ML20041D972
| ML20041D972 | |
| Person / Time | |
|---|---|
| Issue date: | 02/28/1982 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20041D969 | List: |
| References | |
| SECY-82-072, SECY-82-72, NUDOCS 8203090692 | |
| Download: ML20041D972 (10) | |
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s STEAM GENERATOR STATUS REPORT FEBRUARY 1982 U.S. NUCLEAR REGULATORY COMMISSION 8203090692 820218 PDR SECY B2-072 PDR T
I.
Probicm Defintion A.
Summary of Tube Degradation Degradation of steam generators (SG) manufactured by each of the three pressurized water reactor (FWR) vendors has resulted due to a combination of steam generator mechanical design, themal hydraulics, materials selection, fabrication techniques, and secondary system design and operation.
In the early and mid-1970s, Westinghouse (W) S.G. experienced caustic stress corrosion cracking, and W and Combustion Engineering (CE)
+
S.G.s experienced tube thinning (wastage). These modes of degradation were due to difficulties encountered with phosphate secondary water chemis try.
Because of these difficulties, most W and all CE plants converted to an all volatile (AVT) secondary water treatment.
Al though this conversion greatly reduced the occurence of stress corrosion cracking and wastage, other degradation modes including denting (defomation of the S.G. tubes due to corrosion of the carbon steel support plates) began to occur.
Babcock and Wilcox S.G., which have a significantly different design from W or CE and have operated exclusively with AVT water chemistry, had relatively good operating experience in their early years of operation.
Nevertheless, they have experienced numerous tube leaks.
The principal modes of degradation in B&W units have been fatigue crack growth, confined primarily to limited sets of tubes located on the open inspection lane, and more recently erosion-corrosion and primary side intergranular a ttack.
To date, many different foms of steam generator degradation have been identified including:
stress corrosion cracking,2 wastage, intergranular attack, denting, erosion-corrosion, fatigue cracking, pitting, fretting, and support plate degradation. One or more of these forms of degradation have affected at least 40 operating PWRs and have resulted in extensive S.G. inspections, tube plugging, repair, or replacement.
Recently, foreign Westinghouse S.G.s of the same design as McGuire have experienced tube wear associated with flow induced vibration due to a new integral preheater design. ' References 1, 2, and 3 present detailed discussions of domestic S.G. operating experience.
The economic impact of steam generator degradation has been significant.
Approximately 23% of non-refueling outage time has been attributed to steam generator degradation. The cost of such outages in tems of replacement power alone is very high.
However, perhaps the greatest financial costs incurred to date are thosa associated with steam generator replacenent.
Replacement of the Surry Unit 1 and Unit 2 S.G.s cost approximately $200 million, including cost of makeup power.
Replacement of the Turkey Point S.G.s, currently in progress, will cost an estimated $460 million. NRC staff time involved with these activities is estimated at 6000 manhours for Turkey Point (which included time for a hearing) and 3000 manhours for Surry.
Less radical operations also incur significant costs.
Recent tube sleeving operations at San Onofre involved repair of approximately 7000 degraded tubes at a cost of $70 million.
Proposed sleeving of 3000 tubes at R.E. Ginna has an estimated cost of $20 million.
B.
Safety Significance The safety significance of S.G. tube integrity can be divided
- into three categories:
tube failures under nomal operating conditions; tube failures concurrent with postulated accident conditions; and personnel exposure associated with S.G. inservice inspection (ISI), repair, and replacement.
The majority of the S.G. tube failures that have occurred under nomal operating conditions were small stable leaks sometimes requiring plant shutdown, inspection, and corrective actions, but for the most part small enough (e.g., below technical specification leak rate limit) that operations continued until a scheduled shutdown.
However, four significant S.G. tube ruptures have occurred in domestic PWRs since 1975.
These everti occurred on February 26, 1975, at Point Beach Unit 1, Septaber 15, 1976, at Surry Unit 2, October 2,1979, at Prairie Island Unit 1 and on January 25, 1982, at R. E. Ginna.
The first three of these events were evaluated in NUREG-0651, " Evaluation of Steam Generator Tube Rupture Events." The report includes an evaluation of system response, operator action, and radiological consequences during the three events.
The leak rate associated with these events ranged from about 80 gpm to 390 gpm.
The conclusion of the report is that no significant offsite doses or systems inadequacies occurred during the tube rupture events analyzed. However, the potential for more significant consequences was recognized and a number of procedural recommendations were made to correct the deficiencies that were noted.
