ML20154J312

Trip Rept of 980607-10 Visit to Paris,France to Exchange Ideas & Data on Experience with Fatigue in LWR Piping Caused by Mixing & Stratification

ML20154J312
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Issue date:	09/29/1998
From:	Hartzman M, Lund A NRC (Affiliation Not Assigned)
To:	Hackett E, Wessman R NRC (Affiliation Not Assigned), NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
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ML20154H218	List: ML20154H213 ML20154J312
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NUDOCS 9810150213
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• a atuq\ UNITED STATES j u 2

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30006 4 001

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em, September 29,1998 l

MEMORANDUM TO: Richard H. Wessman, Chief Mechanical Engineering Branch i

Division of Engineering, NRR Edwin M. Hackett, Acting Chief Electrical, Materials and Mechanical Engineering Branch Division of Engineering Technology, RES FROM; Mark Hartzman, Senior Mechanical Engineer g Mechanice' Engineering Branch tWg Division of Engineering, NRR A. Loulso Lund, Materials Engineer 9,3 8

Electrical, Materials and Mechanical neering Branch l DMslon of Engineering Technology, RES 3UBJECT: TRIP REPORT, MEETING OF SPECIALISTS ON EXPERIENCE WITH THERMAL FATIGUE IN LWR PlPING CAUSED BY MIXING AND STRATIFICATION I

MEETING DATE: June 7-10,1998 LOCATION: Paris, France.

NRC ATTENDEES: Louise Lund, RES Mark Hartzman, NRR l

Conclusions e The purpose of the meeting was to conduct an intemational exchange of ideas, experience and data on thermal stratification leading to piping failures in nuclear plants 1 worldwide in recent years. The meeting participants were primarily regulators, NSSS vendors and utilities.

L. Lund presented a paper titled "USNRC Regulatory Perspective on Unanticipated l Thermal Fatigue in LWR Piping", co-authored by L Lund and M. Hartzman (copy

! attached).

l e Six attendees were invited to participate in a panel discussion held after the last session with each member making a summary statement. M. Hartzman stated that, after listening to the papers presented at the meeting, the main conclusion stated in the NRC l paper remained valid, that is, the basic thermohydraulic phenomenon which caused the I cracks in Farley and Tihange is still not fully understood. The other conclusion he stated '

is that monitoring of pressure or temperature or leakage remains the most effective method to prevent thermal cycle fatigue failure when the exact causes are not known.

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• The final papers will be published in a proceedings volume at the end of 1998.

I Introduction The purpose of the meeting was to exchange ideas and data on a class of problems which have appeared in nuclear plants worldwide in recent years. These problems are i caused by thermalloading conditions which were not considered in the design of nuclear piping systems because they were unknown at the time the plants were designed. These loading conditions result from the flow of hot and cold fluids within a pipeline, termed stratified conditions.

The meeting was attended by representatives of regulatory bodies, NSSS vendors and utilities. The American participants were the NRC representatives, Louise Lund and Mark Hartzman, and a representative of StructuralIntegrity Associates. The only American NSSS vendor was Westinghouse, represented through its Belgian subsidiary. 1 The other participants were mainly from France. There were also representatives from  !

Belgium, Finland, Netherlands, Germany, Czech Republic, Japan, Sweden, United l Kingdom, Switzerland, Hungary and South Korea.

Background l j

Stratified conditions within a pipe may exist under steady state conditions or under transient conditions. Under steady state conditions, stratification causes thermal bending stresses in the pipe. Under transient conditions, often occurring inadvertently as a result -

of isolation valve leakage, the stratification becomes cyclic in nature, causing local I alternating thermal stresses, possibly leading to fatigue cracking.

Cracking of piping resulting from thermal stratification Decame evident in 1984. Thermal fatigue cracking due to inadvertent leakage in unisolable lines connected to the reactor i coolant system (RCL), caused by transient thermal stratified conditions, became evident in 1987 and 1988 with the events in Farley 2, Tihange 1 and Cenkal 1, and were reported in Bulletin 88-08 and its supplements. Steady state bending stresses due to thermal stratification were reported in Bulletin 88-11. Other similar incidents have also been reported since then in French, German, and Finnish plants. The latest incidents occurred in Oconee 2 and Civaux 1, a French plant.

Highlights e Five sessions were held on the following topics:

1. Operating Experience

2. Thermal Hydraulic Phenomena

3. Response of Materials and Structures

4. Monitoring Aspects

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! 5. Inspection Programs, Mitigation and Prevention, Safety implication l e A number of papers were presented at each of these sessions, all in English. Not all l papers were presented in final form, and some will be revised before publication. The papers and related overheads will be published in a proceedings volume in late 1998.

e L. Lund presented a paper titled "USNRC Regulatory Perspective on Unanticipated i Thermal Fatigue in LWR Piping", co-authored by L Lund and M. Hartzman. The basic conclusion was that the thermohydraulic phenomena which caused the Farley 2 and Tihange 1 events were still unknown, that monitoring of susceptible piping (except in special circumstances) was the most effective method to provide assurance that these events would not recur, and that analytical methods used in lieu of monitoring must reflect these events.

• A Japanese paper presented the results of a full scale simulation of the Farley 2 event.

l. Temperature measurements were taken at various locations along the axiallength of the piping. Fatigue calculations of the induced thermal stresses indicate that the fatigue failure would occur at the check valve weld. This does not reflect the failure at Farley 2, l where the failure occurred at the elbow weld. This indicates that the test did not really adequately simulate the Farley 2 event. From the paper, they found the minimum level of valve leakage that would lead to thermal stratification by using the experimental simulation of the Farley 2 event, and then assessed the maximum levels of leakage past the valves for a sampling of valves in their operating plants. They found that the leakage

through the valves was below the minimum level thought to cause thermal stratification l from their testing, so this gave them assurance that thermal stratification was not

' deemed a problem for the pipelines assessed. They continue to monitor leakage from valves in lines thought to be susceptible to thermal stratification, to provide an early indication of possible thermal stratification.

