.does-not recommend any action on the part of NRC.
This is' consistent with NRC policy that AE00 office recommendations be made on the basis of case studies which are more formal and require extensive reviews including technical community peer review.
The NRC program o'fices of Nuclear Reactor Regulation (NPRi, Inspection and En'orcement (IE), Nuclear flaterial. Safety and Safeguards (NMSS) and Nuclear Regulatory Research (RES) evaluated the report to determine the need for immediate actinn, long term action, prioritization as a new generic issue, or additional study.
The sta#f concluded that no immediate action was needed.
The AE0D report states that "there does not appear to be a direct-relationship between these events and a specific sa#ety oroblem that needs immediate attention."
The staff was aware of the ongoing efforts by industry, begun two years earlier, to address erosion in the types of oiping systems identified in the AE00 report.
The report was distributed to staff for review of the long term and generic safety implications.
Although the AE00 report provided useful information for ongoinq NRC nrograms, the reviaw did not identify the need for any new generic issues.
The instances of erosion-corrosion noted in the AE00= report occurred in two phase systems (steam containing soma moisturel, sinnie phase systems with suspended solids, and single phase systems in
' bypass piping downstream of orifices or valves installed to
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control flow.
These problems had already been identified within the industry and appropriately acted upon, in some cases several years before the AE0D report was issued.
Several utilities had implemented voluntary programs prior to the issuance of NRC IE Information Notice 82-22 and Institute of Nuclear Power Operation (INPO) Significant Operator Event Ranort (SOER) 82-11.
Erosion-corrosion-in single phase systems of carbon steel material had also been known to cause thinning in heat exchangers at regions of local flow disturbance, such as J-tubes in certain nuclear steam generators, and in conventional boiler feedwater tube inlets.
Voluntary programs have been used to monitor and repair such thinning.
Gross thinning due to erosion-corrosion in large diameter single phase piping systems, resulting in the kind of catastrophic oioing failure experienced at Surrv, had not been previously exoerienced.
Instead, more localized through-wall thinning with leakage has been experienced, and this has been limited to minimum flow bypass i
lines.
The NRC is extremely concerned about any degradation of nuclear power plant equipnent which has relevance to the safety of the plant.
We will continue to monitor individual plant performance j
and ov'erall industry experience in this area.
Where olant specific problems occur, the need for generic action will be
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assessed and corrective measures will be evaluated for their acceptability and effectiveness.
Answers to the specific questions in your January 30, 1087 letter j
are enclosed.
Commissioner Asselstine disagrees with this letter.
He will provide his views in a separate letter.
l Sincerely, Gbvde Ov.
\\
Lando W. 2.
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- 1. Answers to Specific Ouestions 2.
IE Information Notice No. 82-22
- 3. Relevant Internal Letters and Memoranda 4
Report on the Technical Meeting dated 1/30/87
- 5. Minutes of Technical Meeting dated 2/4/87 cc:
Rep. Philip Sharp
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FNCLnSURE 1 QUESTION 1.
When did regular testing and monitoring of steam ~
pipes begin?
What promoted this action?
Dlease.
provide all appropriate documentation.
ANSWER.
Regular testing and monitoring of safety-related oicing beoan on a generic basis when the American Society of Mechanical Enqineers issued Section XI of the ASME Boiler and Pressure Vessel Code in 1970.
Regulatory requirements for these inspections are set forth in 10 CFR 50.55a(g).
The ASME Code requires that certain welds and heat affected base material adjacent to welds in piping be exanined periodically.
The purpose of these examinations is to detect cracks in these areas so that corrective action can be taken to protect the integrity of the pioing pressure boundary.
Neither the Code nor NRC regulations require the examination of pipe walls outside of welds or associated heat affected zones.
In addition to the above reauirements, we understand that some licensees have voluntarily been examinino piping weldments and associated haat affected base material in non-safety related i
piping including those in steam lines in this category.
In addition to the measurement of weldnents and heat affected zones for cracking, some licensees voluntarily initiated inspection programs for pipe wall thinning, either before or,iust after the issuances of NRC IE Information Notice 82-?? and INPO SOER 82-11 in 1982.
The issuance of IE Motice 87 '? was due to failures at 4
several plants in 1982 in non-safety related steam extraction lines.