The present disposition of each of the recommendations is discussed in a recent memo to Commissioner Bradford from W. Dircks (Ref. 4).
The present design basis for assuring that plants are acceptably protected against S.G. tube rupture events is a postulated double-ended rupture of a single S.G. tube.
This assumption is intended to provide a bounding leak rate for a spectrum of rupture geometries in a single tube and a spectrum of smaller leaks in multiple tubes within a single S.G.
The consequences of multiple tube failures, in excess of the design base, have not yet been rigorously studied.
Rapid degradation between inspections of a large number of tubes could create the potential for multiple tube failures in the event of a plant transient or failure of a single tube and the accompanying jet impingement and tube whip could cause failure of additional tubes.
Furthemore, the potential for complicating circumstances involving multiple equipment failures such as the stuck open PORV during the Ginna incident and possible steam bubble fomation in the primary system have not been evaluated. Another concern is ruptures in multiple S.G.s.
In this event, unless the plant can be rapidly depressurized and brought onto Residual Heat Removal, there is the potential to continuously lose emergency core cooling water outside of containment. The above concerns are being addressed as part of the TMI Action Plan.
Item I.C.1 '
in the TMI Action Plan addresses S.G. tube failures coupled with other failures (such as a stuck open safety relief valve in the secondary system), ruptures of multiple tubes, and simultaneous ruptures in multiple S.G.s.
The purpose of this effort is not to expand the plant design basis but to assure that operator emergency procedures provide proper guidance for safely controlling the plant during these types of events.
Although rigorous analyses of many of the scenarios postulated above have not been completed, ISI, leak rate limits, and tube plugging requirements are intended to guard against such occurrences (See Section II).
In summary, the consequences of S.G. tube ruptures under nomal operating conditions have been small; however, such evenM can present a significant challenge to plant operators and safety systems.
8 During postulated accident conditions, such as main steam line break (itSLB), feedwater line break, or LOCA, the S.G. tubes are subject to increased pressure differentials and possible pressure. waves (e.g.,
subcooled decompression phenomena) and vibrational loadings.
These loads increase the potential for failure of degraded S.G. tubes which could exacerbate the accident sequence.
In the event of MSLB, failed S.G. tubes would provide a leakage path from the primary to secondary system and several potential leak paths for radioactivity to the environment would then exist.
In the event of a LOCA, the core reflood rate could be retarded by steam binding.
This phenomenon is associated with a cold leg break, in which reflood of the core requires displacing steam generated in the core through the hot leg, the affected steam generator, and out of the cold leg break.
S.G. tube failures would create a secondary to primary leak path which aggravates the steam binding effect and could lead to ineffective reflooding of the core. Analytical and experimental evaluations of this phenomenon are contained in References 4 and 5.
Large MSLBs and LOCAs are considered extremely low probability events, but are postulated as bounding conditions.
More realistic events might include small and intemediate size NSLBs or LOCAs.
Although these postulated accidents pose a less severe challenge to S.G. tube integrity, tube rupture (s) leading to or following such events could have serious consequences.
This is particularly true if fuel damage has occurred as in the case of Three Mile Island.
The final area of concern is the radiatien exposure of personnel involved in S.G. inspection, repair, and replacement.
Reference 3 presents a summary of data on S.G. related personnel exposure for selected plants from 1974 to 1980.
In recent years, as much as 25% of some plants annual occupational exposure has resulted from routine S.G.
inspection and maintenance and as high as 60% for S.G. replacement.
Recent tube sleeving operations at San Onofre incurred 3500 man rem exposure and similar operations are planned for other plants.
II.
Regulatory Approach The NRC approach to assuring S.G. tube integrity under all operating conditions is based on inservice inspection (ISI), primary to secondary leakage rate limits, and preventive tube plugging requirements.
Guidance for perfoming ISI is provided in R.G.1.83, " Inservice Inspection of S.G. Tubes," and plant technical specifications include requirements for ISI.
Typical plant specifications require periodic inspections of 3% of the S.G. tubes in the plant and augmented ISI in the event tube degradation is detected.
Required frequency of inspection is generally flexible enough to allow inspections to be perfomed concurrent with refueling outages.
Certain incidents such as tube leakage require unscheduled ISIS.
Furthemore, many plants with extensive degradation problems have licensing amendments imposing higher frequency and larger size inspections.
