ML20059K666
| ML20059K666 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/17/1990 |
| From: | Wallace E TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEB-89-001, IEB-89-1, NUDOCS 9009240140 | |
| Download: ML20059K666 (15) | |
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TENNESSEE VALLEY AUTHORITY CH ATTANOOG A, TENNESSEE 37 dol SN 157B Lookout Place SEP 171990 l
l U.S. Nuclear Regulatory Commission ATTN Document Control Dest Washington, D.C.
20555 Gentlement In the Matter of
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Docket Nos. 50-327 j
Tennessee Valley. Authority
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i SEQUOYAH NUCLEAR PLANT (SQN) - UNIT 1 JUSTIFICATION FOR CONTINUED OPERATION (JCO) - NRC BULLETIN 89-01 j
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Reference:
TVA letter to NRC dated June 12, 1989 "NRC Bulletin 89-01:
Failure of Westinghouse Steam Generator Tube Mechanical Plugs" As identified to NRC in an August 31, 1990, telephone call, a small circumferential crack was found in one of five Westinghouse steam generator tube plugs from heat No. NX5222 removed from SQN Unit 1 during the Cycle 4 refueling outage.
In addition, one plug from heat No. h75222 experienced axial cracks below the expander. Ten plugs from this,same heat remain in service in SQN Unit 1 Steam Generator 1 hot legs. As requested by NRC in the August 31, 1990, telephone call with TVA and Westinghouse representatives, enclosed is an evaluation that provides details of the plug cracks and a JC0 until 1993, which is consistent with the consnitment contained in the referenced letter. This evaluation is based primarily on the JC0 provided in Section 3.0 of WCAP-12244, Revisi)n 3, " Steam Generator Tube Plug Integrity
- Sumirary Report."
No new TVA commitments are contained in this letter.
Plea e direct questions concerning this issue to Kathy S. Whitaker at (615) 843-7748.
I Very truly yours, TENNESSEE VALLEY AUTHORITY w
E. G. Wallace Manager Nuclear Licensing and Regulatory Affairs Enclosures cet See page 2 i
90092a0140 900917 f-
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PDC An Equal Opportunity Employer Il l
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,c U.S.-Nuclear Regulatory Commission f!(f jf jg{gg t
i ec (Enclosures):
Ms. S. C. Black, Deputy Director Project Directorate 11-4 U.S. Nuclear Regulatory Commission 3
One White Flint, North 11555 Rockville Pike Rockville, Mary 1snd 20852 i
Mr. J. N. Donohew Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike i
Rockville, Maryland 20852 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee ?7379 Mr. B. A. Wilson, Project Chief U.S. Nuclear Regulatory Commission Region II i
101 Marietta Street, NW, Suite 2900
- Atlanta, Georgia 30323 4
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JUSTIFICATION FOR CONTINUED OPERATION
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Westinghouse Energy Systems Nuclear and Myanced Electric Corooration
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Box 355 httsburgh Pennsylvania 15230-0355
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i Mr. D. F. Goetcheus TVA 90 989 Manager Steam Generator /NSSS Dept.
NS-0PLS-OPL-II-90 600 Tennessee Valley Authority September 7, 1990-P. O. Box 2000 Soddy Daisy, TN 37379 i
Tennessee Valley Authority Sequoyah Nuclear Plants Unit 1 STEAM GENERATOR TUBE PLUGS PWSCC JUSTIFICATION FOR CONTINUED OPERATION
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Dear Mr. Goetcheus:
In accordance with our discussions, attached is a Justification For Continued Operation for Steam Generator Tube Plugs which may txperience primary water stress corrosion cracking (PWSCC) for Unit 1 operation.
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If you have any questions, please do not hesitate to contact us.
Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION k*
B. J. Garry, Manager TVA Sequoyah Project:
i Customer Projects Department LVT/ee cc:
D. M. Lafever R. G. Davis P. G. Trudel W. R. Smith i
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Attachment To TVA-90-989 i
Page 1 of 11 JUSTIFICATION FOR CONTINUED OPERATION i
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SEQUOYAH UNIT 1 FOR l
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EXPERIENCE PWSCC
1.0 INTRODUCTION
In response to NRC Bulletin 89-01, seventy-nine Westinghouse Alloy 600 mechanical plugs have been removed from the Tennessee Valley Authority Sequoyah Unit I steam generators. Out of the 79 plugs removed, a y
total of 21 plugs (14 hot leg, 7 cold leg) have been examined both i
, visually and using-a 40X microscope.
