ML19260C487
| ML19260C487 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| REF-GTECI-A-03, REF-GTECI-A-04, REF-GTECI-A-05, REF-GTECI-SG, TASK-A-03, TASK-A-04, TASK-A-05, TASK-A-3, TASK-A-4, TASK-A-5, TASK-OR NUDOCS 8001030701 | |
| Download: ML19260C487 (12) | |
Text
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WESTINGHOUSE, COMBUSTION ENGINEERING & BABC0CK AND WILC0X STEAM GENERATOR TUBE INTEGRITY Lead NRR Organization:
J1 vision of Operating Reactors (DOR)
Lead Supervisor:
Darrell G. Eisenhut, Acting Director, DOR
! 2 Task Managers:
Westinghouse (A-3):
J. Strosnider Combustion Engineering (A-4): F. Almeter Babcock & Wilcox (A-5): J. Strosnider Applicability:
Westinghouse, Combustion Engineering, and Babcock & Wilcox Pressurized Water Reactors Projected Completion Date:
May 1980 1700 084 soci0307O t
TASK A-3, A-4, and A-S REV. No. 2 August 1979 1.
DESCRIPTION OF PROBLEM Pressurized water reactor steam generator tube integrity can be 2
degraded by corrosion induced wastage, cracking, reduction in the tube diameter (denting) and vibration induced fatigue cracks. The primary concern of these Task Action Plans is the capability of degraded tubes to maintain their integrity during nonpal operation and under postulated accident conditions (LOCA or a main steam line break) with adequate safety margins and the establishment of inspection and plugging criteria needed to provide assurance of such integrity.
Westinghouse (W) and Combustion Engineering (CE) steam generator tubes have suffered degradation due to wastage and stress corrosion cracking.
Both types of degradation ha'te been decreased Dy changes in secondary water chemistry. Degradation due to denting which leads to primary side stress corrosion cracks is the major form of tube degradation at present. The extent of steam generator tube degradation has been less severe in Babcock and Wilcox (bow) steam generators than in W or CE. The most significant form of tube degradation in B&W generators has been cracks of unknown origin propagated circumferentially by flow induced vibration. This phenomenon has been limited to a localized area of tubes adjacent to an open inspection lane in the steam generators and has only occurred in the Oconee Units. A second form of tuDe degradation described as an erosion-corrosion phenomenon nas been observed at Oconee and other B&W units.
2.
PLAN FOR PROBLEM RESOLUTION The approach taken in the Generic Task Action Plans is to integrate technical studies in the three area of systems analyses, inservice inspection, and tube integrity to estaolish improved criteria to ensure adequate tube integrity and safe steam generator operation under all conditions.
Improved criteria will be developed for tube plugging, inservice inspection, and steam generator design and operation. The purpose of the systems analyses is to evaluate the consequences of different numbers of steam generator tube failures during postulated accident conditions (LOCA, MSLB) considering predicted fuel Dehavior, ECCS performance, radiologial consequences, and containment response.
The results will be used to define a tolerable level of steam genera-tor tube leakage during postulated accidents. The major emphasis in the inservice inspection portion of the tasks is to develop a statistically based inservice inspection program which will provide assurance that no more than the tolerable level of tube leakage oefined 1700 085 l2 by the systems analyses would occur in an accioent. The tube integrity portion of the tasks is primarily concerned with experimental verifica-tion of tube behavior during postulated acciaents, development of tube plugging criteria, and definiton of operating procedures and design to minimize tube degradation. Specific tasks which must be performeo in each of the above areas include the following.
A. SYSTEMS ANALYSES OF TRANSIENTS AND POSTULATED ACCIDENTS 1.
Analyses of LOCA With Concurrent Steam Generator Tube Failures LOCA analyses already performed will be reviewea to cefine a tolerable level of secondary to primary leakage through the steam generator based on fuel integrity, ECCS performance, ano containment response consicerations. Adcitional calculations will be performed as needed.
2.