• A paper was presented on the Genkai 1 event, described in Supplement 3 of Bulletin 88-

08. It described the results of the examination of the cracked pipe after the event. It also i described the results of a test which simulated the event, which verified the explanation l shown in the Supplement. A separate test also revealed that there may have been high l residual stresses at the weld and heat affected zone.

• A paper was presented on the Tihange 1 event, reported in Supplement 1 of Bulletin 88-

08. It reported the results of the metallographic examination, which verified that the cause of the failure was the same as that at Farley 2. No fatigue stress studies were made, since temperature data was not measured. The utility implemented pressure monitoring as a preventive measure. I e The Westinghouse representative from Belgium presented a paper titled " Isolation Valve Leakage: An Overview of US Methods for Evaluation and Mitigation". The paper was essentially a presentation of the EPRl/ Westinghouse program TASCS. In the discussion L following the presentation, the NRC made clear that this methodology had not been l accepted for implementation in the US.

1 -e The Germans seem to use considerable thermal monitoring and fatigue measuring

. techniques, as well as advanced numerical techniques.

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4 o Residual welding stresses played a significant role in a number of reported failures.

l e No further information was provided about the event at Civaux 1 on May 12,1998 e L Lund provided copies of Information Notice 97-46:"Unisolable Crack in High-Pressure injection Piping," to members of the audience who expressed an interest in the Oconee Unit 2 event .

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Technical Tour l

! L. Lund also participated in an optional technical tour of the Framatome factory and CETIC, the

!- Maintenance Preparation and Qualification Center for PWRs, at Chalon St. Marcel.

! The Framatome facility is the manufacturing site for heavy components, such as reactor vessels, l steam generators, and pressurizers. The site also houses their scientific and technical support personnel in the engineering department and in the Technical Center, which contains j laboratories for developing and qualifying welding and NDE techniques. 1

(' The CETIC facility was created by EDF and Framatome. It is a training facility that provides full-scale mock-ups supplemented with actual components, to provide a realistic training experience ,

l for nuclear wc kers. This facility is used primarily for validation of maintenance procedures, )

I' qualification of prototype tools, and re-qualification of used tools. The facility is licensed to l handle "slightly" contaminated tools, primarily to support the re-qualification of used tools. The l facility is open to any nuclear power industry manufacturer or electric utility worldwide.

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USNRC Regulatory Perspective on Unanticipated Thermal Fatigue in LWR Piping A. L Lund and M. Hartzman U. S. Nuclear Regulatory Commission l

l Washington, DC, USA i

l' Abstract The regulatory responso of the United States Nuclear Regulatory Commission (USNRC) to t

unanticipated thermal fatigue in light water reactor (LWR) primary piping systems is based on a i

long history of such events in both US and foreign commercial nuclear plants. This history includes operational conditions leading to surge line stratification, and cracks in feedwater nozzles, high pressure safety injection lines, and residual heat removal lines. US nuclear power plant experience with themial fatigue has prompted two studies by the USNRC Office for l Analysis and Evaluation of Operational Data (AEOD). In addition, the USNRC has provided l generic communications to its licensees addressing problems arising from thermal stresses and l stratification. Thermal fatigue in general has also been examined by the NRC as a generic j safety issue. All these evaluations have alerted the staff to concems over the broader l implications of unanticipated thermal fatigue in piping, including potential problems with plant conditions not being consistent with licensing basis design and inspection commitments. Of particular concem are those problems arising from unanticipated thermal fatigue in unisolable piping connected to the reactor coolant system. Current USNRC acceptance criteria of licensee actions are based on the USNRC's mission to preserve the public health and safety.

! Introduction l

The regulatory response of the USNRC to unanticipated thermal fatigue in light water reactor (LWR) primary piping systems is based on a long history of such events in both American and foreign commercial nuclear plants. This history includes operational conditions leading to surge line stratification, and cracks in feedwater nozzles, high pressure safety injection lines, and i

residual heat removal lines. US nuclear power plant ~ experience with thermal fatigue has prompted two studies by the USNRC Office for Analysis and Evaluation of Operational Data (AEOD): Review of Thermal Stratification Operating Experience (Reference 1) and Primary

, System Leaks (current ongoing study). In addition, the USNRC has provided generic i communications to its licensees addressing problems arising from thermal stresses and stratification (Bulletin 79-13; Generic Letter 85-20; Bulletins 88-08 with Supplements 1,2, and 3;  ;

Bulletin 88-11; NUREG/CR-6456). Thermal fatigue in general has also been examined by the USNRC as a generic safety issue, most recently for (Generic Safety Issue) GSI-190,

• Fatigue i Evaluation of Metal Components for 60-year Plant Lifen. All these evaluations have alerted the '

staff to concems over the broader implications of unanticipated thermal fatigue in piping, l including potential problems with plant conditions not consistent with licensing basis design and inspection commitments. A complete list of relevant USNRC publications is shown at the end of this paper.

, Experience has shown that under certain circumstances, thermal fatigue caused by l unanticipated flow-induced thermal stratification can lead to through-wall cracking in pipes. Of F

particular concem are those segments of piping attached to the reactor coolant system (RCS) which cannot be isolated once a through-wall crack develops.

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[ in an effort to alert and provide guidance to US commercial nuclear power plant licensees on this subject, the USNRC issued Bulletin 88-08 in June 1988, in response to an event at Farley Unit 2.

i Supplements 1 and 2 were also issued in June 1988 and August 1988, respectively, in response i

to a similar event at Tihange Unit 1, Belgium. Supplement 3 was issued in April 1989, in response to an event at Genkai Unit 1, Japan. This supplement addresses the conditions for a

particular case of potential out-leakage from the RCS, which appears to have been a unique occurrence. The bulletin states that thermal fatigue and through-wall cracking can occur in a

unisolable piping connected to the RCS, as a result of cold in-leakage through an isolating block valve, where the upstream pressure is higher than RCS pressure. The bulletin further suggests that because closed valves often leak, a possibly unanalyzed design condition and a potential

! safety issue may exist for those reactors that are susceptible to these conditions, since subjecting flawed piping to excessive stresses induced by a seismic event, waterhammer, or i

some other transient could conceivably result in total rupture of the pipe.