A copy of IEN 82-22 is enclosed.
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OUESTION 2.
Why did the NRC not require testing and monitoring on water systems, given the data in the AE00 Engineering Evaluation Report?
Please explain in detail, including all internal letters and memoranda on this subject.
ANSWER.
The data in the AE00 report (the frequency and consequences of pipe erosion problem) and the informatfor available prior to the AE00 report, did not suggest that new NRC regulatory action was needed to protect plant workers or the public from any undue radiological health risks.
As discussed in the response to-question #5, the staff is evaluating the effects of considering on-site non-radiological infuries or fatalities on our prioritization process.
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Copies of all relevant internal letters and memoranda are enclosed.
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r QUESTION 3.
Did the,NRC provide licensees with the June 11, i
1984 Report on Erosion in Nuclear Power Plants, or any of its conclusions?
If so, when?
If not, why not?
i ANSWER.
In accordance with standard NRC nolicy on AEOD Engineering 3
Evaluations, this study was provided to licensees and the industry in general in the following wavs:
1.
Copies of the " Transmittal of AE00 Engineering and-4 Technical Evaluation Reports"'(July 17, 1984 memorandum for C. J. Heltemes) were sent to the institute for Nuclear Power Operations (INP0); the Nuclear Safety Analysis Center (NSAC) and the Nuclear Operations 2
Analysis Center (NOAC) at the Oak D.idge National Laboratory.
2.
An abstract of the Engineering Evaluation was published in the November 1984 i ssue of " Dower Reactor Events" (NUREG/BR-0051 Vol.6, No. 3).
All licensees receive at least three copies of the Power Reactor Events publication with one copy going directiv to the plant manager.
Instructions wera.also included for obtaininq a copy-of the report i tself from the P!PC Public Document Room.
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QUESTION 4.
Are there other systems the NRC has identified which were expected to last for-the life of the plant but.which may be sub.fect to early failure?
Please identify each such systen, and any NRC-research or inspection effort dedicated.to each' system.
ANSWER.
NRC has a number of research prograns in place which are investigating both systems and components for their potential for early failure due to aging effects.
These programs are desinned to identify specific aging mechanisms and determine what steos will be necessary to deal with the aging ohenomenon.
Results from this research will be usad to determine whether requirements mandating replacement of specific systens-or components at designated times are needed.
While our knowledge and understanding of aging mechanisms and their effect on systems and components is increased through our research programs, we believe that most aging phenomena can he readily managed and do not pose a safety problem provided the necessary compensatory measures (maintenance, surveillance, repair, replacement, etc.) are effectively implemented. -Such measures have identified portions of piping systems, valves, cumps, electrical items, steam generators and other items in nuclear plants requiring repair or replacement.
With proper maintenance and surveillance oronrams in olace, combined with required inservice inspection and test of safety related systems and components, the effect of aqing is being adequately addressed by the utilities.
As results from our aning research indicate the need for change, such as increased inspection frequency, changes will be made wherever appropriate.
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t QUESTION 5.
What actions does the NRC plan in response to the Surry accident?
Does the Commission expect increased inspection of secondary systems?
If not, why not?
ANSWER The Nuclear Regulatory Commission has organized a coordinated interoffice effort addressing the many asoects of the Surry accident involving the nffices of Nuclear Reactor Requiation, Inspection and Enforcement and Nuclear Regulatory Research.
NRC is coordinating its efforts with the Flectric Power Research Institute (EPRI) and the Institute for Nuclear Power Operations (INP01 One of the first actions taken by the NRC was to convene a panel of nationally recognized experts in the technical disciplines related to the Surry failure.
This panel met in an open public meeting on January 15, 1987.
Two menoranda related to that meeting are enclosed.
The technical review of the Surry failure at the.lanuary 15th meeting (in which there was extensive industry participation) provided some of the basis for tha NRC coordinated effort.
The specific tasks which the NRC Offices have underway are the following:
a)
Prioritization of the Surry event using the criteria described in NUREG 0933 to establish the safety significance, recognizing that the failure of feedwater piping is a required design basis event for the plant.
b)
Review of the method for treating on-site non-radiological injuries or fatalities as part of the prioritization and cost-benefit analysis processes in order to determine whether or not there is need for revision to the prioritization process for these events.
c)
Preparation of a report to the appropriate national codes and standards bodies to make them aware of the facts surrounding the Surry event so that they may consider appropriate revisions to their documents.