Analyses of MSLB With Concurrent Steam Generator Tube Faiulures MSLB analyses already performed will be reviewed to cefine a tolerable level of primary to seconaary leakage through the steam generator during a steamline break. For a break insice containment, the tolerable primary to secondary leakage will be based on containment response considerations. For a break outside containment, the tolerable primary to seconaary leakage will be based on raciological considerations including fuel behavior during the accident.
Bounding calculations for the above scenarios may allow icentifica-tion of the critical case ano detailed calculations to cefine the tolerable leak rate may only have to be performed for the governing conoitions. The results will be plant specific, cepenoing on the plant site, ECCS design, and fuel cuty, so that development of a single quantitative criterion may not be possible.
B.
EVALUATION OF ISI 1.
Generic Evaluation of ISI Results from inservice inspections of operating reactors will be reviewed and evaluateo as they relate to the Task Action Plan.
In adoition, the results of industry and NRC experimental and analytical stuaies will De reviewed to evaluate the safety of continued operation of pressurized water reactors.
~
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3 2.
Develop Statistically Based ISI Program 2
Statistically based inservice inspection programs will cepena on an equivalent number of tube failures, calculated from the tolerable leak rate defined by the systems analyses. Therefore, inservice inspection programs will be established for varying numbers of tolerable tune failures, since this parameter will be plant specific. Statistical analyses will also be performed to define the error associated with eday current testing.
These results will be consicered in the oevelopment of criteria cefining an acceptaole ISI plan, to provide adequate assurance that no more than the toleraole number of tuces woulo f ail during an accident.
3.
Evaluation of ISI Methoos A qualitative review of the development of eddy current probes, coils, aad multi-frequency techniques to improve eday current testing accuracy and to better quantify various cefects including dents, cracks in dented regions, and circumferential fatigue cracks will be performco. Results of this review will indicate promising areas for further research in improving eddy current testing accuracy.
C.
EVALUATION OF STEAM GENERATOR TUBE INTGEGRITY 1.
Mechanical Integrity The mechanical integrity of steam generator tubes under normal operating and postulated accident conditions (LOCA, SSE, and MSLB) will be reviewed ano evaluated. Specifically, experi-mentally obtainea burst and collapse pressures will De reviewea and used to develop improved tube plugging criteria.
2.
Material Integrity Corrosion and other materials related tube degracation phenomena will be reviewec as they relate to tube plugging criteria. Particular emphasis will be placed on the tuce centing phenomenen and the evaluation of field experience ano laDoratory data to define plugging criteria for centea tubes.
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. 3.
Steam Generator Design and Operations Steam generator and secondary system design and operating 2
procedureswill be evaluated as they relate to mechanical and material aspects of steam generator tube integrity. Specific emphasis will be placed on identifying improved aesign and operating procedures, to reduce the potential for centing and other corrosion related pheonmena in Westinghouse and Combustion Engineering steam generators anc fatigue cracking and erosion-corrosion in Babcock and Wilcox steam generators.
Specific areas.of secondary system operation, materials selec-tion, and mechanical design to be addressed include seconaary water chemistry, main condenser integrity, potential for con-taminant ingress from cemineralizers, ano potential for metal ion transfer from the main condenser and other secondary system equipment Euch as feeawater heaters.
D.
ISI COST-BENEFIT ANALYSES Because the above studies could potentially result in increased inservice inspection requirements, a cost-benefit analyses is being performed to evaluate the impact of such requirements. The analyses will quantify the impact of inservice inspections for varying tube sample sizes considering parameters like personnel exposure anc cowntime. These costs will De weighea against the potential savings from eliminated unscheduled shutdowns to prmiae a realistic evalua-tion of the impact of new criteria.
E.
RESULTING CRITERIA Evaluation of the interaction between systems analyses, statisti-cal analyses, and tube integrity analyses ano integration of the above stuaies will allow the following criteria to be developea.
1.