2 As a result, Bulletin 88-08 requested that licensees plan and implement a program to provide  !

l continuing assur6 < a for the life of the plant that the unisolable sections of piping connected to i l the RCS will not be subjected to the type of cyclic thermal loading associated with leaking  ;

, isolation valves, described in the Bulletin. Several options for providing this assurance were  ;

i listed in Action 3 of the Bulletin. Similar incidents have since also occurred in foreign plants, ,

i most notably in France at Dampierre Unit 2 in 1992 and Dampierre Unit 1 in 1996. A brief history '

i of the US commercial nuclear plant experience with thermal stratification is given in the next i section.

j US Commercial Nuclear Experience Feedwater Nozzle Cracks

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. Thermal stratification was first ideritified in 1979 as the root cause for feedwater nozzle cracks, j which occurred at a number of pressurized-water reactor (PWR) plants. As a result, USNRC IE Bulletin 79-13 was issued by the USNRC describing operating events involving cracking in i feedwater system piping, including cracks in the feedwater elbows adjacent to welds in elbows of j steam generator nozzles at 15 Westinghouse and Combustion Engineering-designed plants.

2 investigations into the root cause of the cracking suggested thermal stratified flow conditions in s the feedwater pipe weld region during zero and low power operations. As a result of these findings, the bulletin requested all PWR licensees to: (1) perform radiographic and ultrasonic examination of all feedwater nozzle-to-pipe welds and of adjacent pipe and nozzle areas, (2) repair or replace the cracked piping identified from the examination; (3) perform another examination at the next refueling outage, or (4) change operational procedures such that the feedwater level within the steam generator is maintained essentially constant and no intermittent cold auxiliary feedwater injection is utilized during start-up, hot standby, or cold shutdown operations.

Farley Unit 2 No new industry-wide thermal stratification concem arose until an unisolable leak event from the primary system at Farley Unit 2 in late 1987. The leak was located in the RCS loop B cold-leg safety injection line between a swing check valve and the RCS loop. A through-wall crack was identified at a weld connecting the elbow closest to the RCS and a horizontal spool on the six-2

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inch safety injechon line. Failure analysis on the defective section of piping, followed by temperature measurements and vibration measurements of the affected line, revealed that the crack was caused by fatigue due to cyclic thermal stresses attributed to cyclic thermal stratification. The maximum temperature difference was measured to be 211F. This temperature difference was caused by a small leak of less than one gallon per minute flowing through a closed bypass valve of the cold charging water line. Metallurgical evaluations confirmed that cracking was initiated on the inside diameter (ID) surface and progressed radially outward to the outside diameter. Because the cracking appeared at severallocations over a broader region of the ID surface, rather than from a single isolated initiation site, the licensee concluded that the cracks were caused by high cycle fatigue.

Trojan Nuclear Plant in April 1988, the licensee M the Trojan Nuclear Plant discovered that a pressurizer surge line had been displaced, and was in contact with one of the pipe whip restraints. Shifting of the line had been observed since 1982, when the licensee began to monitor the line, and the response had been to adjust shims and gap sizes on the basis of analysis of various postulated conditions.

The licensee performed a piping stress analysis, which indicated that the surge line under stratified flow conditions would deflect downward, contact pipe whip restraints, and undergo plastic deformation which would result in the cold set of the pipe that was observed in the surge line. As corrective actions, the licensee proposed to perform inspections and nondestructive examination of the line, conduct piping integrity evaluation, and initiate a pipe monitoring program to measure the actual temperature distribution and line movements.

Oconee Unit 2 in April 1997, Oconee Unit 2 was shut down resulting from unidentified reactor RCS leakage exceeding the technical specification limit of 1 gpm. An unisolable leak was found in the high-pressure injection line/make up (HPIMU) line, from a through-wall crack in the weld connecting the HPIMU pipe and the safe-end of the reactor coolant loop nozzle. Preliminary analysis indicated that the degradation was caused by high-cycle fatigue due to a combination of thermal cycling and flow induced vibration. A gap was found in the contact area between the thermal sleeve aid the safe end, indicating that the thermal sleeve may not have been securely attached. .

The theimal sleeve was cracked, with parts missing. The licensee hypothesized that this was j the resu4 of attemate heating and cooling of the weld by intermittent mixing of the hot reactor coolant leaking through the gap in the contact area between the loose thermal sleeve and the safe-end, and the cooler normal makeup water flowing through the HPIMU line. However, other phenomena may have also been involved, and the failure in this line is still under investigation by the licensee. A summary of this event is given in NRC Information Notice 97-46, dated July 9, 1997. Based on this event, as well as for license renewal issues, the NRC staff recently reexamined the requirements given in Section XI of the ASME Code for inspection of these lines, to address the apparent inconsistency in the Code requirements for surface examination versus volumetric non-destructive examination for this piping. This staff initiative is discussed in greater detail in a later section of this paper. This phenomenon has been identified as a probable cause for similar safe-end cracking at Crystal River Unit 3 and other B&W plants in the early 1980's, and addressed in USNRC Information Notice 82-09 and Generic Letter 85-20.

US Commercial Nuclear Industry Response

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In response to Bulletin 88-08, and other generic USNRC communications, the licensees in the 3

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commercial nuclear power industry established a variety of corrective actions aimed at preventing problems from thermal stratification. Temperature monitoring programs were installed at many plants. Other actions included increased inspections of the piping, valve replacements, operations procedural changes, and changing pipe supports to accommodate thermal stresses.

LWR Owners Groups established programs to investigate cyclic thermal stratification as a generic issue. Additionally, the Electric Power Research Institute (EPRI) developed a program called Thermal Stratification, Cycling and Striping (TASCS) (Reference 1) to provide an analytical methodology for addressing unanticipated thermal fatigue problems in piping systems. Although  ;

the basic thermohydraulic phenomena are still not well understood, this has led to a better ,

understmaing of the operational conditions under which potential fatigue failure from cyclic  !

thermal stratification can occur, as no new cracking attributable to thermal stratification had been observed in the United States until the Oconee 2 incident in 1997.