The ISI requirements were developed largely through a combination of engineering judgement and operating experience. More rigorous statistically based ISI programs have been developed as part of Unresolved Safety Issues A-3, A-4, and A-5 (see Section V). The purpose of the required ISIS is to detemine if tube degradation is occurring in the S.G., assess the rate of tube degradation based on results of successive inspections, and identify those tubes requiring plugging or repair.
i Primary to secondary leak rate limits are an extremely important requirement for ensuring safe S.G. operation.
Some foms of tube degradation have been observed to degrade tubes beyond the prescribed plugging limit
, during the interval between inspections.
Technical Specification primary to secondary leak rate limits requiring shutdown, ISI, and corrective actions provide protection against unacceptable levels of degradation between inspections.
Many serious conditions of tube degradation have been detected by monitoring of primary to secondary leakage and subsequent inspection.
Primary to secondary leak rate limits exist in each plant's technical specifications.
The bases for these limits are twofold.
First, the leak rate limit ensures that the calculated dosage contribution from tube leakage will be limited to a small fraction of the allowable limits in the event of a S.G. tube rupture or MSLB.
Second, the leak rate limit is intended to correspond to a defect size that would not be expected to result in tube rupture under nomal or postulated accident conditions. :
Finally, degradation limits for tube plugging exist in the plant Technical Specifications.
Criteria for establishing the tube plugging limits are presented in R.G.1.121, " Basis for Plugging Degraded Pressurized Water Reactor Steam Generator Tubes." These criteria require that the plugging limit include margins for eddy current testing error and continued depMation between inspections. Thus, it is important to have a good estimate of the rate of degradation based on successive ISI results and an understanding of the degradation phenomena.
The primary focus of the current NRC philosophy is directed at maintaining primary system integrity.
This is accomplished primarily through the requirements described above for ISI, leak rate monitoring, and tube plugging.
In a sense, it is directed at treating the symptoms and not the cause of S.G. degradation, which lies primarily in secondary system design and operations.
This philosophy has been debated extensively, but the current position regards eliminating the problem at its source as an industry responsibility.
III. Current Corrective Actions An effective solution to S.G. tube degradation problems would require major changes in S.G. mechanical design, themal-hydraulics, materials selection, fabrication techniques, and changes in the secondary system design and operation.
Elimination of S.G. degradation requires a systems approach integrating all of these considerations. There are no simple corrective actions.
This is particularly true for those plants which have significant operating time and have experienced S.G. degradation.
Design changes in operating S.G.s that would be necessary to eliminate degradation problems are virtually impossible.
For example, tube to tubesheet crevices already contaminated with corrosive environments are virtually impossible to clean, carbon steel support plates cannot be replaced with more corrosion resistant materials, and residual fabrication stresses cannot be removed. Thus, corrective actions may prolong S.G.
life, but tube degradation is expected to continue in operating plants.
Once the secondary systen is contaminated by an aggressive environment it is difficult to reverse the adverse affects.
For example, caustic stress corrosion cracking and wastage, due to residual phosphate water chenistry conditions, still continue in some plants long after conversion to AVT water chemistry.
i Several corrective actions, however, have been proposed and l
i are in use.
These fixes include.such actions as tube sleeving, sludge lancing, soaking and flushing, reduced operating temperatures to slow corrosion, boric acid injection to arrest denting, support plate modifications to retard denting, S.G. replacement, and improvements in secondary system design and operation.
Secondary system improvements include prompt correction of condenser in-leakage, condenser retubing, removal of copper based alloys from the secondary system, and addition of demineralizing systems. An industry constituted secondary water chemistry guidelines committee, under chaimanship of EPRI, is developing generic chemistry limits and operating guidelines.
NRR has been in contact With this committee for the past year and will review a copy of the draft reports prior to issue.
Chemical cleaning has also been proposed but has not been implemented due to uncertainties regarding its longer-tem affect on S.G. integrity.
?ndustry efforts are currently underway to eliminate these uncertainties and chemical cleaning may become a viable optien in the near future.
These fixes have met with varying degrees of success, but none of then is a panacea.
Furthemare, short tem solutions to one problem may create other problems.
Conversion from phosphate to AVT water chemistry, which minimized wastage and stress corrosion cracking but was followed by denting, is a case in point.
Finally it should be noted that the majority of the plants under review for operating licenses have S.G.s of similar design to those currently in operation, so that the potential for S.G. tube degradation exists in these plants as well.
IV.
NRC, Industry, and Foreign Research tnd Development Activities NRC's steam generator research program addresses improved eddy current inspection techniques for steam generator tubing, stress corrosion cracking of steam generator tubing and evaluation of tube integrity.
