ML19305D408
| ML19305D408 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1980 |
| From: | Liaw B, Strosnider J Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0571, NUREG-571, NUDOCS 8004150042 | |
| Download: ML19305D408 (41) | |
Text
-
NUREG-0571 2o**2o }2ob]Yg_
x Summary of Tube Integrity
. Operating Experience With Once-Through Steam Generators l
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B. D. Liaw, J. Strosnider Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 0@ 50316)/
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Available from GP0 Sales Program Division of Technical Information and Document Control U.S. Nuclear Regulatory Commission Washington, D.C.
20555 and National Technical Information Service Springfield, Virginia 22161
NUREG-0571 Summary of Tube Integrity Operating Experience With Once-Through Steam Generators s = "; 2 a m n it'" " ' *
- B. D. Liaw, J. Strosnider Division of Operating Reactors Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission W chington, D.C 20566 f ~%,
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ABSTRACT Operating problems have occurred in the steam generators designed by each of the three manufacturers of pressurized water reactor nuclear steam supply systems:
Babcock and Wilcox, Combustion Engineering, and Westinghouse Electric Corporation.
This report focuses on the problems associated with steam generators of the once-through type that are designed and manufactured by Babcock and Wilcox.
It identifies the operational experiences observed to date and the position the NRC staff has taken to ensure steam generator tube integrity.
It should be noted that a number of research efforts related to these problems are currently under way and that the information included in this report represents our current understanding of each issue.
A similar report, NUREG-0523, discusses tube integrity experience with the U-tube type of steam generator designed by Westinghouse and Combustion Engineering.
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TABLE OF CONTENTS Page ABSTRACT...........................................................
iii 1.
INTRODUCTION.......................................................
1 2.
OPERATING EXPERIENCE...............................................
9 2.1 Experience at Oconee Nuclear Station..........................
9 2.1.1 Circumferential Fatigue Cracks.........................
9 2.1.2 Surface Defects at the 14th Tube Support Plate.........
13 2.1.3 Serpentine Depression..................................
13 2.1.4 Axial Surface Defects at the 15th Tube Support Plate...
13 2.1.5 Circumferential Wear at the 15th Tube Support Plate....
13 2.1.6 Pitting at the Tube Support Plate......................
13 2.1.7 Cavitation.............................................
14 2.2 Experience at Other B&W Uni ts.................................
14 2.2.1 Arkansas Unit 1........................................
14 2.2.2 C rys tal Ri ve r Uni t 3..................................
14 2.2.3 Davis-Besse Unit 1.....................................
16 2.2.4 Rancho Seco Unit 1.................
16 2.2.5 Three Mile Island Unit 1...............................
17 2.2.6 Three Mile Island Unit 2...............................
17 3.
DISCUSSION.........................................................
19 3.1 Steam Generator Inservice Inspection Requirements.............
19 3.2 Tube Plugging Criteria........................................
19 3.3 Prima ry to Secondary Leakage Rate Limit.......................
21 3.4 Turbine Stop Valve Testing...................................
21 3.5 Tube Sleeving.................................................
21 4.
SUMMARY
AND CONCLUSIONS............................................
22 5.
REFERENCES.........................................................
23 APPENDIX A - Task A Babcock and Wilcox Steam Generator Tube Integrity.................................
A-1 APPENDIX B - Rancho Seco Supplement to Monthly Plarit O p e ra t i o n s R e p o r t..........................................
B-1 y
4 LIST OF FIGURES Figure Title M
1 Pressurized Water Reactor (JWR) Cooling Cycles................
2 2
Schematics of PWR Reactor Coolant System......................
3 3
Once-Through Steam Generator..................................
4 4
B&W Tube Support Plate Design.................................
6 5
Steam Generator Tube Identification System....................
12 I
vi
LIST OF TABLES Table Title Page 1
Typical Feedwater Chemistry Specifications....................
7 1
2 Operating PWR's With B&W Steam Generators.....................
8 3
Tube Leaks at 0conee..........................................
10 4
Tubes Removed from the Oconee Steam Generators for Examination.................................................
11 5
Arkansas Unit 1 Steam Generator (S.G.) Inservice Inspection Results..........................................
15 6
Summary of B&W Steam Generator Inspections....................
18 7
Steam Generator Tube Inspection...............................
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SUMMARY
OF OPERATING EXPERIENCE WITH ONCE-THROUGH STEAM GENERATORS i
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1.
INTRODUCTION Nuclear power plants using the pressurized water reactor (PWR) design concept contain three separate cooling cycles.
The three cooling cycles of a typical PWR are shown in Figure 1.
The first cooling cycle consists of the primary coolant system that pumps pressurized coolant water through the heat generating core of~the reactor where it picks up heat.
The second cooling cycle consists of large heat exchangers called steam generators, a steam-driven turbine genera-tor, a steam condenser, feedwater pumps, and associated piping systems.
Heat generated in the primary coolant system is transferred to the secondary system 1
through steam generators. The water in the secondary coolant system boils in the steam generator creating steam that is used to drive the turbine generator.
Aftar it passes through the turbine generator, the steam is condensed back into water in the steam condenser. The secondary cooling water is returned to steam generators by the feed pumps, thereby completing the cycle. The third cooling cycle is the condenser cooling water system that provides cold water to condense the steam back to water in the steam condenser. The basis for this closed-cycle system is to ensure that the radioactive primary coolant water, the secondary cooling water, and the condenser cooling water are separated from each ott.er.
The steam generator is the connecting link between the radioactive primary and nonradioactive secondary coolant system and is, therefore, a principal part of the reactor coolant pressure boundary.
Figure 2 shows the major components of the primary coolant system for a Babcock and Wilcox reactor.
Two major types of steam generators are currently in use in pressurized water reactors in the United States.
These are the recirculating type manufactured by Westinghouse and Combustion Engineering and the once-through type manufactured by Babcock & Wilcox (B&W).
Experience with the recirculating type of steam generators was reported in NUREG-0523, " Summary of Operating Experience With Recirculating Steam Generators" (Ref. 1).
This report discusses the operating l
experience with B&W once-through steam generators.
Figure 3 shows a typical once-through steam generator.
In this steam generator design, there are approximately 16,000 straight tubes that terminate in a tube sheet at each end of the steam generator.
These tubes have an outside diameter of approximately 0.625 inch and a wall thickness of 0.035 inch.
Primary coolant water enters the steam generator inlet plenum and flows from top to bottom through the steam generator tubes and exits through the steam generator outlet plenum.
The secondary coolant water enters through the steam generator feedwater nozzles, flows down through an annulus around the downcomer section, and flows up around the tubes.
The auxiliary feedwater is introduced through an internal header located in the upper region of'the steam generator.(Figure 3).
Boiling occurs about two-thirds of the way up the tube bundle.
The area above the boiling region is the steam region in which the tubes are exposed to steam (in contrast to the U-tubes in the recirculating types of steam generators that remain covered with water).
The steam is superheated in this region.
After passing through the superheated region near the upper tube sheet, the steam flow shifts from vertically upward to radially outward then abruptly downward 1
CONTAINMENT STRUCTURE CONTROL $ [
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PRES $URE INJECil0N N0Z2LE (2)
FEE 0 WATER INLETS (2)
- COLO LEG PIPING REACTOR VESSEL Figure 2. Schematic of B&W Reactor Coolant System.
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TUBE SHEET PRIMARY OUTLET NOZZLES Figure 3. Once-Through Steam Generator.
4
l through the annulus between the shell and the baffle and then through the steam generator outlet nozzles.
The tubes are supported vertically by the tube sheets i
and are restrained from horizontal movement along their length by 16 tube support plates (the exact number depends on design, but it is generally 15 or 16).