USNRC Sta# Response l The regulatory basis for Bulletin 88-08 is General Design Criterion (GDC) 14 of 10CFR50 (Title 10 of the Code of Federal Regulations, Part 50), Appendix A, which requires that the reactor coolant pressure boundary be designed so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. The Actions required of licensees by the Bulletin, and the USNRC staff review and acceptance of the responses, a.re intended to provide the assurance that GDC 14 is in effect satisfied for the life of the plant. The objectives of the Bulletin were: (a) to advice licensees of a potential safety issue due to an unanalyzed design condition in unisolable lines connected to the RCS, (b) to determine that this problem did not I exist in their plants, or correct it if it did, and (c) to provide continuing assurance that such a l 4

problem will not develop during the life of the plant.

Action 3 of the Bulletin lists three acceptable options for providing the required assurance: (a) redesigning and modifying the piping to withstand the additional thermal cyclic loading due to leakage, in addition to the licensing basis design thermal operational loading, (b) instrumenting the piping to detect leakage and (c) ensure a favorable pressure differential which will prevent the in-leakage or out-leakage in case of a defective valve.

Initially, the response of most licensees was to adopt option (b), by installing thermocouples on lines identified (in Action 1 of the Bulletin) to be susceptible to thermal cycling. Assuming no leaking isolation valves, the objective of the monitoring was to detect unanticipated isolation valve leak initiation, and to take preventive measures as soon as practical to repair or replace the leaking valve or valves.

In recent years, EPRI (Reference 2) and others have developed simplified methods, such as ,

TASCS, to evaluate the thermal effects of valve in-leakage on susceptible piping systems. l Basically, these are tools to assist in addressing Action 3 of the Bulletin, in accordance with I option (a) above. These methods are based on a combination of analytical theory, experimental work and plant measurements. They permit the user, in principle, to determine the effect of the thermal cycling on the fatigue life of the piping, implicitly addressing the USNRC concems.

j The USNRC staff finds such an approach acceptable. It recognizes that shutting down a plant to repair valves imposes a hardship and may have an adverse effect on the safe operation of the plant. It is therefore of interest to the USNRC that licensees be able to determine reliably the '

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l time period from the onset of in-leakage to crack initiation, the most likely location of the crack,

and the time for the crack to propagate to the surface. This provides a reasonable assurance of i safety by determining an acceptable time interval until the repairs can be performed. However, j such an approach must have a reliable, technically defensible basis, and address the j fundamental phenomena observed in the events described in the Bulletin, its supplements, and

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other similar subsequent events. A review of the available literature, both public and proprietary, indicates that the basic thermohydraulic phenomena which caused the cracking in the piping at Fariey and Tihange do not appear to be completely understood, as of the writing of this paper.

in Reference 3, Nakamori and Hanzawa describe experimental work performed in Japan. A test specimen simulating the Farley safety injection line was subjected to the actual operating a ; conditions which existed at the time the leak was detected. Of particular interest in this paper are the axial temperature distributions recorded along the top and the bottom of the pipe with il' i

three in-leakage flow rates: 0.132 gpm (30kg/h), 0.4 gpm (100 kg/h), and 0.88 gpm (200 kg/h).

l (The leak-rate reported in the bulletin was 0.7 gpm (159 kg/h)). At 0.132 gpm the temperatures i! along the top and the bottom were the same as the main pipe, with no cycling. At 0.4 gpm, the

! temperatures along the top and bottom were the same up to the second elbow weld, where the crack at Farley occurred. There was evidence of cycling in the fluid, but little on the inside or outside of the wall at this location. The largest cycling occurred at the check valve weld, located

at 7.8 inside diameters from the main pipe nozzle. The same distribution occurred for 0.88 gpm, t

except the temperatures were lower in the horizontal segment. However, as with 0.4 gpm, i there appeared to be no evidence of temperature cycling in the metal wall at the location of the elbow weld where the crack occurred. The largest temperature cycling in the fluid was again j recorded at the check valve outlet weld. A fatigue assessment at 0.4 gpm was also performed, indicating that fatigue failure would have occurred at this weld. This does not correspond to the i location of the through-wall crack found at Farley.

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! There appears to be a paradox here. At Fadey, the thermal fatigue failure occurred at the weld i between the elbow and the horizontal segment, roughly 5.2 inside diameters from the main pipe nozzle. Yet the reported temperatures in the simulation indicate no thermal stratification or

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cycling at this location. The top to-bottom temperatures are almost identical. The maximum thermal cycling in the fluid was stated to be at the check valve outlet. Then the question arises,

! what caused the fatigue failure at Farley? At Tihange, the cracking occurred in both the end l j welds and the base metal of the first elbow, at a location similar to that of the elbow in Farley.

i Assuming the simulation temperature distribution is applicable to Tihange, it also doesn't explain j the cause of these cracks.

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! The conclusion of the USNRC staff is that, until data consistent with the failures that occurred at i Farley and Tihange, and other reported similar failures, is determined, and the thermohydraulic i phenomena which caused the failures are well understood and reproducible by analytical means, there is no assurance that a given analytical method will provide a reliable assessment under all potential cyclic stratification circumstances, except in special cases where the technique is obviously conservative with respect to known data. The Fadey and Tihange events should be considered as benchmarks for demonstration of method adequacy. The USNRC staff therefore also concludes that temperature or pressure monitoring should be maintained in systems known to be susceptible to the phenomena described in the Bulletin, unless assurance can be provided, based on in-situ measurements, that the leak rates are low and non-cyclic, until a repair or a replacement of the leaking isolation valve can be effected at the next available shut down.

The USNRC staff has recently accepted a few licensees' responses to the bulletin, wh'ch have 5

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^ proposed discontinuing existing temperature monitoring programs. The acceptance was based on the evaluation of proprietary information provided by the licensees, and the following

considerabons

. 1. Monitored temperature traces indicated isolation valve leakage without thermal stratification cycling overlong periods of time (one or more operating cycles).

. 2. The upstream source of the in-leakage was a high temperature source, such as a regenerative heat exchanger. ,

! 3. The nearest check valve was located at least 25 diameters from the RCL nozzle.

j 4. Isolation valve leakage was monitored by other means.