The plugs examined were from heats NX4523 (9 Hot Leg / 7 Cold Leg) and NX5222 (5 Hot Leg). Of the hot leg plugs examined, one very small circumferential crack was detected above the expander in a plug manufactured from Heat NX5222. This circumferential crack consisted of two small cracks where the total circumferential extent was 0.095 inch or approximately 15 degrees around the circumference with a maximum depth of penetration of 0.009 inch.
In addition to the circumferential crack above the expander, one Heat NX5222 plug experienced axial cracks below the expander. Axial cracks were also
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i Attachment To TVA-90-989 i
Page 2 of 11 detected below the expander in 7 of the 9 hot leg plugs manufactured from Heat NX4523; six of these plugs had through wall axial cracks.
Small circumferential cracking below the expander was also noted in 3 of the 7 plugs which had axial cracks.
No cracking was detgeted in any of the cold leg plugs.
All hot leg plugs manufactured from Heat NX4b23 have been replaced with Alloy 690 plugs.
However, ten hot leg plugs manufactured from Heat NX5222 remain in service in Steam Generator 1 of Sequoyah Unit 1.
As a consequence, a safety evaluation has been completed to establish the justification for continued plant operatjon with steam generator tube plugs which may have a propensity to experience PWSCC.
This evaluation is specific to Sequoyah and is largely based on the Justification for Continued Operation provided as Section 3.0 of WCAP-12244, Rev. 3, " Steam Generator Tube Plug Integrity Summary Report".
This report was prepared to document analyses, evaluation, testing, and metallurgical evaluations used by the Westinghouse Electric Corporation to assess the safety significance of primary water stress corrosion cracking in steam ~ generator tube plugs manufactured from Alloy 600 material.
In addition to the five plugs of heat NX5222 removed from Sequoyah Unit 1, three plugs from this same heat were removed from another utility's plant and examined by Westinghouse. These plugs, having PWSCC usage factors of 0.12, did not exhibit any cracks.
(As defined in WCAP-12244, the PWSCC usage factor is a specific plug's normalized EFPD divided by the normalized EFPD for the plug removed from Hillstone Unit 2 with the highest circumferential crack growth rate.
'When the PWSCC usage factor is 1.0, the remaining EFPD to minimum ligament calculated by the algorithm will be zero). The five Sequoyah Unit 1 plugs had PWSCC usage factors of 0.21.
Heat NX5222 has been characterized as one with a semi-continuous grain boundary carbide distribution and was assigned a microstructural factor of 4.4 based on
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1 Attachment To TVA-90-989 Page 3 of 11 the corrosion test results contained in WCAP 12244.
In accordance j
with the program specified in WCAP-12519, which was issued in February 1990 to the NRC, the algorithm would be modified as follows:
1.
Since a circumferential crack has been found above the expander for heat NX5222, the microstructural factor for heat NX5222 will be set at 1.0.
2.
The apparent.rowth rate for heat Nx5222 will be established based on the most conservative value obtained from metallurgical examination results for plugs from that heat.
..e Accordingly, the apparent growth rate normalized to 622.5 degreas Fahrenheit is 0.0336 mils /EFPD which compares to the normalized apparent i
growth rate for the Millstone plug of 0.1679 mils /EFPD. The combined effect of these two changes, lowering the microstructural factor but using 1
an actual apparent growth rate specific to heat NX5222, will not appreciably alter the service times to reach minimum ligament for this heat contained in WCAP 12244, Rev. 3.
In view of this small change, there is no expected effect on the utilities having plugs from this heat.
The conservatism employed in the growth rate calculation was to assume that it had grown 360 degrees around the circumference to the depth measured when, in fact, it had only grown 15 degrees.
For the NX5222 plugs remaining in Sequoyah Unit 1, repair actions'are not required until 1993.
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Attachment To TVA 90-989 1
P1ge 4 of 11 t
1 2.0 EVALUATION The root cause of the steam generator primtry to secondary leakage event' j
at North Anna Unit 1 in March of 1989 was a steam generator tubg plug top release as a result of the occurrence of circumferential1y oriented PWSCC r
above the top of the mechanical plug expander.
Laboratory test results and metallurgical examination of plugs removed from service that were believed to be dripping / leaking support the determination that the motphology of PWSCC in mechanical plugs is predominantly axial and is limited in azimuthal extent when the crack passes through wall. As the i
visual examination of plugs manufactured from Heat NX5222,.which were removed from service has revealed the presence of PWSCC,,the Westinghouse designed plugs manufactured from Alloy 600 material remaining in service j
of the same heat'in Steam Generator 1 are evaluated below for the potential for plug top release.