Tube Plugging Criteria Using results of the mechanical and material tube integrity studies, tube plugging criteria for thinned, cracked, ano dented tubes will be establishea. These criteria will proviae input for a revision of Regulatory Guide 1.121, " Bases for Plugging of Degraced PWR Steam Generator Tubes." Criteria for plugging thinned or cracked tubes will consicer th? ainimum allowable wall thickness defined by the mechanical tube integrity studies and statistically based allowances for eday current testing error and operational oegracation between inspections. Criteria for plugging dented tubes will include the incubation time for stress corrosion cracking in cented tubes based on the magnitude and rate of strain in the tubes ana the environment.
2.
Inservice Inspection Criteria Results from the evaluations of ISI discussed above will be used to propose a revision of Regulatory Guice 1.63, 1700 088
. " Inservice Inspection of Pressurized Water Reactor Steam 2
Generator Tubes."
3.
Steam Generator Design and Operating Criteria Based on the above tube intgerity studies, licensee and vencor proposed modifications in steam generator design and operating techniques will be reviewed ano evaluated and appropriate design and operating criteria for operating plants and plants not yet licensed will be established.
3.
BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLETION OF TASK The safety issue addressed by this Task Action Plan is applicable 2
to Pressurized Water Reactors.
For PWRs currently licensed for operation, we have concluded that, 12 pending completion of these TAPS, continuea operation coes not I
constitute an undue risk to the health ano safety of the public for the following reasons:
Augmented inservice inspection requirements and preventive tuce plugging criteria have been established to provide assurance that a great majority of degraded tubes will De identified and removed from service before leaks develop.
Primary to seconaary leakage rate limits, and associatea surveillance requirements, have been establishea to provide assurance that the occurrence of abnormal tube degracation curing operation will be cetectea and appropriate corrective action will be taken such that an indiviaual cefect will not De-come unstable under normal operating, transient or accicent conditions.
Observea through-wall cracks at dented locations, i.e. tuDes/
support plate intersections, have been small and stable (no rapid failures) during normal operation.
In adaition, since such cracks are constrained by the support plates, they are not anticipated to become unstable (burst) during postulated accidents.
Continuous feeaback from operating experience and the TAP efforts will be utilizea to update interim criteria anc requirements.
1700 DM
. For plants experiencing severe degradation, the following adoitional interim bases were also considered:
The prooability of the design basis accicent during normal operation is small and the probabilty that the accident would occur during the short period of time between the leak detection and the plant shutcown is even smaller.
A small amount of leakage (e.g., less than the Technical Specification limit) can be tolerated during normal operation without exceeding the offsite dosage limits of 10 CFR Part 20.
Some small amount of leakage can De toleratea during postulatea l2 accidents.
The above-mentioned rationale, which constitutes the basis for continued operation of licensed PWR facilities, also supports con-tinued licensing of new facilities. Further, to the extent practicable, for facilities not yet licensed for operation " state-of-the-art" design improvements and operating procedures which are expected to decrease the potential for or rate of steam generator tube cegradation are required by the staff. The following design and operational factors are considered by the staff in the conduct of its reviews:
Designs to provide improved circulation to eliminate low flow areas, ano to facilitate sluage removal.
Designs to minimize flow induced vibration and cavitation.
Designs to provide increased flow around the tubes at the support plate.
Selection of material for tube support plates with improvec corrosion resistance.
Material selection (chemistry), processing ar.d heat treatment to minimize the susceptibility of tubes to stress corrosion cracking.
Secondary system water chemistry control.
Secondary side material selection (concensers, feeawater heaters, turbine discs anc blades, elbows, etc.), and water cleanup system to minimize erosion ano the resulting sludge and corrosion procuct builoup in the steam generators.
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. 4.
NRR TECHNICAL ORGANIZATIONS INVOLVED A.
Engineering Branch, Division of Operating Reactors, has the primary lead responsibility for the overall review and evalu-2 ation of steam generator tube integrity. This includes opera-tional experiences, tube failure mechanisms and potential repairs, plugging criteria, ISI requirements, tube failure proDanility, leakage rate limits,and secondary coolant system chemistry control.
Manpower Estimates
.5 Manyear FY 1979 1.5 Manyear FY 1960 B.