The support plate-tube intersection design shown in Figure 4 is used for all support plates except a portion of the 15th support plate which has drilled tube holes.
This plate has drilled holes near the periphery which enhance steam flow through the center of the bundle and into the upper annulus of the steam generator.
Each once-through steam generator is subjected to a full-furnace stress relief after fabrication, including the installation of steam generator tubes.
This treatment reduces the magnitude of residual stresses and produces a microstructure with improved resistance to stress corrosion cracking.
All B&W steam generators are operated with all-volatile secondary water treatment and full-flow condensate demineralizers.
Hydrogen and ammonia are added for dissolved oxygen and pH control, respectively.
Typical feedwater chemistry specifications recommended by B&W are given in Table 1 (Ref. 2).
The feedwater chemistry specifications vary somewhat from plant to plant depending on the materials used in the condenser, feedwater heaters, and other secondary system equipment.
There are currently nine PWR nuclear units with B&W steam generators operating in the United States.
These units are listed in Table 2.
Each unit uses two once-through steam generators.
Section 2 of this report describes the operating experience at these units and defines the nature of the operational problems that have occurred to date.
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Table'1.
TYPICAL B&W FEE 0 WATER CHEMISTRY SPECIFICATIONS Power Operation Average Values June '73 Nov. '73 Feedwater Preop.
To To Chemistry Spec.
Avo. Value Nov. '73 March '74 pH @ 77*F (s 25'C) 9.3-9.5 9.4 9.4 9.5 Cation cond.,
< 0.5 0.45 0.5 0.4 pmho/cm Total solids, PPB
< 50 Fe, PPB
< 10*
50 10
< 10 SiO, PPB 20 20
< 20
< 20 2
Na, PPB 20 15
< 10(2-3) 0, PPB
<7
< 10
< 10
<7 2
N H, PPB 24 150 30-50
'5-100 Pb, PPB
<1
<1
<1
<1 Cu, PPB
<2
<2
<2
<2
- 100 PFB for preoperational testing.
i 7
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. Table.2. OPERATING PWR'S WITH B&W STEAM GENERATORS Date of Initial Plant Name Connercial Operation 4
Arkansas Unit 1 12/19/74 Crystal River Unit 3
'3/13/77 Davis-Besse Unit 1 11/20/77 Oconee Unit 1 7/15/73 Oconee Unit 2 9/9/74 Oconee Unit 3 12/16/74 Rancho Seco Unit 1 4/17/75.
- Three Mile Island Unit 1 9/2/74
- Three Mile Island Unit 2 12/30/78 l
- Currently shut down following the TMI-2 accident.
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2.
OPERATING EXPERIENCE 2.1 Experience at Oconee Nuclear Station As of Ncvember 1979, 17 tube leaks have occurred in B&W steam generators.
Sixteen of these leaks occurred at the Oconee Nuclear Station and are summarized in Table 3.
In October 1979, Crystal River Unit 3 experienced a small tube-to-tube sheet seal weld leak, described in Section 2.2.2.
Among those found to be defective, seven tubes given in Table 4 have'been removed from the Oconee units for laboratory examination, and B&W has categorized the following forms j
of observed tube degradation (Ref. 3).
(See also Appendix A.)
2.1.1 Circumferential Fatigue Cracks The first tube leak in a B&W steam generator occurred on July 21, 1976, in the Oconee Unit 3 steam generator B.
Subsequent plant shutdown and inspection revealed that the leaking tube was the lith tube in column 77 (C77T11) near the open inspection lane.* Figure 5.shows the steam generator tube identifica-tion system.
Eddy current testing (ECT) revealed that the defect was located at the uppermost 15th support plate level.
The tube was removed from service i
by plugging.
This event was the first indication of abnormal degradation in any B&W steam generator.
Prior to this date, two inservice inspections of the Oconee Unit 1 steam generators and one inspection of the Oconee Unit 2 steam
~
generators had been performed.
These inspections included eddy current testing in accordance with Regulatory Guide 1.83 and visual and fiber optic inspections.
The eddy current testing did not reveal any tubes with greater than 20 percent wall penetration and, in general, no evidence of abnormal degradation had been observed.**
A second steam generator tube leak occurred on October 31, 1976, in the Oconee Unit 1 steam generator A.
The tube, C77T17, was again located on the open lane, and visual inspection using fiber optics revealed a circumferential crack near the upper tube sheet. This tube was also removed from service by plugging.
On December 4, 1976, a leak developed in the Oconee Unit 2 steam generator B.
The leaking tube was identified as C77T23, again a lane tube, and was determined to have a 270 circumferential crack at the upper tube sheet.
This tube was removed from the steam generator and subjected to visual, chemical, and metal-lurgical examinations.
The metallurgical examination revealed that the crack had initiated on the outside surface of the tube and propagated through the wall, and then had continued circumferential1y in both directions around the tube.
The propagation of the crack around the tube was attributable to a high-frequency, low-stress fatigue mechanism.
The crack-initiating mechanism has not been identified although B&W has suggested that initiation may have been environmentally assisted (Refs. 3, 4).
A total of seven circumferential fatigue-cracks have been observed to date but only at the Oconee units (Refs. 5, 6).
" Column 76 was left untubed from the center to the outer periphery of the tube bundle, thus leaving an open tube lane to facilitate the steam generator inspection.
- Eddy current testing is not reliable at or below 20 percent and eddy current indications below that value are therefore not considered indicative of l
abnormal tube degradation.
9
Table 3.
TUBE LEAKS AT OCONEE Date Generator Row Tube Elevation Condition 12/76 1B 114 109 14th support plate ECT indication 12/76 1B 75 18 Upper tube sheet 300 circumferential crack **
1/77 1B 75*
12' Upper tube sheet 350 circumferential crack 2/77 1B 32 13 14th support plate ECT indication 3/77 1B 77*
25 Upper. tube sheet Weld crack 5/77 1B 77*
15 Upper tube sheet Crack 4/78 1B 74 2
Upper tube sheet 45 -90* circumfer-ential crack 4/78 1B 69 1
Upper tube sheet Weld crack 7/79 1B 73 130 14th support plate Not identified 12/76 2B 77*
23 Upper tube sheet 270 circumferential crack 1/78 2B 77*
25 Upper tube sheet 90 circumferential crack 7/76 3B 77*
11 15th support plate ECT indication 2/77 3B 77*
19 15th support plate 45 circumferential
~;
crack 6/77 3B 78 1
15th support plate 90 circumferential crack
-7/77 3B 77*
2 Upper tube sheet 60*-90* circumfer-ential crack 10/76 1A 77*
17 Upper tube sheet Crack l
- Tubes located on open lane.
l
- Probable. leaker.
10
Table 4.
TUBES REMOVED FROM THE OCONEE STEAM GENERATORS FOR EXAMINATION Steam Generator Row Tube Disposition 1B 75 18 Had been previously stabilized due to a 300 circumferential crack at upper tube sheet 1B 83 117 ECT indication at 14th tube support plate 1B 43 108 ECT indication at 14th tube support plate 1B 77 25 Leaker 2B 77 27 Distorted ECT signal at upper tube sheet 28 75 9
ECT indication at 15th tube support plate 2B 77 23 Leaker
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2.1.2 Surface Defects at the 14th Tube Support Plate During steam generator inspections at Oconee Unit 1 in September 1977, several tubes in the outer periphery of the bundle had eddy current indications greater than 40 percent through-wall near the 14th support plate level.
One tube with a 40 to 50 percent through-wall indication and one tube with approximately a 90 percent through-wall indication were removed for examination.
Visual inspec-tion of the tube with the 40 to 50 percent indication revealed an eroded area with the shape of an inverted pyramid about 1/8 inch long and 1/16 inch wide and approximately 0.02 inch deep.