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OUESTIdN 5.
(Continued) d)
Preparation and issuance of a Supolement to the IE Information Notice IN 86-106 to provide the industry with additional information made available'since the -
event.
e)
Collection of information fron licensees related to wall thinning measurements together with correlation and dissemination of the data.
f)
Review of the role of states, OSHA and the NRC dealing with non-radiological accidents.
- Depending on the outcome of the above tasks., particularly those relative to safety significance of events in non-safety related systems with potential for affecting only on-site personnel, action will be taken to ensure that any necessary-inspections of secondary non-safety related systems will be made.
The results of a preliminary survey made by the utilities indicates that of 29 plants which have inspected main feedwater piping to date, all have measured pipe wall thickness well within allowable limits.
Only the two Surry units have detected severe wall thinning of this oiping. INP0 prnmotly issued a Significant Event Report following the event and is' evaluating need for further action.
EPRI has in preparation a technical white' paper to provide information on the subject.
OUESTION 6.
Please provide a status report regarding the Commission's review of plant life extension.
Include in the status report all NRC. programs and industry programs dedicated to this effort, how much money is being spent, and what technical ouestions are beino addressed.
ANSWER' The Commission has initiated the development of a policy on extending nuclear power plant licenses beyond 40 years.
A request for public comment regarding this policy was, published in the Federal Register on Nnvenber 6, 1986 with a 60 day comment period.
Due to-numerous requests for extension, the comment period was extended until February 2, 1987.
The sta## hopes to brief the Commissinn by May 30, 1987 on the public comments received.
The staff plans to provide the Commission with an Information Paper by Sept. 1987 reporting on the possible options for relicensing and staff plans for completion of the development of a proposed policy on license.'enewal.
A Commission Paper on Proposed License Renewal Policy is planned for Sept. 1988, with any required necessary implementinq reaulations to be published for initial comment about Sept. 1989.
We believe that the final regulation might become effective during 199?.
The NRC funded research programs dealing with the technical basis in suoport of life extension will cost $1.15M in FY 87 and about $2.7M in FY 88.
The above fioures represent work clearly identified as associated with decisions regarding operation beyond normal operating life, usually about 40 years.
However, a great daal of our work, especially associated with aging, will orovide an essential base for making life extension decisions.
Some of the major policy questions to be addressed concerning life extension
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are contained in the Federal Register notice of November 6, 1986.
The technical questions to be addressed are associated with the ability to identify, understand and predict potential dearadation mechanisms such as fatigue, radiation damage, wear, creep, erosion, and water hammer their effect on service life of' specific components and systems, and the effect of degradation on overall plant safety.
Industry has been increasingly active over the last few years in the area of license renewal beyond 40 years.
A joint industry, EPRI, and DOE program was initiated in 1984 to identify issues associated with life extension.
In 1985, detailed technical studies funded.iointly by DOE and EPRI were started on two pilot
QllESTION 6.
(Continued) plants (Surry 1 and Monticello) and a study on the regulatory aspects of life extension was commissioned by AIF's National Environmental Studies Project.
Preliminary results were presented at a joint industry /EPRI/n0E life extension seninar in 4
August 1086.
Industry has established a framework to continue its efforts and integrate life extension activities.
A steerino committee called NUPLEX (Nuclear Plant life Extension) has been established to coordinate the efforts of several working arouos focusing on technical activities, codes and standards, licensing natters, and policy considerations.
Organizations which generate National Consensus Codes and Standards, particularly those associated with our national professional technical societies, such as ASME, IEEE, and ASCE,_
are also becoming very-active in this area.
ASME is planning to establish a special coordinating committee to direct the overall thrust of plant life extension activities associated with all national standards writing bodies.
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UNITED STATES i
's NUCLEAR REGULATORY COMMISSION 5,
[
W ASHING TON, D. C. 20555
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,, gy OFFICE OF THE SECRETARY NOTE FOR: Document Control Desk FROM:
Correspondence & Rec s
The enclosed document (s) are to be entered into the DCS. An advanced has been sent to the Public Document Room.