Accident Analysis Branch, DSS, and the Environmental Evaluation Branch, 00R, have the responsibility for the review and evaluation of the offsite dosage during a postulated MSLB or LOCA with vary-ing amounts of steam generator tube leakage. Working with the Analysis Branch, DSS, they will establish a tolerable leak rate through the steam generator during the postulated accidents, or criteria for establishing such a rate on a plant specific basis, as appropriate.
The Environmental Evaluation Branch is also responsiole for the cost-benefit analyses described in task 20. This work will be performed under a technical assistance program with Battelle Pacific North-west Laboratory.
Manpower Estimates Accident Analysis Branch Environmental Evaluations Branch
.10 Manyear FY 1979
.10 Manyear FY 1979
.15 Manyear FY 1980
.10 Manyear FY 19e0 C.
Reactor Safety Branch, Division of Operating Reactors has the lead responsibility for the review ano evaluation of the effects of secondary to primary steam generator tube leakage on ECCS and core performance curing a LOCA and defining a tolerable leak rate.
Manpower Estimates:
.10 Manyears FY 1979
.10 Manyears FY 1980 1700 091 D.
Analysis Branch, Division of Systems Safety, has the leaa responsibility in developing necessary analytical capabilities and evaluating the response of the primary and seconaary systems to a MSLB with varying numDers of concurrent steam generator tube failures. This analysis will provide an evaluation of core response and primary to secondary leak rates during 2
the postulated MSLB accident which can be used by the Accident Analysis and Environmental Evaluation Branches in their radiological dose calculations.
Manpower Estimates:
.10 Manyears FY 1979
.15 Manyears FY 1960 E.
Mechanical Engineering and Materials Engineering Branches, Division of Systems Safety have responsibility for the review of mecnanical and material tuDe integrity and lead responsibility for the review of new design, material, and operating methods.
The activities involved will include the review and evaluation-of applicant's and vendors proposed improvements on the design and/or operation of the steam generators for items such as secondary coolant chemistry, design modifications to avoid tube corrosion and denting, condenser design to avoid inleakage, ISI requirements, recommendation for revision of Regulatory Guides and Sections 10.3.6 and 10.3.7 of the Standard Review Plan.
Manpower Estimates Mechanical Engineering Branch Materials Engineering Branch
.1L Manyear FY 1979
.10 Manyear FY 1979
.25 Manyear FY 1960
.25 Manyear FY 1980 F.
Containment Systems Branch, DSS, has the responsibility for tne review and evaluation of containment response during a postulated LOCA or MSLB inside containment with concurrent steam generator tube failures. The tolerable leak rate through the steam generator will be defined for the postulated accident condition.
Manpower Estimates
.10 Manyear FY 1979
.10 Manyaar FY 1960 5.
TECHNICAL ASSISTANCE A. Contractor:
Brookhaven National Laboratory (BNL) - 00R Funds Required:
SK 1979; SK FY 1900 1700 092
. This program is needed to obtain technical consultation ano assistance to review information in areas of water chemistry and corrosion analysis, monitored by EB/00R. This program 2
will provice assistance in accomplishing Tasks C2 and C3.
B.
Contractor: Brookhaven National Laboratory (BNL) - DSS 2
Funds Required: 125K FY 1979 The purpose of this program is to develop the necessary analytical capaDilities and evaluate the effects of steam generator tube ruptures concurrent with MSLB. The results of this program will be used to determine a tolerable leak rate during postulated accident condtions. This program will assist in accomplisying Task A2.
C.
Contractor: Sandia Laboratories, - 00R Funds Required: 39K FY 1979; 10K FY 1980 The purpose cf this program is to perform a statistical analysis of steam generator tube failures in operating reactors in order to establish the bases for the sampling plan for inservice inspection. This program will assist in 2
accomplishing Tasks B1 and B2.
D.
Contractor: Battelle Pacific Nortnwest Laboratory Funds Required: 80K FY 1980 The purpose of this program is to perform the cost-benefit analyses described in task 2D.
6.