Inspection of the second tube revealed a shallower canJle-flame-shaped spot covering a larger tube surface area approxi-mately 1 inc'i long and 0.3 inch wide.
During the inspection that included ECT of approximately 7000 tubes in both steam generators, 32 defective tubes (greater than 40 perces;t wall thinning) and 44 degraded tubes (between 20 and 40 percent wall thinning) were found in steam generator B, and five defective and eight degraded tubes were found in steam generator A.
The most likely cause of this type of degradation is postulated as either liquid impingement erosion, corrosion pitting, or a combination of the two (Ref. 3).
2.1.3 Serpentine Depression Some of the tubes removed for laboratory examination had what B&W describes as serpentine depressions located at or'below the lower surface of the upper tube sheet.
These depressions follow a twisting path about 20 mils wide and 2 mils deep circumscribing up to half the tube's circumference.
These depressions have been associated with some of the through-wall cracks and partial through-wall microcracks.
B&W has suggested that under-deposit corrosion, electrochemical corrosion, or wear between the tube and the tube sheet could be possible causes of this type of degradation.
2.1.4 Axial Surface Defect at the 15th Tube Support Plate An axially aligned defect located on the outside tube surface has been observed in connection with through-wall cracking and partial through wall microcracks at the 15th tube support plate elevation. The defect was located at the "la'd" areas of the tube support plate and was 3 to 5 mils deep.
Fretting and/or chemical milling are believed to be the possible causes.
2.1.5 Circumferential Wear at the 15th Tube Support Plate Wear as a result of fretting between the tube and tube support plate land was observed at the 15th support plate.
The wear consisted of scratches 8 to 10 mils deep and 3 mils in length circumferentially around the tube.
A large amount of magnetite was packed on the surface.
2.1.6 Pitting at the 15th Tube Support Plate One tube removed for examination had pitting at the 15th tube support plate land contact.
The pits were round in shape and a maximum of 6 mils deep.
Pitting was not observed on any other tubes.
B&W has attributed the observed pitting to local chemical conditions that develop in the space between the tube and tube support plate probably during shutdown.
13
2.1.7 Cavitation Scattered areas of cavitation were also observed near the upper tube sheet on one tube, and cavitation may have also beer, associated with the surface defects at the previously discussed 14th tube support plate.
2.2 Experience at Other B&W Units Circumferential cracking has been limited to tubes in the Oconee-steam generators; however, other forms of degradation have also been observed at other units.
2.2.1 Arkansas Unit 1 Baseline inspection of the Arkansas Unit 1 steam generators conducted _in 1977 revealed no defective tubes.
The first inservice inspection at the unit was conducted in February 1977 after approximately 500 days of effective full power operation. Approximately 3.2 percent of the tubes were inspected in each of the two steam generators.
No defective tubes were detected; however, three tubes in steam generators A and one tube in steam generator B had ECT indica-tions of shallow secondary side cracking.
The tubes were all located within two or three rows of the open inspection lane.
The defects were located at various elevations between the 2nd and 14th tube support plates.
The maximum indicated depth was 30 percent through wall.'
The second inservice inspection at Unit 1 was conducted in February 1978.
During the initial 3 percent inspection, three defective tubes were discovered in steam generator A.
An additional 6 percent sample in steam generator A revealed two additional defective tubes.
An additional 12 percent inspection was conducted in steam generator A.
Approximately 9.2 percent of the tubes in steam generator B were inspected with no defective tubes discovered.
Table 5 summarizes the tube degradation observed during the two inspections.
Defective tubes were plugged except for tube C77T17 that was removed from steam generator A for examination.
Metallurgical examination of the tube revealed two cracks. One tube crack was at the lower surface of the upper tube sheet elevation and the second crack was approximately 5/8 inch above that.
The i
cracking was intergranular in nature and Arkansas Power and Light Company has reported stress corrosion as the cracking mechanism (Ref. 7).
The examination indicated that fatigue was not an active mechanism in the cracking process and that inside diameter (ID) constriction did not exist near the rolled area.
The cavitation-erosion tube degradation oliserved in other B&W units was not reported to have been observed at Arkansas Unit 1 as of February 1978.
2.2.2 Crystal River Unit 3 Crystal River Unit 3 began commercial opr. ration on March 13, 1977.
In March 1978, the breakup of a burnable poison.-od assembly resulted in damage to the l
tube-to-tube-sheet welds in the Unit ? steam generator B.
In addition, debris became lodged in a small nercentage of the tubes in steam generator B.
Correc-tive actions taken by Florida Power Corporation included video inspection of the damaged tube stubs, leak testing, a free path check of 100 percent of the A and B steam generator tubes, eddy current testing, and tube plugging.
I 14
'1 Table 5.
ARKANSAS UNIT 1 STEAM GENERATOR INSERVICE INSPECTION RESULTS Steam Generator
- Tube 2/77. Inspection 2/78 Inspection A
79-13 30%
NI**
A 77-6 25%
NI A
79-16.
NI 35%
A 75-27 Distorted signal 70%-
A 75-15 NI 70%-
A 79-15 30%
45%
B 77-41 20%
20%
B 74-15 20%
20%
B 74-12 30%
30%
B 77-10 NI 25%
A 77-17 NI 40%
A 11-3 NI 40%
" Indications of 20% or less are not considered reliable indications of degradation.
- NI - no indication.
15
The tube stubs terminate approximately 0.3 inch out of the tube sheet.
The tubes are expanded approximately 1 inch into the tube sheet and have a 0.1-inch fillet seal weld.
Visual examination of steam generator A revealed no debris on the upper tube sheet and no tube end or tube-to-tube-sheet weld damage.
Over 90 percent of the tube ends in steam generator B experienced some degree of damage in the form of cold working.
No cracks were observed.
The leaktightness of the seal welds in both steam generators was verified by pressurizing the partially filled secondary side of the steam generator with l
helium and inspecting each weld individually with a mass spectrometer capable of detecting a 10 8 cc/sec leak.
No leaks were observed.
A 100 percent free path check of all tubes in steam generators A and B was performed.
Seven tubes in steam generator B, in which debris had lodged in them that could not be removed, were plugged.
Eddy current inspection of 3 percent of the tubes, plus the 19 tubes from which debris was removed, was conducted in steam generator B.
Seven percent of the tubes in steam generator A were inspected, resulting in the plugging of one tube that had an ECT indication.
No other significant ECT indications were observed (Ref. 8).
Results of the steam generator inspection conducted during the May 1979 refueling outage are not available at this time.
2.2.3 Davis-Besse Unit 1 Davis-Besse Unit 1 achieved initial criticality on August 12, 1977, and began commercial operation on November 21, 1977.
No preservice inspection was performed and no inservice inspection has been conducted to date.
The technical specifications for Unit 1 currently require that an inservice inspection of the steam generator be conducted before August 12, 1979.
2.2.4 Rancho Seco Unit 1 The first inservice inspection (ISI) at Rancho Seco Unit 1 was conducted in September 1977 after approximately 460 effective full power days of operation.
The inspection included approximately 9.0 percent of the tubes in steam generator A and 3.3 percent of the tubes in steam generator B.
No defective tubes were found in steam generator B.
In steam generator A, four defective tubes were idt.ntified and plugged.
The defects were located at the 15th tube support plate e;evation.
No specimens were removed, and therefore the nature j
of the defects could not be positively identified.
However, Sacramento fiunicipal Utility District nas attributed the degradation to wear.
An additional four i
tubes with simila indications at the same elevation, but less than 49 percent through-wall, were plugged as preventive measures.
The second inservice inspection was conducted in November 1978.