5. Main pipe flow was very low, j 6. Leaking isolation valves had already been replaced, or a commitmewas made to repair

or replace them at the first upcoming refueling outage.

Other USNRC Staff Initiatives

! As a result of the Oconee 2 event (and license renewal issues), the NRC staff recently i

reexamined the requirements given in Section XI of the ASME Code for inservice inspection (ISI) i of high pressure safety injection (HPSI) piping, for both Subsection lWB (for Class 1 piping) and i Subsection IWC (for Class 2 piping).

l The criteria in Subsection IWC require that Class 2 HPSI piping down to nominal pipe size (NPS) 1% receive both a volumetne and a surface examination as part of a facility ISI program. The criteria in Subsection lWB require only that a surface examination be performed for Class 1 l piping less than NPS 4, with one provision excluding piping of NPS 1 and smaller from examinabon. Therefore, for the HPSI system, the inspection criteria for Class 2 piping between i NPS 4 and NPS 1%, inclusive, are.more comprehensive than those for Class 1 piping of the j

same size range. The NRC staff has submitted a formal request to the ASME Code to examine l

) this discrepancy and to pursue its resolution in future Code Editions. '

i To address the NRC staff's immediate concoms regarding the inspection of the Class 1 HPSI

} piping, the staff published in theFederalRegistera proposed rule with the intent of amending the

requirements of 10 CFR 50.55a (see Reference 4). The proposed Rule change would effectively I require licensees to implement volumetric examinations of the Class 1 HPSI piping welds for those sections of piping with size greater than NPS 1 on a schedule consistent with their current ISI program requirements. The public comment period for the proposed 10CFR50.55a rulemaking was recently closed, and the USNRC staff is in the process of responding to the comments received.

In addition, the USNRC staff has also published a proposed Generic Letter in th5ederal Register which is intended to: (1) slert U.S. licensees to the discrepancy noted above; (2) to request information on previous actions taken by U.S. licensees; and (3) to verify the integrity of ,

the subject piping and to request information on the licensee's plans for future inspections. This information is necessary to permit the USNRC staff to assess the quality and level of information available today on HPSI system integrity, and to determine if additional action is required to ensure that the integrity of the system is maintained. The public comment period for the proposed Generic Letter remains open until 29 May 1998.

Conclusion l

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l (1) Based on the above considecations, the USNRC staff concludes that, at present, temperature or pressure monitoring remains the most reliable method to provide the l assurance for the life of the plant, requested by Action 3 of the Bulletin regarding the effects of unanticipated thermal fatigue in unisolable piping connected to the RCS. Under certain conditions, and on case by case basis, the USNRC staff may accept the elimination of an already installed monitoring capability.

(2) The USNRC staff has initiated rulemaking to correct an apparent inconsistency in I.Sl examination requirements for ASME Class 2 and Class 1 for NPS 4 inches nominal and under. It has also requested the incorporation of surface and volumetric requirements for inspection of Class 1 piping 4 NPS and under into the ASME Code Section XI.

Acknowledgments

' l The authors wish to express their appreciation for the technical assistance of Matthew Mitchell,  !

USNRC, in providing insight on other USNRC staff initiatives discussed in this paper. l References

1. Su, N. T. (1990). < Review of Thermal Stratification Operating Experiences U.S. Nuclear Regulatory Commission Report AEOD/S902.

2. EPRI Report TR-103581, < Thermal Stratification, Cycling and Striping (TASCS)> March ib94. (Proprietary) l 3. Nakamori, M., and K. Hanzawa (1995). < Valve Maintenance Guideline to Prevent Thermal Fatigue Damage of the Reactor Pressure Boundary > PVP-Vol. 313-2, Volume 2, l p 95-102. Presented at the 1995 Joint ASME/JSME Pressure Vessels and Piping Conference, Honolulu, Hawaii, July 23 27,1995 l 4. United States Federal Register,82 FR 83892,1997.

NRC Generic Communications Bulletin 79-13 Cracking in Feedwater System Piping

Information Notice 82-09 Cracking in Piping of Makeup Coolant Lines at B&W Plants l Information Notice 84-87 Piping Thermal Deflection Induced by Stratified Flow l Generic Letter 85-20 Resolution of Generic issue 89

High Pressure injection / Makeup i Nozzle Cracking in Babcock and Wilcox Plants Information Notice 88-01 Safety injection Pipe Failure l Bulletin 88-08 Thermal Stresses in Piping Connected to Reactor Coolant l l (incl. Supps.1,2, and 3) Systems t

i information Notice 88-80 Unexpected Piping Movement Attributed to Thermal Stratification 7 l f

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Bulletin 88-11 Pressurizer Surge Line Thermal Stratification Information Notice 97-46 Unisolable Crack in High-Pressure injection Piping NUREG/CR-6456 Review of Industry Efforts to Manage Pressurized Water Reactor Feedwater Nozzle, Piping, and Feedring Cracking and Wall Thinning AEOD S902 Review of Thermal Stratification Operating Experience NUREG/CR-6582 Assessment of Pressurized Water Reactor Primary System Leaks NRC Generic Safety issues (GSI)

GSI-190 Fatigue Evaluation of Metal Components for 60 year Plant Life (ongoing)

GSI-166 Adequacy of Fatigue Life of Metal Components (closed 2/97)

GSI-14 ' PWR Pipe Cracks (closed 10/85)

GSI-156-6.1 Pipe Break Effects on Systems and Components (enhanced prioritization) r 8

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"*g USNRC Regulatory Perspective on Unanticipated l

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/ Thermal Fatigue in LWR Piping  :

A. L. Lund and M. Hartzman  !

U.S.NuclearRegulatory Commission  !

Washington,DC, USA i

Presented at the Specialists Meeting on Experience with Thermal Fatigue in LWR Piping Caused by Mixing and Stratification i

June 7 - 12,1998 1 Paris, France l

I p#'"%, ,

(g.....), US Commercial Nuclear Experience O FeedwaterNozzle Cracks at 15 Westinghouse and Combustion Engineering-designedplants e Cracking in FW system piping adjacent to welds in elbows of steam generatornozzles

• Root cause: thermal stratified flow conditions during 7.ero and low poweroperations

g ,,

4

( .....