The objective of the eddy-current program is to upgrade and improve eddy-current inspection probes, techniques and associated instrumentation for inservice inspection of steam generator tubing to improve the ability to identify and characterize tube defects.
Specific objectives include improving defect detection and characterization as affected by tube diameter and thickness variations, tube denting, probe wobble, tubesheet and tube support interference, and defect location and type.
The stress corrosion cracking program is developing data and models which will be used to predict the stress corrosion cracking initiation and service life of Inconel 600 steam generator tubing.
The testing program includes variables which influence stress corrosion cracking such as temp,erature, stress, strain and strain rate, metallurgical structures and processing, and ingredients in the primary and secondary cool ant.
A steam genentor, with service induced degradation will be used for the validation of the accuracy and confidence limits of nondestructive inspection instrumentation and techniques; burst and collapse tests on field degraded tubes to validate tube integrity models; and for developing i
data for validation of previously developed stress corrosion cracking l
predictive models, chemical cleaning and decontamination, dose-rate reduction and secondary side characterization.
In addition, statistically based sampling models for inservice inspection programs will be confimed and/or improved utilizing the first ever confimed data base.
L
, There are many ongoing programs addressing S.G. issues at EPRI, mst of which are sponsored by the S.G. Owner's Group, and the rest by EPRI itself.
The programs address the following areas:
(1) chemistry and corrosion, (2) materials selection and testing, (3) themal hydraulic and structural testing and analysis, and (4) nondestructive examination (NDE).
Efforts in the chemistry and corrosion area are directed at examining the causes of corrosion related degradation such as denting, intergranular attack, and stress corrosion cracking, and identifying potential fixes such as alternative secondary water chemistry treatments. Materials selection and testing efforts are directed at characterizing and evaluating the suitability of alternative tubing and S.G. materials. This includes consideration of new heat treatments for tubing and compatability of S.G. tubing with structural materials.
Testing and analysis in themal hydraulics and structures is directed at secondary side S.G. design and perfomance and their effect on S.G. tube integrity. The EPRI nondestructive examination programs focus on development of improved inspection techniques.
These techniques include multiple frequency /multiparameter eddy current testing, automatic eddy current signal analysis, profilometry for quantifying dent configuration and strain levels in dented tubes, and methods for evaluating the condition of the tube support plates.
In addition, EPRI has established the NDE Center in Charlotte, NC, dedicated to providing good NDE techniques, and effectively transferring research and development results to the industry.
Research and development activities underway on steam generators outside the USA are being funded at high levels in several countries.
The Japanese are conducting a very large program with emphasis on themal/
hydraulics, and also on water chemistry and tube testing.
To date, we have received little infomation on the progress or results of their p rog rams.
The French have work underway on eddy current NDE, crevice chemistry, and decontamination. There is work underway in Sweden on water chemistry.
The Gemans have work undemay in eddy current NDE, and at KWU on primary side decontamination and secondary side cleaning; however, Geman steam generators are tubed with Incolloy 800 so much of their research is less relevant to ours.
Finally, the Italians have undemay a large program which will allow them to make new designs to avoid current and possible future problems.
V.
Long Tem Approach A.
Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity In 1978, the NRC established Unresolved Safety Issues A-3, A-4, and A-5 (USI) regarding degradation in W, CE, and B&W steam generators, respectively.
A draft report, NUREG-0844, presenting the proposed NRC
' staff resolution of these generic safety issues has been prepared and is currently being reviewed by NRR management prior to transmittal to the Committee for Review of Generic Requirements and the Commission and publication for public comment.
The report integrates technical studies in the areas of systems analyses, inservice inspection (ISI), and tube integrity to establish improved criteria for ensuring adequate tube integrity and safe steam generator operation under all conditions.
O 9 In the systms analyses portion of the report, the consequences of steam generator tube failures during nomal operation and postulated loss-of-coolant and main steam line break accidents are evaluated. The evaluation considers predicted fuel behavior, emergency core cooling system perfomance, radiological consequences, and containment response.
The results of the systems analyses lead to proposed criteria for establishing a tolerable level of steam generator leakage during postulated accidents.
ISI techniques are then evaluated, and statistically based ISI programs presented which, if implemented, would provide additional assurance that no more than the tolerable level of tube leakage, defined by the systems analyses, would occur during nomal or postulated accident conditions.
In the tube integrity portion of the report, the behavior of degraded tubes during nomal and postulated accident conditions and tube plugging criteria are evaluated.