The inspection included 100 percent of the lane tubes (tubes within three rows of either side of the open inspection lane) and approximately 3.5 percent and 3.2 percent of the remaining tubes in steam generators A and B, respectively.
The inspection revealed 205 lane tubes in steam generator A and 167 lane tubes in steam generator B to have 0.001-to 0.003-inch " dings," which are a reduction in tube diameter over a small part of the tube's circumferenre'with no change in wall _
thickness.
Seven off-lane tubes in steam generator A and five off-lane tubes' l
in steam generator B had indicators of 0.005-to 0 006-inch " dings" at the 16
l l
t 15th support plate.
The licensee has attributed these phenomena to tube vibra-tion.
In addition, four tubes in steam generator B were discovered to have between 20 and 40 percent outside diameter tube wall thinning (Ref. 9).
(See also Appendix B.)
2.2.5 Three Mile Island Unit 1 Three Mile Island Unit 1 has performed two inservice inspections in addition to a 3 percent bas.nline inspection.
The first inservice inspection was conducted in March 1977 after 257 days of commercial operation.
Approximately 6.3 and 3.2 percent of the tubes were inspected in steam generators A and B, respectively.
Two defective tubes in steam generator A and six defective tubes in steam generator B were plugged as precautionary measures.
The nature of defects were not characterized (Ref. 10).
The second inservice inspection was conducted in March 1978 after approximately 544 full power days of operation.
Approximately 12 and 9 percent of the tubes were inspected in steam generators A and B, respectively. One defective tube was discovered and plugged in steam generator A.
The tube was located on the periphery of the tube bundle with an approximately 65 percent through-wall defect 3 inches above the 3rd support plate.
The defect has been categorized as a manufacturing flaw (Ref. 11).
No defective tubes were discovered in steam generator B.
2.2.6 Three Mile Island Unit 2 Three Mile Island Unit 2 achieved initial criticality on March 28, 1978, and started commercial operation in December 1978.
An inservice inspection of the steam generator was not performed prior to the March 28, 1979, accident.
- However, a baseline inspection of 100 percent of the tubes in both steam generators was conducted in November and December 1977 following the hot functional tests.
This was the first B&W unit to conduct a 100 percent baseline inspection.
The inspection revealed nine tubes in steam generator A and 21 tubes in steam generator B with greater than 40 percent wall thinning.
Two tubes in steam generator A and 10 tubes in steam generator B had degradation between 20 and 40 percent.
In addition to the thinned tubes, appro).imately 2.4 percent of the tubes in steam generator A and 2.2 percent of the tubes in steam generator B had indications described as 0.001-to 0.003-inch "Jings." All tubes with greater than 40 percent through-wall defects were plugged.
In addition, four tube samples were removed for laboratory examination.
The examination revealed that the " ding-like" indications were not assoc'ated with a ieduction in wall thickness and would not result in a decrease in tube integrity.
B&W performed a comprehensive review of the steam generator manufacturing process and concluded l
that the " dings" and other tube anomalies detected during the inspection occurred during manufacture of the tubes and the steam generators.
A good correlation I
between the locations of " ding" indications on the outer surfaces of tubes and i
the support plate spacings tends to confirm the conclusion that tubes had been scratched by the edges of support plate crevices during the tubing process.*
Table 6 summarizes the results of B&W steam generator inspections described above.
l
- This conclusion was reached at a meeting with representatives from Metropolitan Edison Company, Babcock & Wilcox, and the NRC staff on January 25, 1978.
17
m.
.,__,4 L
Table 6.
SUMMARY
OF B&W STEAM GENERATOR INSPECTIONS Crystal Davis-Rancho Three Mlle Three Mile Inspection Arkansas 1 River 1 Besse 1 Oconee 1 Oconee 2 Oconee 3 Seco 1 Island 1 Island 2 Activities 5G A 5G B 5G A SG B 5G A 5G B 5G A SG B 5G A 5G B SG A SG B 5G A 5G B SG A 5G B 5G A SG B Preservice Inspection ~
None Before 1/77 None None Norte None None Before 3/76 12/77
% inspection 100%
100%
3%
3%
100%- 100%
No. tubes degraded NA NA
< 2.4% < 2.2%
No. tubes plugged NA NA 9
21 Primary degradation mechanisms NA NA manufacturing ist ISI 2/77 3/74 8/79 11/74 4/76 11/76 9/77 3/77 None 100% free path
% inspection 3.2%
3.2%
3.2%
3.1%
3.1%
3.8%
3.2%
9%
3.3%
6.3%
3.2%
No. tubes degraded 3 1
1
'i NA 02 os 02 g2 02 4
NA 02 NA NA No. tubes plugged. 0 0
1 7
NA 0
0 0
0 0
0 4
0 2
6 Primary degradation Poison roa mechanisms IGSCC breakup NA NA NA NA NA NA NA wear NA 2nd ISI 2/78 5/79 3/76 8/77 10/77 11/78 3/79 100% of lane tubes
% inspection 21.4%
9.2%
NA 3%
3.2%
3.3%
6.4%
7%
7% 3% of remaining 12%
9%
No. tubes degraded 4.2%
4.2%
NA 02 02 g2 NA 02 6%
212 172-NA NA No. tubes plugged 5 1
NA 0
0 0
4 0
0 0
0 1
Primary degradation dings -
manufacturing mechanism IGSCC NA NA NA NA NA vibration flow 3rd ISI 8/77 6/78
% inspection 16%
32%
5.7%
7.1%
No. tubes degraded 8
44 02 9
No. tubes plugged 5
34 0
0 Primary degradation impingement NA NA mechanism erosion or pitting l
Leak History None t
None see Table 3 see Table 3 see Table 3 None None None prior to l
3/28/79 i
l Tubes Removed 1-4 3
l NA = Information not available Note: 1 Leak at two tube / tube sheet welds in October 1979 due to damage caused by poison rod breakup in March 1978.
8ECT indications < 20% through wall.
8Not including inspections resulting from forced outages.
l l
i n
3.
DISCUSSION The NRC staff approach to ensure steam generator tube integrity during normal operation and under accident conditions is threefold:
(a) inservice inspection requirements, (b) preventive tube plugging, and (c) primary-to-secondary leakage rate limits. The following sections discuss the status in each of the three areas with respect to B&W designed steam generators.
6 3.1 Steam Generator Inservice Inspection Requirements The program for inservice inspection of steam generator tubes, as set forth in the B&W Standard Technical Specifications, is a modification of Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes." It is an augmented program designed to pro ;de more extensive inspection of steam generators with evidence of abnormal tube degradation.
Table 7, which is taken from the Standard Technical Specifications, outlines the inspection requirements.
Degraded tubes are those that have a reduction in wall thickness of greater than 20 percent but less than the plugging limit, which is the maximum allowable reduction in tube wall thickness minus an operational allowance.
The NRC has approved a modified version of the Standard Technical Specifications for Three Mile Island Unit 1, Arkansas Unit 1, and Davis-Besse and is currently reviewing such specifications for Oconee Units 1, 2, and 3.
This modified version treats tubes in areas of unique operating conditions or physical construction separately from the randomly selected tube samples.
Specifically, tubes within three rows of the open inspection lane, where fatigue cracks have occurred in the Oconee units, and tubes that pass through drilled holes in the 15th support plate rather than the broached openings, are subject to 100 percent inspection.
Inspection of these tubes is not considered to be part of the required 3 percent random inspection, and the results of the inspection of these tubes are not used in classifying the random inspection results into the C1, C2, or C3 category.
This form of inspection therefore distinguishes between random and deterministic forms of degradation.
Similar technical specifications are under review for other B&W units including the Oconee units.