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ENCLOSURE 2 1
UNITED STATES NUCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEM WASHINGTON, D. C.
20555 July 9, 1982 FAILURES IN TURBINE EXHAUST LINES IE INFORMATION NOTICE NO. 82-22:
Addressees:
All nuclear power reactor facilities holding an operating license or construction permit.
Purpose:
This information notice is provided as an early notification of events that may It is expected that recipients will review the have safety significance.
No specific action or information for applicability to their facilities.
response is required at this time.
Description of Circumstances:
Duke Power Company's Oc~onee Unit 2, while operating at 95%
rupture of a 24-in.-diameter, long-radius elbow in On June 28, 1982, power, experienced a 4-f t2the feedwater heat extraction line which is supplied st The rupture has been attributed to piping degradation t turbine exhaust.
results from steam erosion.
the absence of main steam (turbine header) line pressure, believed a steamline The main turbine break had occurred and manually tripped the tsactor.
l-Systems and related automatically tripped as a result of the reactor trip.
parameters responded as expected following the reactor trip and subsequent recovery.
The steam jet destroyed a non-safety-related electrical load center in the lower elevation of the turbine building and certain non-safety-related instru-mentation in the vicinity, but did not render any essential equipment inoperable.
Two persons suffered steam burns, serious enough to.be hospitalized overnight.
i Initial indication of extraction steamline degradation at the Oconee facility was discovered in 1976 when a pinhole leak occurred on a similar line in Unit 3.
Subsequent to this discovery, a maintenance surveillance program utilizing In 1980, two ultrasonic examination of extraction steam lines was begun.
elbows on Unit 3 identical to the failed elbow on Unit'2 were replaced because I
of steam erosion.
In March 1982, prior to the failure, ultrasonic-inspection revealed substantial erosion of the Unit 2 elbow in the extraction.line; however, the erosion was
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The licensee theorizes that less than the licensee's criterion for rejection.
-0204210392-1 i
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TN 82-22 l
July 9, 1982
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Page 2 of 2 sustained reduced power operation and resultant lower quality steam contributed to accelerated erosion and failure of the elbow.
Additionally, the previous inspection program may not have identified the section where the line was thinnest.
The failed elbow and the identical elbow on the other "C" feedwater heater supply line'.have been replaced.
The two corresponding 24-in, elbows on Oconee Unit 1 were ultrasonically inspected on July 1, 1982.
The inspection revealed that a 4-in.2 area in one of the elbows had been reduced in thickness from 0.375 in. to 0.100 in.
Power operation was subsequently adjusted to 80% and the "C" feedwater heaters isolated so that the affected steam line was maintained at 125 psig with no steam flow.
The factor of safety based on material yield at the above pressure is slightly greater than two.
Elbow replacement will be performed on Unit 1 after returning Unit 2 to power.
In addition, the Institute of Nuclear Power Operations (INPO) has identified four other similar failures of steamlines also apparently resulting from steam erosion.
These failures resulted in plant shutdown.
They are Vermont Yankee on 1/27/82; Trojan.1 on 1/9/82; Zion 1 on 2/12/82; and Browns Ferry 1 on 6/24/82.
For example, in Vermont Yankee, a leak occurred in the 12-in.-diameter drain line between the "C" moisture separator and the heater drain tank, blowing steam into the heater bay area.
On Zion 1 a steam leak occurred in the 150 psig high pressure exhaust steam line from the Unit 1 turbine.
The leak originated from an 8-in. crack on a weld joining 24-in.-diameter piping with the 37.5-in.-diameter high pressure steam exhaust pipe leading to the moisture separator reheater.
INPO will issue a Significant Event Report on Nuclear Notepad shortly and is preparing a Significant Operating Experience Report which is expected to contain recommendations on this subject.
If you have any questions regardir.g this matter please contact the Regional Administrator of the appropriate NRC Regional Office, or this office.
rd. Jordan, Director Divi ~ n of Engineering and e
Qua ity Assurance Office of Inspection and Enforcement Technical
Contact:
- 0. Merrill 301-492-4513
Attachment:
List of Recently Is_ sued IE Information Notices