) US Commercial Nuclear Experience l

O Farley Unit 2 -- Late 1987 I

e Through-wall crack at a weld connecting an elbow to a horizontal  !

spool on a 6-in. unisolable SI kne

-i e Temperature measurements indicated cyclic thermal stratification (max. top-to-bottom difference 215 F)  ;

i e Small leak (< Igpm) through a closed bypass valve of the cold chargingwaterline I 4

3 i

f %o,,

( i US Commercial Nuclear Experience O TrojanNuclearPlant-- April 1988 i

e Discovered displaced pressurizer surge line in contact with one of the pipe whip restraints

• Piping stress analysis indicated that stratified flow conditions could produce the observed displacement

• Corrective actions proposed: inspections and NDE of the line, as well as pipe temperature and displacement monitoring program l

4

j . ..

/

(g'~"\)

US Commercial Nuclear Experience O Oconee Unit 2 -- April 1997 e Unisolable leak from through-wall crack in a weld connecting HPI/MU line and the safe-end of the reactor coolant loop nozzle e Failure in line still under investigation by licensee i

e Gap found in contact area between thennal sleeve and safe end, and thermal sleeve was cracked with parts missing

• Preliminary analysis indicates that high-cycle fatigue due to thermal cycling and flow induced vibration contributed to failure 5 i i

__-.--__.-__._-_--____<r -

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) US Commercial Nuclear Industry Response O Corrective actions initiated or concluded:

e Implemented temperature monitoring programs L

1.

[ e Increasedinspections ofpiping l

e Replacedleakingisolationvalves l

l

• Implemented operations procedural changes I

. Modified pipe supports to accommodate thermal stratification loads 6

5 p 1

{g%,, i US Commercial Nuclear Industry Response I

O EPRI Program TASCS - Thermal Stratification, Cycling and Striping  ;

• Developed by Westinghouse under EPRI contract in response to I

! Bulletin 88-08

• Identifies piping susceptible to thermal stratification

]

• Provides analytical methodology for determining thermal stratification loads in piping systems

'7 l

l! USNRC StaffResponse i

t O Bulletin 79-13 issued in 1979, requesting all PWR licensees to: .

l e Perform RT and UT examination of all FW nozzle-to-pipe welds .

! and adjacent areas e Repair or replace the cracked piping identified from the examination -

• Perform another examination at next refueling outage

• Change operational procedures to minimize intermittent cold auxiliary FW injection during start-up, hot standby, or cold shutdown conditions 8 i

1 -

1 i e s

.l.

(g'~~%,,f LUSNRC Staff Response 4: ~

Information Notice 97-46 issued July 9,1997, in response to event at .

. Oconee 2 4

! -ASMG i: NRC re-examined inspection requirements in Section XI of B&PV -

! Code: inconsistency ofrequirement for surface examination versus volumetric examination for Class 1 and 2 piping between 1 inch and 4- 1 inches I

t

. .t 9

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(, ,

) USXRC Staff Response O Bulletin 88-08 issued in June 1988 to address unanticipated cyclic thermal stratification loads due to leaking isolation valves:

e Advises licensees of a potential safety issue due to an unanalyzed ,

design condition in unisolable lines connected to the RCS e Requests that licensees determine that this problem does not exist in the licensee's plant, or correct the problem ifit does exist e Requests that licensees provide continuing assurance that this safety issue will not develop during the life of the plant 10

/t i

( ) -USNRC Staff Response j i

The regulatory basis for Bulletin 88-08 is General Design Criterion 1.4 ~

g of 10 CFR50, Appendix A, which requires that the reactor coolant .

boundary be designed so as to have an extremely low probability of

^

abnormal leakage, ofrapidly propagating failure, and of gross rupture j t

I1

( ) USKRC Staff Response The actions requested by Bulletin 88-08 are intended to provide the assurance that GDC 14 is in effect satisfied for the life of the plant by:

Action -1. Identifyingsusceptiblesystems Action 2. Checkingcurrentadequacy Action 3. Providing continuing safety assurance for the life of the plant through one of the following options:

12

.+' ' %,

j

(, USNRC Staff Response L

1. Redesigning and modifying the piping to withstand the additional thermal cyclic loading due to leakage

2. Monitoring (by instrumenting) the piping to detect leakage .

3. Ensuring a favorable pressure differential which will prevent the in-leakage or out-leakage in case of a ~

defectivevalve  ;

13

%p"%,,)

USNRC Staff Response Most licensees who identified susceptible piping under Action 1 adopted Option 2, by installing thermocouples on the piping.

Within the past several years, industry developed simplified methods under TASCS to address Option 1 and discontinue Option 2, where applicable.

TASCS based on a combination of analytical theory and test results.

It permits the calculation of(static) thermal stratification loadmg resulting from isolation valve leakage which, when combined with a fatigue analysis and an assumed thermal cyclic frequency, permits the calculation of the time interval from initiation ofunanticipated valve leakage to crack initiation, if shorter than the remaining life of the plant.

14

fi LUS3RC Staff Response

, The NRC finds the intent of the TASCS approach acceptable, since it is ofinterest to the NRC 'that licensees determine reliably the -

-following:

! If an isolation valve starts leaking:

1. Where will the pipe crack?

2. How long before crack initiation occurs? l i

3. How long before the crack propagates to the surface?  ;

This provides a margin ofsafety to permit repair or replacement is i

i

( .....

) USXRC Staff Response i

Any analytical approach to answer these questions must:

j

• Have atechnically defensible basis a

Address the phenomena described in the Bulletin, its supplements, and other similar subsequent events.

16

i i USNRC Staff Response '

• s...../ .

Review of available literature, both public and proprietary, indicates that the basic thermo-hydraulic phenomena which caused Farley, and other similar events, is still not well understood.

! An example is the experimental simulation described in a paper by

! Nakamori and Hanzawa,in 1995:

• Only actual simulation known to the NRC staff

• Simulated actual operating conditions ofFarley safety injection line e Simulation indicated no thermal stratification or cycling at the location of the failure at Farley, as shown in the following figure 17

_.__m__ __._____._..__ _ ____ _ . _ _ ___.__ _ _ . _ _ . _ _ _____ _ _ _ _ _ _ _

A f ~~"%,,

( ) USXRC Staff Response

• The fatigue analysis predicted the failure at the check valve weld.