3.2 Tube Plugging Criteria The plugging limit is established in accordance with criteria in Regulatory Guide 1.121, " Basis for Plugging Degraded PWR Steam Generator Tubes." B&W has conducted burst and collapse tests on steam generator tubes with simulated defects to establish the extent of " allowable" tube degradation. -Specimens tested included undamaged tubes and tubes with varying depths of longitudinal slits, circumferential slits, uniform thinning, and long, flat defects on tube outer surface.
The burst and collapse tests were run at normal operating temperature.
Based on the burst and collapse test data and on calculations performed in accordance with Regulatory Guide 1.121, B&W has calculated the j-maximum defect depths allowable under normal operating or accident conditions.
l Including a margin for continued degradation between inspections and for error l
in eddy current testing, the plugging limit for B&W steam generator tubes has been conservatively established as 40 percent.
This tube plugging criteria assures that tubes will not become degraded to the extent that they could fail j
during postulated accident conditions prior to the next inservice inspection.
Preliminary results of an NRC-sponsored research program at Battelle Pacific 19
TABLC 7. STEAM GENERATOR TUBE INSPECTION
- 1ST SAMPLE INSP8'CTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1"*
None N/A N/A N/A fl/A S Tubes per S.G."
C-2*"
Plug defective tubes C-1 None N/A N/A and inspect additional Plug defective tubes C-1 None 2S tubes in.his S. G.
C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G.
Perform action for C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N/A N/A sample E$
C-3"
- Inspect all tubes in All other this S. G., plug de-S. G.s are None N/A N/A fective tubes and C-1 inspect 2S tubes in Some S. G.s Perform action for N/A N/A each other S. G.
c ; i,ut no C-2 result of second additional sample Prompt notification S. G. are to NRC pursuant C-3 to specification Additional inspect all tubes in 6.9.1 S. G. is C-3 each S. G. and plug defective tubes.
Prompt notification N/A N/A to NRC pursuant to specification 6.9.1
- Source: B&W Standard Technical Specifications.
" S = 3 $ % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection.
"
- C-1: Less than 5% of the total tubes inspected are degraded tubes and more of the inspected tubes are defective.
"
- C-2: One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% anj 10%
of the total tubes inspected are degraded tubes.
"* C-3: More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
f
Northwest Laboratory (Ref. 12) confirm the conservatism of the 40 percent plugging criteria.
3.3 Primary to Secondary Leakage Rate Limit In addition to tube plugging, a primary to secondary leakage rate limit is included in the technical specifications for each nuclear power plant.
The leakage rate is based on radiological and mechanical considerations.
A total steam generator tube leakage limit for all steam generators not isolated from the reactor coolant system ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The leakage rate limit is also limited by the leak rate associated with the form of through-wall defect that is most limiting under postulated accident conditions.
By limiting the allowable leak rate during normal operation, it is ensured that any tube that develops a through-wall defect during operation will not fail during postulated accident conditions.
3.4 Turbine Stop Valve Testing An important issue receiving continued attention is the relationship between turbine stop valve testing and steam generator tube failures.
In February 1977, B&W identified turbine stop valve testing as a potential cause of tube failures.
Instrumentation was installed in Oconee Unit 2 steam generator B and a series of tests were conducted.
The tests indicated that turbine stop valve testing does result in steam generator pressuro and flow transients.
Results of similar tests conducted at Three Mile Island Unit 2 have not been fully evaluated.
Although the relationship between stop valve testing and the tube failures has not been specifically identified, B&W has recommended several corrective actions for the Oconee and other B&W units.
The actions taken at the Oconee units included hardware modifications to mitigate the effects of stop valve testing.
In addition, all B&W units have reduced the frequency and power level of stop valve testing.
A decrease in testing frequency generally leads to a decrease in turbine stop valve reliability, which is important for ensuring that the risk from turbine missiles is sufficiently small.
Therefore, the frequency of testing must be carefully determined to minimize detrimental effects to the steam generators while providing the necessary level of valve reliability.
3.5 Tube Sleeving B&W has developed steam generator tube sleeves intended to stiffen the tubes, thereby reducing dynamic stresses resulting from flow-induced vibration.
The tube sleeves vary in length from approximately 1 foot to 1 feet and are secured inside the generator tube by two expanded regions.
One or more sleeves can be installed in a given tube to achieve the desired vibration characteristics.
The sleeves are not intended to perform as part of the primary coolant system boundary and are not used for repairing degraded tubes.
The tube sleeves were qualified analytically and experimentally, and demonstration programs involving installation of a small number of tube sleeves were approved by the NRC for the Oconee units and Three Mile Island Unit 2.
No inservice inspection of the steam generators was performed at Three Mile Island Unit 2 prior to the March 28, 1979, accident.
Inspection of the tube sleeves at the Oconee units are being performed during each inservice inspection.
21
4.
SUMMARY
AND CONCLUSIONS To date, the Oconee Nuclear Units are the only plants where steam generator tube leaks have occurred.
The circumferential fatigue crack is the most signif-icant form of degradation that has been observed because it poses the greatest safety threat.
It has also resulted in the greatest amount of down time at B&W units.
However, the NRC licensing requirements, including inservice inspec-tions, tube plugging,'and leakage rate limits, address this and the other types of degradation that have been observed in several B&W units and ensure safe plant operation.
In relation to steam generator problems in B&W and other manufacturers' steam generators, the NRC is currently conducting.several programs leading to better definition and control of steam generator tube degradation phenomena (Ref. 13).
The three tasks addressing B&W, Westinghouse, and Combustion Engineering steam generator tube integrity have been ranked in the top twenty NRC generic issues.
These three tasks involve confirmatory tube integrity tests, studies of inservice inspection techniques, water chemistry analyses, and probability and consequence analyses.
The results of these programs, expected to be completed in late 1980, will provide bases for possible revisions to regulatory guides addressing inservice inspection and tube integrity criteria.
Task Action Plans A-3, A-4, and A-5 regarding steam generator tube integrity are attached as Appendix A.
For operating B&W plants, the NRC staff has concluded that continued operation does not constitute an undue risk to the health and safety of the public for the following reasons:
1.
Inservice inspection requirements and preventive tube plugging criteria have been established to provide assurance that degraded tubes will be identified and removed from service before leakage develops.
2.
Primary-to-secondary leakage rate limits, and associated surveillance requirements, have been established to provide assurance that the occurrence of tube cracking during operation will be detected and that appropriate corrective action will be taken expeditiously.
Early experience with leaks in B&W steam generators indicates that if a
. crack occurs it will propagate around the circumference of the tube in a short period of time (several hours) provided that there is flow of sufficient energy to drive it.
However, the crack formed will produce an identifiable leak, the detection of which will result in prompt corrective action (i.e., plant shutdown).
3.
The probability of the design basis accident occurring during normal operation is small, and the probability that the accident would occur during the short period of time between the leak detection and the plant shutdown is even smaller.
4.
Even if an accident occurs when there are cracked tubes, the conserva-tively calculated consequences are still acceptably small until the plant is shut down.
5.
A small amount of leakage (e.g., less than the Technical Specification limit) can be tolerated during normal operation without exceeding the offsite dose limits of 10 CFR Part 20.
22
5.
REFERENCES 1.
D. G. Eisenhut, B. D. Liaw, J. Strosnider, " Summary of Operating Experience With Recirculating Steam Generators." USNRC Report NUREG-0523, January 1979.
Available for purchase from the National Technical Information Service, Springfield, Virginia 22161.
2.
D. F. Levstek, W. Muller, F. J. Pocock, " Alloy 600 Experience in Nuclear Once-Through Steam Generators." Presented at the 16th CEFA Seminar, February 24-26, 1975.
Available in NRC PDR for inspection and copying for a fee.
3.