Proposed changes in operating procedures and design changes to minimize tube degradation are also identified.
Implementation of the proposed requirements and criteria developed in the program for resolution of the USI are not expected to totally eliminate S.G. degradation. The intent of the proposed requirements is to establish a logical approach to evaluating steam generator tube integrity and ensuring safe steam generator operation. The draft NUREG-0844 recommends criteria and requirements that can be used to evaluate current and future degradation programs in steam generators.
The establishment of maximum allowable steam generator tube leak rates during postulated accident conditions and associated tolerable number of defective tubes is a major contribution to the evaluation of steam generator tube degradation problems.
It provides objective criteria against which steam generator tube integrity can be evaluated.
Similarly, the development of statistical ISI programs provides a rational, scientific basis that can be used to establish and evaluate ISI requirements that will ensure the above criteria are satisfied. Results from NRC S.G. research programs are expected to lay the experimental basis for many of these criteria.
In keeping with the NRC's current and past philosophy on this issue, the proposed regulatory requirements developed in the draft report focus on ISI programs and techniques and tube plugging criteria.
The primary responsibility for attacking the problem at its source and eliminating S.G. degradation is the industry's.
However, several of the requirements proposed in NUREG-0844 are intended to promote industry efforts in this area.
For example, one requirement is to ensure that all operating plants have implemented an approved secondary water chemistry monitoring and control program. This is a requirement in the most recent version of the NRR standard review plan for licensing of new plants.
In addition, this type of program has been implemented at some but not all operating plants.
Under this requirement, it is the industry's responsibility to establish specific water chemistry limits and effective monitoring techniques. This will ensure that each utility at least considers the importance of secondary system water chemistry and puts in the effort to develop a comprehensive water chemistry program.
Similarly, ISI requirenents for condensers are proposed. These requirements will hopefully reduce the frequency of condenser in-leakage and encourage anr
.. utilities to improve condenser perfomance.
Use of noncopper based alloys when retubing condensers and feedwater heaters is also a requirement.
Additional requirenents are proposed for plants in the preoperating license stage and many recommendations for operating and future plants are made. The intent of the proposed requirements as stated in the report is to leave primary responsibility for correcting the S.G. problem in the hands of the industry, to allow the industry flexibility in addressing the issue, but at the same time, to strongly encourage proper industry actions.
B.
Comprehensive flRC/ Industry Program The preceeding review has attempted to summarize the status of the S.G. issue at this time. As indicated, the NRC has many ongoing efforts to address this multifaceted problem.
However, to date, joint flRC and industry cooperative efforts on this issue have not been extensive.
This is due largely to the different focuses on the issue.
flRC is primarily concerned with requiring adequate ISI and corrective actions to ensure primary system integrity, while the industry has been concerned with developing fixes to prolong S.G. service life and reliability.
ftRC and industry efforts have been primarily complementary in nature.
However, to the extent that reliability implies safety and vice-versa the flRC and industry efforts are synonomous.
Therefore, the staff is pursuing the development of a joint f1RC and industry program to address both near-tem and long-tem actions required for continued safe operation of steam generators and ultimate resolution of the S.G. degradation problem.
The intent is to evaluate the degree to which the flRC can expand its role in prevention of tube degradation and work with the industry to solve this problem.
Efforts to detemine the feasibility of this type of cooperative program have been initiated and proposals for a joint flRC and indust ~ry program will be presented in a later document
~
e REFERENCES 1.
Eisenhut, Liaw, Strosnider, " Summary of Operating Experience with Recirculating Steam Generators," NUREG-0523, January 1979.
2.
Liaw, Strosnider, " Summary of Tube Integrity Operating Experience with Once-Through Steam Generator," NUREG-0571, March 1980.
3.
SECY-81-664, "Information Report - Steam Generator Tube Experience,"
from W. J. Dircks to the Canmissioners, November 24, 1981.
4.
Memorandum for Commissioner Bradford fran W. J. Dircks, Status of Recanmendations Made in NUREG-0651, " Evaluation of Steam Generator Tube Rupture Events," to be transmitted.
5.
EG&G Idaho, Inc. Report TREE-NUREG-1213 (NUREG/CR-0175), " Investigation of the Influence of Simulated Steam Generator Tube Ruptures During Loss-of-Coolant Experiments in Semiscale MOD-1 Systems," May 1978.
6.
EG&G Idaho, Inc. Report CAAP-TR-032, " Steam Generator Tube Rupture -
Effects on a LOCA," November 1978.
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