This is inconsistent with location of through-wall crack found at the elbow weld in Farley l

• Data used to support adequacy of the simplified methods of TASCS 18

~

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( ) Current USXRC Staff View i

The USNRC staff therefore concludes that: '

• Current simplified analytical methods do not have a defensible technicalbasis

• Temperature or pressure monitoring should be maintained in '

systems known to be susceptible to thermal stratification e Analytical methods should be evaluated with events ofFarley and Tihange as benchmarks ,

s 19

f ~*%,

(g..... ),Overview of Recent Licensee Submittals The USNRC staff has recently accepted a few licensee submittals proposing the discontinuation ofexisting temperature monitoring i 1

programs based on a combination of the following:

• In-situ monitored temperature traces indicated isolation valve leakage without thermal stratification cycling over long periods of time (one or more operating cycles). i

• The upstream source of the in-leakage was a high temperature source, such as a regenerative heat exchanger.

, i 20

i

(, j Overview of Recent Licensee Submittals i

)

• The nearest check valve was located at least 25 diameters from the RCL nozzle. '

• Isolation valve leakage was monitored by other means. t i

• Main pipe flow was very low. '

• Leaking isolation valves had already been replaced, or a commitment was made to repair or replace them at the first upcoming refueling outage. .

i J

21

....l f s,,

( ) USNRC Staff Conclusions ]

1. Test data consistent with the Farley/Tihange failures have not reported lin the literature i

2. Thermo-hydraulic phenomena which caused the events in Farley, i Tihange and other plants are not well understood

3. No simplified analytical methods which can describe these events have t beenreported

4. Temperature / pressure monitoring appears to be the most reliable method to prevent Farley/Tihange type failures in susceptible systems

5. Elimination of an existing monitoring installation may be acceptable under certain conditions, if defensible 22

f - s,,,

ig/ i USXRC Generic Communications '

\,.....  ;

1 A complete list ofrelevant bulletins, information notices, generic letters and other communications is shown in the paper t

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• UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 July 9,1997 i

NRC INFORMATION NOTICE 97-48: UNISOLABLE CRACK IN HIGH-PRESSURE INJECTION PlPING s

Adoressees All holders of operating licenses or construction permits for nuclear power reactors.

Pumose The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to the discovery of a leaking cracked weld in an unisolable section of a combined makeup (MU) and high-pressure injection (HPI) line at Oconee Unit 2. It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required.

Descriotion of Circumstances On April 22,1997, at 12:50 p.m., Oconee Unit 2 was shut down because of unidentified reactor coolant system (RCS) leakage exceeding the technical specification limit of 1 gpm.

From the time of initial leak indications on April 21, at approximately 10:45 p.m., until reactor pressure was sufficiently reduced, the leakage rate rose from approximately 2 gpm to a maximum leakage rate of approximately 12 gpm. A subsequent containment entry identified an unisolable leak in the MUMPI line 2A1 from a through-wall crack in the weld connecting the MUMPI pipe and the safe-end of the 2A1 reactor coolant loop (RCL) nozzle.

Discussion The Oconee 2A1 MU/HPI nozzle assembly consists of the MU/HPl 2% inch diameter pipe / safe-end/ thermal sleeve (see Figure 1 - Original Design). The sleeve is attached by contact rolling to the inner surface of the safe-end. A 1-inch diameter " warming" line taps into the bottom of the MUMPl pipe immediately upstream of the pipe / safe-end weld where l the through-wall crack was found. This line permits a small continuous MU flow (3 gpm) to reduce nozzle thermal transients due to changes in normal MU flow. All Oconee units have two combined MUMPl lines and two additional HPl lines connected to the RCS. However, the thermal sleeve configuration in Oconee Unit 1 is different from that in Units 2 and 3.

Preliminary analysis indicates that crack initiation and propagation in the weld was caused by high-cycle fatigue due to a combination of thermal cycling and flow induced vibration. The e *

-9707020306-

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IN 97-46 July 9,1997 Page 2 of 4 metallurgical examination of the weld determined that the crack consisted of a 360' inside surface flaw. The flaw depth increased gradually from about 30 percent into the wall untilit became through-wall over a 77* arc length (see Figure 2). The examination found a gap in the contact area between the thermal sleeve and the safe end, indicative of loss of contact

' that caused the thermal sleeve in this line to be loose (see Figure 1). The thermal sleeve was found to be cracked, with portions missing from the end that extends into the RCS flow path. Significant wear damage was observed at both the upstream (the rolled end) and the downstream end. Cracking was also found in the pipe in the vicinity of the " warming" line nozzle. Video examinations of the other thermal sleeves of the HPl system showed no evidence of damage. Ultrasonic Testing (UT) and Radiographic Testing (RT) of the welds and the thermal sleeves iri the other HPi nozzles showed no indications of cracking or loosening, or other signs of degradation. Figure 1 shows a comparison of the original and new thermal sleeve designs. The thermal sleeve in the 2A1 MUIHPI line was replaced during the current outage with the new design thermal sleeve.

! Although the root cause of the cracking is not well understood, the licensee has identified a number of thermal / mechanical conditions that may have contributed to the crack propagation of the 2A1 pipe to safe-end weld. The precise contribution to cracking of each of these conditions is not presently known. However, the licensee has hypothesized that, in addition

! to the thermal cycling experienced at the nozzle during heat up/ cool down and other plant transients, a likely contributor to the fatigue may have been the alternate heating and cooling of the weld by intermittent mixing of the hot reactor coolant leaking through the gap in the contact area between the loose thermal sleeve and the safe-end, and the cooler normal makeup water flowing through the associated MU/HPl line. Although the precise contribution of the gap is unknown, it is believed that a gap may be a prerequisite for cracking in the piping since the cracked pipes also had gaps between the thermal sleeve and the safe end.

This phenomenon was identified as the probable cause for similar safe-end cracking observed at Crystal River and other B&W plants (including Oconee) in the early 1980's. This issue was previously addressed in Information Notice 82-09 and Generic Letter 85-20.