Minutes from the EPRI Corrosion Advisory Committee Meeting, Palo Alto, California, February 6-8, 1978.
Available in NRC PDR for inspection and copying for a fee.
4.
Oconee Nuclear Station Steam Generator Tube Leak Status Report, Dockets 50-269 and 50-270, August 26, 1977.
Available in NRC DPR for inspection and copying for a fee.
5.
Letter from W. O. Parker, Jr., Duke Power Company, to E. G. Case, NRC,
Subject:
" Steam Generator Operating History Questionnaire,"
February 16, 1978.
Available in NRC DDR for inspection and copying for a fee.
6.
Letter from W. O. Parker, Jr., Duke Power Company, to H. R. Denton, NRC,
Subject:
" Updated Steam Generator History Questionnaire,"
November 28, 1978. Available in NRC PDR for inspection and copying for a fee.
7.
Letter from D. H. Williams, Arkansas Power and Light Company, to K. R. Goller, NRC,
Subject:
" Arkansas Nuclear One, Units 1 and 2 Operating History Questionnairr," Dockets 50-313 and 50-368, June 15, 1978.
Available in NRC PDR fr e inspection and copying for a fee.
8.
Memorandum from V. S. Noonan, NRC, to R. W. Reid, transmitting
" Evaluation of Crystal Rive Unit 3 Steam Generator Damage Resulting from Burnable Poison Rod Atsembly Break-up," Docket 50-302, August 25, 1978.
Available in NRC PD( for inspection and copying for a fee.
4 9.
Letter from J. J. Mattimoe, Sacramento Municipal Utility District, to K. R. Goller, NRC,
Subject:
" Steam Generator Operating History Questionnaire," February 16, 1977.
Available in NRC PDR for inspection and copying for a fee.
10.
Letter from J. G. Herbein, Metropolitan Edison Company, to H. R. Denton, NRC,
Subject:
" Steam Generator Operating History Questionnaire,"
February 13, 1978.
Available in NRC PDR for inspection and copying for a fee.
11.
Letter from J. G. Herbein, Metropolitan Edison Company, to R. W. Reid, NRC,
Subject:
"Three Mile Island Nuclear Station Unit 1 Once-Through Steam Generator Inservice Inspection," July 28, 1978 Docket 50-289.
Available in NRC PDR for inspection and copying for a fee.
23
12.
J. M. Alzheimer, R. A. Clark, C. J. Morris, M. Vagins, " Steam Generator Tube Integrity Program, Phase I Report," USNRC Report NUREG/CR-0718, September 1979.
Available for purchase from the National Technical Information Service, Springfield, Virginia 22161.
13.
U.S. Nuclear Regulatory Commission, "NRC Programs for the Resolution of Generic Issues Related to Nuclear Power Plants, Report to Congress,"
USNRC Report NUREG-0410, January 1978.
Available for purchase from the National Technical Information Service, Springfield, Virginia 22161.
i 1
24 -
e c
7 a
f i
APPENDIX A TASK ACTION PLANS A-3, A-4, A-5 t
WESTINGHOUSE, COMBUSTION ENGINEERING, AND BABC0CK & WILC0X STEAM GENERATOR TUBE INTEGRITY.
t s
a 4
4 1
1 4 -
I h~
A 1
.. _,.. -,.. -,.,. +
TASK ACTION PLANS A-3, A-4, A-5 WESTINGHOUSE, COMBUSTION ENGINEERING, AND BABC0CK & WILCOX STEAM GENERATOR TUBE INTEGRITY Lead NRR Organization:
Division of Operating Reactors (DOR)
Task Manager:
Westinghouse (A-3):
J. Strosnider Combustion Engineering (A-4):
F. Almeter Babcock and Wilcox (A-5):
J. Strosnider Applicability:
Westinghouse, Combustion Engin:ering, and Babcock & Wilcox Pressurized Water Reactors Projected Completion Date:
May 1980 l
f
Tasks A-3, A-4, and A-5 R::v. No. 2 August 1979 1.
DESCRIPTION OF PROBLEM Pressurized water reactor steam generator tube integrity can be degraded by corrosion-induced wastage, cracking, reduction in 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 normal 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 have been decreased by changes in secondary water chemistry.
Degradation due to denting which leads to primary side s_ tress 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 (B&W) steam generators than in W or CE.
The most significant form of tube degradation in B&W generators has been cracks of unknown urigin 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 tube degradation described as an erosion-corrosion phenomenon has 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 tech-nical studies in the three areas of systems analyses, inservice inspection, and 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 behavior, ECCS performance, radiological con-sequences, and containment response.
The results will be used to define a tolerable level of steam generator tutie 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 defined by the systems analyses would occur in an accident.
The tube integrity portion of the tasks is primarily concerned with experimental verification of tube behavior during postulated accidents, development of tube plugging criteria, and definition of operating procedures and design to minimize tube degradation.
3pecific tasks which must be performed in each of the above areas include the following.
A.
Systems Analyses of TransSnts and Postulated Accidets (1) Analyses of LOCA with Concurrent Steam Generator Tube Failures LOCA analyses already performed will be reviewed to define a tolerable level of secondary to primary leakage through the steam A-3,4,5/1
Tcsks A-3, A-4, rnd A-5 Rev. No. 2 August 1979 generator based on fuel integrity, ECCS performance, and con-tainment response considerations. Additional calculations will be performed as needed.
(2) Analyses of MSLB with Concurrent Steam Generator Tube Failures MSLB analyses already performed will be reviewed to define a tolerable level of primary to secondary leakage through the steam generator during a steam line break.
For a break inside contain-ment, the tolerable primary to secondary leakage will be based on containment response considerations.
For a break outside containment, the tolerable primary to secondary leakage will be based on radiological considerations including fuel behavior during the accident.
Bounding calculations for the above scenarious may allow identifica-tion of the critical case, and detailed calculations to define the tolerable leak rate may only have to be performed for the governing conditions.
The results will be plant specific, depending on the plant site, ECCS design, and fuel duty, 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 evaluated as they relate to the Task Action Plan.
In addition, the results of industry and NRC experimental and analytical studies will be reviewed to evaluate the safety of continued operation of pressurized water reactors.
(2) Develop Statistically Based ISI Program Statistically based inservice inspection programs will depend 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 tube failures, since this parameter will be plant-specific.
Statistical analyses will also be performed to define the error associated with eddy current testing.
These results will be considered in the development of criteria defin-ing an acceptable ISI plan, to provide adequate assurance that no more than the tolerance number of tubes would fail during an accident.
(3) Evaluation of ISI Methods A qualitative review of the development of eddy current probes, coils, and multi-frequency. techniques to improve eddy current testing accuracy and to better quantify various defects includ-ing dents, cracks in dented regions, and circumferential fatigue cracks will be performed.
Results of this review will indicate A-3,4,5/2
Tasks A-3, A-4, and A-5 R:v. Ns. 2 August 1979 promising areas for further research in improving eddy current testing accuracy.
C.
Evaluation of Steam Generator Tube Integrity (1) Mechanical-Integrity The mechanical integrity of steam generator tubes under normal operating and postulated accident conditions (LOCA, SSE, and MSLB) will be reviewed and evaluated.
Specifically, experimentally obtained burst and collapse pressures will be reviewed and used to develop improved tube plugging criteria.
(2) Material Integrity Corrosion and other materials related to tube degradation phenomena will be reviewed as they relate to tube plugging criteria.
Particular emphasis will be placed on the tube denting i
phenomenon and the evaluation of field experience and laboratory data to define plugging criteria for dented tubes.
(3) Steam Generator Design and Operations Steam generator and secondary system design and operating procedures will be evaluated as they relate to mechanical and material aspects of steam generator tube integrity.