2.1 Leakage Past The Plug Precludes Piug Top Release 4
i The following conditions must occur in order for the top of a Westinghouse
- mechanical plug to release:
1.
A 3600 circumferential crack must proceed through the thickness of the plug shell without penetrating through-wall until a ligament remains insufficient to resist primary pressure.
2.
A plug land above the circumferential crack must seal and not allow leakage behind the plug to equalize pressure across the plug.
3.
The inactive tube must not have a through-wall penetration when it was L
removed from service.
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Attachment To TVA-90-989 Page 5 of 11 Nine of the 10 Heat NX5222 plugs are installed in Row 1 tubes which have experienced PHSCC in the U-bend of the tube. One Heat NX5222 plug is installed in a Row 10 tube; this plug has a cable sicbilizer, which extends around the U-bend to the cold leg straight portion of the tube installed on top of the plug. This tube was preventively plugged and stabilized as it was identified as potentially being susceptible to fatigue-induced cracking of the type experienced at North Anna Unit 1.
Plug top release is not expected to occur in any tubes where through wall cracks are present. The potential for plug top release in tubes where either partial through wall cracks were present or no degr,adation was
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present when the tube was removed from service has also been evaluated.
Typically, in additian to the lands beneath the expander, only the first land immediately above the expander seals in the mechanical plug.
The location of circumferential cracking, as it has occurred in the field, is f
limited to between the first and second lands immediately above the expander. The second land above the expander does not normally provide a seal. Mechanical plugs with either axial or circumferential cracks at this location are expected to leak past the seal and into the inactive tube precluding plug top release.
l 2.2 Plug Top Release For Transients For Westinghouse designed steam ~ gene;ators, plug to tube interface force l
calculations for plugs located ncar the tubesheet center, as is the case l
with the nine Row 1 tubes and one Row 10 tube with NX5222 plugs, show that negative contact forces result for a number of normal and upset operating conditions due to tubesheet bow.
However, the tubesheet bow that occurs in bringing a plant to hot standby exceeds the tubesheet bow that occurs during most secondary side transients including the transient which occurred at North Anna Unit 1 (i.e., the main feedwater regulator valve
Attachment To TVA-90-989 Page 6 of Il failed closed due to failure of an air supply line) which may have resulted in the plug top release event. The expe
.ed frequency of these normal plant operation events greatly exceeds thL of the accident events.
In the unlikely event that favorable conditions exist for any plug which would permit the plug top to be released, it is more likely to occur during expected normal operating conditions than as a consequence of a significant transient or accident.
For plugs located in the center of the bundle, forces on the plug land are increased relative to steady state conditions for accident condition loadings (i.e., Steam Line Break Event, Stuck Open Safety. Valves, etc.)
and plug top release during these events is not expected,to occur.
The.:e considerations, along with the inherent variability in PWSCC cracking rates lead to the conclusion that there is a very low likelihood that multiple plugs could exist in a condition which would be capable of resulting in plug top release.
2.3 Single Plug Top Release Event Relative to SGTR Analysis in the FSAR For tubes which are expanded the full depth of the tubesheet, such as is the case with the WEXTEX expanded tubes in the Sequoyah Unit ', steam generators, a separated plug top has approximately 20 inche; of travel before encountering the transition to the unexpanded portion of the tube which will slow the plug top down.
For the plug top to move into the unexpanded portion of the tube it must have sufficient force to deform the plug lands to a diameter where they can more easily pass through the tube. The North Anna Unit 1 event, laboratory testing, and dynamic analysis show that sufficient energy can be achieved by the separated plug top in full depth expanded tubes to pass through all energy barriers inside the tube leading to perforation of the inactive tube.
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l Attachment To TVA-90-989 Page 7 of 11 However, should perforation of the inactive tube occur, the mechanical expander functions as a leak limiting orifice such that the maximum leak rate expected is less than 80 gpm for all sizes of mechanical plugs.
This leakage is less than the design basis tube rupture. More than seven tubes would have to experience the plug top release mechanism simultaneously before primary to secondary flow would exceed that analyzed in the FSAR.
l In order for multiple plugs to be released in the generator at the same time, all plugs must have the same remaining ligament and no leakage past the plug into the inactive tube. In light of the fact the total population of plugs manufactured from Heat NX5'!22 remaining in the Sequoyah Unit 1 steam generators is 10 plugs, this scenario of multiple plug top releases is judged to be highly unlikely.