ASSISTANCE REQUIREMENTS FROM OTHER NRC 0FFICES A.
Office of Nuclear Regulatory Research, Division of Reactor Safety Research, Metallurgy and Materials Branch.
RES has funded, at the request of NRR, a major confirmatory experimental program at Pacific Northwest Laboratory. The activity of this program consists of a series of tests to verify the burst and cyclic strengths of degraded steam generator tubes and the leakage rate data. This program 2
is managed by Metallurgy ana Materials Branch, (Task C1).
RES has also funded, at the request of NRR, a program, to address the factors whicn determine Incenel 600 sus-ceptibility to stress corrosion cracking in primary water. Metallurgical condition, chemistry, temperature, stress and environment will De considerea, (Task C2).
2 1700 093 D
0 MENU B.
Office of Standaros Derelopment, Division of Engineering Standaros, Structures dna Components Standaros Branch.
OSD has funded a confirmatory research program at Battelle Columous Laboratory to evaluate eddy current metnods for inspecting steam, ;enerator tubes as a subcontract to 2
Brookhaven Nationa* Laboratory, (Part of Task B3).
C.
Office of the Management and Program Analysis, Applied 2
Statistic Branch.
Provide assistance to EB/ DOR for statistical assessment of steam generator tube integrity, (Part of Tasks B1 and B2).
D.
ACRS This task is closely related to one of the generic items identified Dy the ACRS and, accordingly, will be coordin-ated with tne committee as the task progresses.
7.
INTERACTIONS WITH OUTSIDE ORGANIZATIONS A.
Licensee (s) of Pressurized Water Operating Reactors All plants experiencing abnomal tube degracation will De closely monitored.
Each licensee with severelv degraced steam generators will suomit an analysis of the con-sequences of tuoe degraaation on tuce integrity ano demonstrate that adequate safety margins exist for continued safe operation.
B.
Vendors 2
The primary interaction with the vendors has been ano continues to be on the investigation program for the resolution of the problems at operating plants, justifica-tion for continueo operation of plants with known tuDe aegradation, and the licensing bases for new plants.
C. EPRI, PWR Owners Group, etc.
Interactions with other organizations such as the Electric Power Research Institue (EPRI) ana the "ad noc" Organizations of PWR owners may also be appropriate regarding the safe operation of steam generators in general and, in particular, the various safety problems associatec with the degradation of steam generators.
The primary purpose for interactions with these organizations 2
is to exchange infoma ion on their experience and research work they are sponsoring.
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8.
POTENTIAL PROBLEMS Except for steam generator replacement, there is no apparent short term resolution of tube denting in affected Westingnouse or CE plants. Programs to resolve tube denting in presently operating plants may bring about a partial solution, by arresting centing 2
through a cleaning program.
However, by establishing plugging criteria for dented tubes, and requiring scheduled inspections varying with the degree of denting observed, ' safety concerns can De minimized to the point where continued operation can be justified.
Unfortunately results of the 2
BNL stress corrosion cracking program sponsored by the Office of Research may not be availaole before the desired task action plan completion dates; therefore, plugging criteria will necessarily De based on preliminary results of the program and largely on operating experience.
The criteria may therefore be somewhat judgemental in nature.
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STANDARD DISTRIBUTION Task A-3, 4, 5 Steam Generator Tube Integrity Central file H. R. Denton Jf Hanauer M. Aycock D. Eisenhut B. Grimes L. Shao G. Lainas E. Adensam J. Hard S. Hosford V. Noonan R. LaGrange F. Alneter J. Strosnider K. Parczewski W. Hazel ton B. Liaw M. Wohl F. Schroeder J. Knight R. Bosnak W. Butler P. Check K. Kniel T. Novak S. Pawlicki Z. Rosztoczy J. Zwolinski F. Odar J. Rajan H. Conrad B. Tarovlin D. Muller W. Kreger R. Houston G. Knighton L. Soffer S. Varga W. Minners F. Hebdon L. Crocker NRC PDR*
Local PDR*
NSIC*
ACRS*
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