Specific emphasis will be placed on identifying improved design and operating procedures, to reduce the potential for denting and other corrosion-related phenomena in Westinghouse and Combustion Engineering steam generators, and fatigue cracking and erosion-corrosion in Babcock and Wilcox steam generators.
Specific areas of secondary system operation, materials selection, and mechanical design to be addressed include secondary water chemistry, main condenser integrity, potential for contaminant ingress from demineralizers, and potential for metal ion transfer from the main condenser and other secondary system equipment such as feedwater heaters.
D.
ISI Cost-Benefit Analyses Because the above studies could potentially result in increased inservice inspection requirements, cost-benefit analyses are 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 and downtime.
These costs will be weighed against the potential savings from eliminated unscheduled shutdowns to provide a realistic evaluation of the impact of new criteria.
A-3,4,5/3
Tcsks A-3, A-4, cnd'A-5 R:v. No. 2 August 1979 E.
Resulting Criteria Evaluation of the interaction between systems analyses, statistical analyses, and tube integrity analyses and integration of the above studies ~will allow the following criteria to be developed.
(1) Tube Plugging Criteria Using results of the mechanical and material tube integrity studies, tube plugging criteria for thinned, cracked, and dented tubes will be established.
These criteria will provide input for a revision of Regulatory Guide 1.121, " Bases for Plugging of Degraded PWR Steam Generator Tubes." Criteria for plugging thinned or cracked tubes will consider the minimum allowable wall thickness defined by the mechanical tube integrity studies and statistically based allowances for eddy current testing error and operational degradation between inspections.
Criteria for plugging dented tubes will include the incubation time for stress corrosion cracking in dented tubes based on the. magnitude and rate of strain in the tubes and the environment.
(2)
Inservice Inspection Criteria Results from the evaluations of ISI discussed above will be used to propose a revision of Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes."
(3) Steam Generator Design and Operating Criteria Based on the above tube integrity studies, licensee and vendor proposed modifications in steam generator design and operating techniques will be reviewed and 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 to pressurized water reactors.
For PWRs currently licensed for operation, we have concluded that, pending completion of these TAPS, continued operation does not constitute an undue risk to the health and safety.of the public for the following reasons:
Augmented inservice inspection requirements and preventative tube plugging criteriaLhave been established to provide assurance that a great majority of degraded tubes will be identified and removed from service before' leaks develop.
Primary to. secondary leakage rate limits, and associated surveillance requirements, have been established to provide assurance that the A-3,4,5/4
Tasks A-3, A-4, and A-5 Rev. No. 2 August 1979 occurrence of.abnor.9al tube degradation during operation will be detected and appropriate corrective action will be taken such that an individual defect will not become unstable under normal operat-ing, transient, or accident conditions.
Observed through-wall cracks at dented. locations, i.e., tubes / support plate intersections, have been small and stable (no rapid failures) during normal operation.
In addition, since such cracks are constrained by the support plates, they are not anticipated to become unstable (burst) during postulated accidents.
Continous feedback from operating experience and the TAP efforts will be utilized to update interim criteria and requirements.
For plants experiencing severe degradation, the following additional interim bases were also considered:
The probability of the design basis accident during normal operation is small and the probability that the accident would occur during the short period of time between the leak detection and the plant shutdown is even smaller.
A small amount of leakage (e.g., less than the Technical Specifica-tion limit) can be tolerated during normal operation without exceeding the offsite dosage limits of 10 CFR Part 20.
Some small amount of leakage can be tolerated during postulated accidents.
The above mentioned rationale, which constitutes the basis for continued operation of licensed PWR facilities, also supports continued licensing of new facilities.
Further, to the extent that is practicable, for I
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 degradation 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, and to facilitate sludge 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 improved corrosion resistance.
Material selection (chemistry), processing and heat treatment to mini-
~
mize the. susceptibility.of tubes to ' stress corrosion cracking.
i A-3,4,5/5
Tasks A-3..A-4, tnd A-5 R2v. No. 2 August 1979 Secondary system water chemistry control.
-Secondary side material selection (condensers, feedwater heaters, turbine discs and blades, elbows, etc.), and water cleanup system to minimize erosion and the resulting sludge and corrosion product buildup in the steam generators.
4.
NRR TECHNICAL ORGANIZATIONS INVOLVED A.
Engineering Branch, Division of Operating Reactors, has the primary lead responsibility for the overall review and evaluation of steam generator tube integrity.
This includes operational experiences,-_
tube failure mechanisms and potential repairs, plugging criteria, ISI requirements, tube failure probability, leakage rate limits, and secondary coolant system chemistry control.
Manpower Estimates:
.5 man year FY 1979 1.75 man years FY 1980 B.
Accident Analysis Branch, DSS, and the Environmental. Evaluation Branch, DOR, have the responsibility for the~ review and evaluation of the-offsite dosage during a postulated MSLB or LOCA with varying 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 responsible for the cost-benefit analyses described in Task 20.
This owrk will be performed under a technical assistance program with Battelle Pacific Northwest Laboratory.
Manpower Estimates:
Accident Analysis Branch
.10 ma~n year,FY 1979
.15 man year FY-1980 Environmental Evaluation Branch
.10 man year FY 1979
.10 man year FY 1980 C.
Reactor Safety Branch, Division of Operating Reactors, has the lead responsibility for the review and evaluation of the effects of
~
secondary to primary steam = generator tube leakage on ECCS and core performance during a LOCA and' defining a tolerable leak rate.
Manpower Estimate:
10 man years FY 1979 10 man years FY 1980 D.
Analysis Branch, Division of Systems Safety, has the lead responsibil-ity in-developing necessary analytical capabilities-and evaluating the response of the primary and. secondary systems to. a MSLB with varying numbers of concurrent steam generator tube failures.
This-A-3,4,5/6-i
,y w
t
=
Tasks A-3, A-4, and A-5 Rsv. N3. 2 August 1979 analysis will provide an evaluation of core response and primary to secondary leak rates during the postulated MSLB accident which can be used by the Accident Analysis and Environmental Evaluation Branches in their radiologicr.1 dose calculations.
Manpower Estimates:
.10 man years FY 1979
.10 man years FY 1980 E.
Mechanical Engineering and Materials Engineering Branches, Division of Systems Safety, have responsibility for the review of mechanical and material tube integrity and lead responsibility for the review of new-design, material, and operating methods.
The activities involved will include the review and evaluation of applicants' 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 inkeakage, ISI requirements, recommendation
]
for revision of Regulatory Guides and Sections 10.3.6 and 10.3.7 of the Stcndard Review Plan.
Manpower Estimates:
1 Mechanical Engineering Branch
.10 man year FY 1979
.25 man year FY 1980 Materials Engineering Branch -.10 man year FY 1979
.25 man year FY 1980 F.
Containment Systems. Branch, DSS, has the responsibility for the 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 man year FY 1979
.10 man year FY 1980 5.
TECHNICAL ASSISTANCE A.
Contractor:
Brookhaven National Laboratory (BNL) - D0R Funds Required:
$5K FY 1979, $5K FY 1980 This program is needed to obtain. technical consultation and assist-ar.ce to review information in areas of water chemistry and corrosion analysis, monitored by EB/ DOR.
This program will provide assistance in accomplishing Tasks C2 and C3.
A-3,4,5/7
Tasks A-3, A-4,'cnd A-5 Rev. No. 2 August 1979 B.
Contractor:
Brookhaven National Laboratory (BNL) - DSS Funds Required:
$125K FY 1979 Tne purpose of this program is to develop the necessary analytical capabilities 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 conditions.
This program will assist in accomplishing Task A2.
C.