Recent re-examination of radiographs made in April 1996 of the Oconee 2A1 nozzle revealed that the licensee had failed to identify the gap which had developed in the safe-end/ thermal sleeve contact area. The licensee also had failed to follow the original recommendations for augmented ultrasonic testing (UT) as listed in NRC Generic Letter 85-20, "High Pressure injection /Make-Up Nozzle Cracking in Babcock and Wilcox Plants," issued November 8, 1985. The licensee performed the recommended UT of the safe ends of the MU/HPI lines; however, they did not inspect the adjacent piping as recommended. In addition, the licensee failed to UT the weld between the safe-end and pipe, a discontinuity where cracking would be expected, and did, form. Also, NRC Bulletin 88-08, Supplement " Thermal Stresses in Piping Connected to Reactor Coolant Systems," issued August 4,1988, emphasized that, because of the difficulty in identifying the types of cracks that were occurring due to thermal stresses, the need exists for enhanced UT and for experienced examination personnel to detect the cracks.

e

• \

1 g

l 1

IN 97-46 July 9,1997 Page 3 of 4 3A1 MU/HPI line was found to have area. n

a. The gap As a in the result of the gap in the 3A1 safe-end, Oconee Unit 3 was , . UT shut down on removed and is presently being metallurgically s also ex revealed cracks in the thermal sleeve. Minor gaps in the other safe-e contact areas were determined not to have grown, the rolled s area of acceptable, no cracking. and UT examinations of the other Oconee Unit e 3 HPI no The Oconee Unit 1 nozzles have a double thermal sleeve design (Figu inspection in the period from 1983 to 1989 indicated. Radiographic that no gap existed in ree of the four thermal sleeves. The thermal sleeve in the 182 (HPI) ,

a not line had a gap grown the licenseeduring include: the inspection period. Advantages of the double therm (1) greater stiffness; flow area, with corresponding increased(2) flow greater velocity. thermal resistance; and (3) reduce Reoulations requires that the reactor coolant ral so as to have pres 1

~

napture. The related generic communications ,

ross listed) additional similar events occurring. events, and the actions that licen Similar Recent Events to the reactor coolant system was .found The damaged pipe inystem Dampierre length was examined and a through wall crack located on an uninterrup on of straight root cause of the cracking, but concludede that the injection system. The licensee also concluded -

on tha a straight portion of a pipe is likely to raise questions about previous assum regarding the root cause of the cracking.

Related Generic Communications I

)

LINES AT B&W PLANTS," dated March 31,1982.NRC INFOR '

INJECTION / MAKEUP NOZZLE CR,ACKING November 11,1985. , ae IN

IN 97 46 July 9,1997 Page 4 of 4 NRC BULLETIN NO. 88-08, " THERMAL STRESSES IN PlPING CONNECTED TO COOLANT SYSTEMS," dated June 22,1988.

NRC BULLETIN NO. 88-08, Supplement 1," THERMAL STRESSES IN PIPING CON TO REACTOR COOLANT SYSTEMS," dated June 24,1988.

NRC BULLETIN NO. 88-08, Supplement 2," THERMAL STRESSES IN PlPING CONN TO REACTOR COOLANT SYSTEMS," dated August 4,1988.

NRC BULLETIN NO. 88-08, Supplement 3, " THERMAL' STRESSES IN PlPING CONN

-TO REACTOR COOLANT SYSTEMS," dated April 11,1989.

NOTICE 97-19, " SAFETY INJECTION SYSTEM WELD FLAW AT

.NRC INFORMATION SEQUOYAH NUCLEAR POWER PLANT, UNIT 2," dated April 18,1997 This information notice requires no specific action or written response. If you have any questions about the information in this notice, please contact one of the technical contact ,

listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

Marylee M. Stosson, Acting Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts: Barry Elliot, NRR Eric Benner, NRR -

301-415-2709 301-415-1171 E-mail: bje@nrc. gov E-mail: ejb1@nrc. gov Kamal Manoly, NRR Mark Hartzman, NRR 301-415-2765 301-415-2755.

E-mail: kam@nrc. gov E-mail: mxh@nrc. gov I Attachments:

p 1. Figure 1 - Thermal Sleeve

2. Figure 2 - Warming Line Flow

3. Figure 3 - Unit 1 Thermal Sleeve

~ 4. List of Recently issued NRC Information Notices e

I t

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SAFE END TO NOZZLE WELO MAIN (BI-E TTALIC) N0ZZLE( C. S. ) COOLANY PIPE d CLAODING7/

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2A1,281,3A1,381,382

Attachment 2.

IN 97-46 .

July 3, 1997 .

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( Attachment 4 i

" IN 97-46 i July 9,1997 l Page 1 of 1 LIST OF RECENTLY ISSUED NRC INFORMATION NOTICES

$~ l Date of l Information issuance issued to Notice No. Subject 06/27/97 All " users and fabricators" 97-47 Inadequate Puncture of type B transportation Tests for Type B packages (as defined in

. Packages Under 10 CFR 10 CFR 171.16(10)(B) 71.73(c)(3) 07/02/97 All holders of OL permits 96-44, Failure of Reactor for nuclear power reactors Supp.1 Trip Breaker from Cracking of Phenolic Material in Secondary Contact Assembly 07/02/97 All holders of OLs or cps-97-45 Environmental for nuclear power reactors Qualification Deficiency for Cables and Contain-

. ment Penetration Pigtails m.

07/01/97 All holders of OLs or cps

-97 44 Failures of Gamma Metrics Wide-Range for test and research reactors Linear Neutron Flux Channels ,

07/01/97 All holders of OLs or cps 97-43 License Condition for nuclear power reactors Compliance 06/27/97 All fuel cycle conversion, 97-42' Management Weaknesses enrichment, and f abrication Resulting in Failure facilities to Comply with Shipping Requirements for Special Nuclear Material 06/27/97 All holders of OLs or cps 97-41 Potentially Undersized for boiling-water reactors Emergency Diesel Generator Oil Coolers p OLVOperating License

. CP = Construction Permit

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