Contractor:
Sandia Laboratories, D0R Funds Required:
$39K FY 1979, $10K FY 1980 The purpose of this program is to perform a statistical analysis of steam generator tube failures in operating reactors in order to estab-lish the bases for the sampling plan for inservice inspection.
This program will assist in accomplishing Tasks B1 and B2.
D.
Contractor:
Battelle Pacific Northwest La.boratory Funds Required:
$80K FY 1980 The purpose of this program is to perform the cost-benefit analyses described in Task 20.
6.
INTERACTIONS WITH OUTSIDE ORGANIZATIONS A.
Licensee (s) of Pressurized Water Operating Reactors All plants experiencing abnormal tube degradation will be closely monitored.
Each licensee with_ severely degraded steam generators will submit an analysis of the consequences of tube degradation on tube integrity and demonstrate that adequate safety margins exist for continued safe operation.
B.
Vendors The primary interaction with the vendors has been and continues to be on the investigation program for the resolution of the problems at operating plants, justifications for continued operation of plants with known tube degradations, and-the licensing bases for new plants.
C.
Interactions with other organizations such as the Electric Power Research Institute (EPRI) and the "ad hoc" organization of PWR owners may also be appropriate regarding the safe operation of steam generators in general and, in particular, the various safety problems associated with the degradation of steam generators.
A-3,4,5/8
Tasks A-3, A-4, and A-5 Rev. No. 2 August 1979 The primary purpose for interactions with these organizations is to exchange information on their experience and research work they are sponsoring.
7.
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 experi-mental 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 is managed by Metallurgy and Materials Branch (Task C1).
RES has also funded, at the request of NRR, a program to address the factors which determine Inconel 600 susceptibility to stress corrosion cracking in primary water.
Metallurgical condition. chemistry, temperature, stress and environment will be considered (Task C2).
B.
Office of Standards Development, Division of Engineering Standards, Structures and Components Standards Branch.
OSD has funded a confirmatory research program at Battelle Columbus Laboratory to evaluate eddy current methods for inspecting steam generator tubes as a subcontract to Brookhaven National Laboratory (part of Task B3).
l C.
Office of Management and Program Analysis, Applied Statistics 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 by the ACRS and, accordingly, will be coordinated with the committee as the task progresses.
I 8.
POTENTIAL PROBLEMS Except for steam generator replacement, there is no apparent short-term resolution of tube denting in affected Westinghouse or CE plants.
Programs to resolve tube denting in presently operating plants may bring about a-partial solution, by arresting denting 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 be minimized t. the point where continued operation can be justified.
Unfortunately, results of the BNL stress corrosion cracking program sponsored by the Office of Research may not be available before the desired Task Action Plan completion dates; therefore, plugging criteria A-3,4,5/9
Ttsks A-3, A-4, and A-5 Rav. N2. 2 August.1979 will necessarily be based on-preliminary results of the program and largely on operating experience.
The criteria may therefore be somewhat judgmental in nature.
e 4
4 l
- i A-3,4,5/10
a
-*c.-
O APPENDIX B RANCHO SECO SUPPLEMENT TO MONTHLY PLANT OPERATIONS REPORT EDDY CURRENT INSPECTION OF OTSG TUBES 1978 REFUELING OUTAGE f
e i
APPENDIX B RANCHO SECO SUPPLEMENT TO MONTHLY PLANT OPERATIONS REPORT EDDY CURRENT INSPECTION OF OTSG TUBES 1978 REFUELING OUTAGE DESCRIPTION A-0TSG B-0TSG No. of Tubes Inspected Bundle-Lane:
3.5% = 540 Tubes Bundle-Lane:
3.2% = 501 Tubes Lana:
- 100%
= 373 Tubes Lane:
s100% = 363 Tubes Total' UU Tubes Total BT4 Ti+w Na. of Tubes Plugged 0
0 Location of Indications Dents (1 to 3 mils) at Upper Dents (1 to 3 inils) at Upper Tube Sheet in $54% of the Lane Tube Sheet in s45% of the Lane Tubes (205 tubes).
Tubes (167 tubes).
Dents (5 to 6 mils) at 15th Dents (5 to 6 mils) at the Tube Support Plate in 7 tubes 15th Tube Support Plate in of tube blindle (excluding the 5 tubes of the tube bundle lane area).
(excluding lane area).
Denting indications at Tube Denting indications at Tube Support Plate were observed Plate were observed during during the previous inspection the previous inspection but but were so small as to be were so small as to be interpreted as permeability interpreted as permeability variation.
Growth attributed variation.
Growth attributed to vibration.
to vibration.
Tube 66-130 was inspected (per Other indications noted are SMUD letter of 10/8/77 to as follows:
Region V) with no indications noted.
& #2 TSP Tube 58/87 <20% OD between LTS
& 1 TSP Tube 53/86 <20% CD between LTS i
& 1 TSP Tube 60/43 38% OD @ 15th TSP (No growth from indication observed last year and noted on NCR S-768 Rev. 1)
B *.
e
APPENDIX B (Continued)
DESCRIPTION A-0TSG B-0TSG Location of Plugged Tubes NA NA G:neral Description of
$100% of Lane (3 rows either Same as "A" whsre inspected tubes side of the open tube lane to are located the WY Axis) and $3% radom sample of the remainder of the entire tuba bundle.
Typt of Inspection All tubes; 100% of length at Same as "A" 400 KHz (some lane tubes were inspected with 400 KHz and 25 KHz simultaneously.
1 L
e B-2
g
)
U.S. NUCLEA') REOUL ATORY COMMtS840N
- 1. REPO.~iT NUM8E R (Assisted by DOC 1 BIBLIOGRAPHIC DATA SHEET NUREG-0571
- 4. TITLE AND SUBTITLE (AcW Vodume Na,if apprcpriesel
- 2. (Leave b/mk1 Summary of Tube Integrity Operating Experience with
- 3. RECIPlENT'S ACCESSlW NO.
Once-Through Steam Generators
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLEMD B. D. Liaw, J. Strosnider Miy" I"*"1979
- 9. PERFORMING ORGANIZATION NAME AND MAILING ADORESS (/nclude lip Code /
DATE REPORT ISSUED Division of Operating Reactors l "^"1980 garch U.S. Nuclear Regulatory Commission Washington, D.C.
20555 8' *"' "" * #
- 8. (Lehre Nmkl
- 12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (/nclude lip Codel
- 10. PROJECT / TASK / WORK UNIT NO.
Same as 9. above.
- 11. CONTRACT NO.
- 13. TYPE OF REPORT PE RIOD COVE RED (/nclussre dares)
Technical Report
- 15. SUPPLEMENTARY NOTES
- 14. (Leave b/m*/
- 16. A85TR ACT Q00 words'or less) l Opsrating problems have occurred in the steam generators designed by each of the three manufacturers of pressurized water reactor nuclear steam supply systems:
Babcock & Wilcox, Combustion Engineering, and Westinghouse Electric Corporation.
This report focuses on the problems associated with steam generators of the once-through type that are designed and manufactured by Babcock & Wilcox.
It identifies the operational experiences observed to date and the position the NRC staff has taken to ensure steam generator tube integrity.
It should be noted that a number of research efforts related to these problems are currently under way, and that the 1
information included in this report represents the staff's current understanding of cach issue. A similar report, NUREG-0523, discusses tube integrity experience with i
the U-tube type of steam generator designed by Westinghouse and Combustion Engineering.
- 17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS 17b. IDENTIFIERS /OPEN-ENOED TERMS
- 18. AVAILABILITY STATEMENT
- 19. SECURITY CLASS (This report /
- 21. NO. OF PAGES Unlimited availability.
- 20. SECURITY CLASS (This papel
- 22. PRICE S
NRC FORM 336 17 77)
_