Additionally, an adjacent tube rupture event is '10t expeeted to occur as a released plug top is calculated not to contain enough energy to perforate the tube walls of both the inactive tube and adjacent active tube o-tubes.
For steam generators with full depth expanded tubes, analysis shows that the highest plug top kinetic energies are achieved only with substantial flow through the expander region'of the plug.
Testing up to a level of approximately 1100 ft-lbs has not resulted in the perforation of an adjacent active tube.
Tests conducted at 800, 960, and 1173 ft-lbs resulted in the perforation of the inactive tube, and lodging of the plug top back into the inactive tube.
This is similar to what occurred at North Anna Unit 1.
Testing has also been conducted on tube plugs where stabilizers have been-installed on top of the mechanical plug.
It is judged that a plug top release in the configuration of the Row 10 tube in steam generator 1 would not result in perforation of the inactive tube either by the stabilizer or the plug top.
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Attachment To TVA 90-989 i
Page 8 of 11 l
2.4 Procedure Response to SGTR and Plug Top Release Discussions j
The Westinghouse Owners Group (WOG) was originally formed to respond to the post-TMI actions required by the NRC.
One of the major contributions of the WOG has been the development of emergency response guidelines,
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ERGS, for Westinghouse designed nuclear steam supply systems.
The ERGS j
are entered any time a reactor trip or safety injection occurs or is j
required, or any time a complete loss of all AC power or the AC emergency busses takes place, i
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Since a reactor trip is associated with a steam generator,. tube rupture event or leak sufficiently large to cause a high radiation alarm on the i
condenser air ejector or secondary side, the ERGS are used to bring the plant to a safe shutdown condition.
Specific ERGS have been developed to address.a number of SGTR situations, including multiple tube ruptures and an SGTR coincident with a LOCA. These guidelines have been developed, in part, based on the history of actual plant performance to a SGTR. An integral part of the ERG development has been the validation of the procedures on plant simulators. A total of 16 different scenarios were used in the validation process, including the rupture of more than one tube.
l The WOG has conducted several workshops to provide training on the ERGS, including one specifically directed to SGTR. A part of a reactor L
operator's initial and recurring training is to address the plant response
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to a number of these different SGTR scenarios.
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i Attachment To TVA-90 989 Page 9 of 11 In the unlikely event that a plug top release event should occur at Sequoyah Unit 1, the expected consequence would be a primary to secondary leak well within the bounds of guidance, training, and validation provided by the ERGS. The release of multiple plug tops would increase the primary to secondary flow rate, but would be enveloped by the ERG validation case of the rupture of multiple tubes.
The conclusion from an extensive review of plug top release scenarios is that the operator would experience no differences in plant response from that of the validated ERG SGTR cases.
2.5 Realistic Radiological Release Considerations
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The consequences of a plug top release coincident with a steam generator tube rupture have been evaluated.
Based on realistic assumptions, the doses resulting from the release of a plug top consisttnt with a SGTR are sig'nificantly less than the results reported in the SAR for a SGTR using standard licensing methodology.
These reductions in releases are due to lower source terms, on-site power, improved dose conversion factors, and meteorology.
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Attachment To TVA-90 989 Page 10 of 11
3.0 CONCLUSION
' A total of ten Westinghouse designed plugs manufactured from Alloy 600 material (Heat NX5222) remain in service in Steam Generator 1 of Sequoyah Unit 1.
These plugs have been evaluated to detarmine the potential for i
steam generator tube plug top release.
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It is concluded that should these 10 plugs experience either axial or circumferential cracks at a location above the expander, it is expected
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that the presence of a through wall crack in the plug would result in a l
leak past the seal and into the inactive tube precluding nlug top release. The consequences of a potential plug top release is further j
minimized in the plug installed in the Row 10 tube by the presence of the cable stabilizer.
In the unlikely event that a plug top should release, it is more likely to occur during normal operating conditions than as a l
consequence of a significant transient or accident especially since the l
identified plugs are located at the center of the tube bundle.
L Since a reactor trip is associated with a steam generator tube rupture event or leak sufficiently large enough to cause a high radiation alarm on the condenser air ejector, the plant Emergency Res a nse Guidelines would be useo to bring the plant to a safe shutdown condition.
The expected l
consequence would be a primary to secondary leak well within the bounds of guidance, training, and validation provided by the ERGS.
- Finally, realistic radiological release considerations should be significantly less than results included in the SAR for a SGTR event.
In light of the above, Tennessee Valley Authority's Sequoyah Unit 1 can continue to operate safely with the ten steam generator tube plugs manufactured from Heat NX5222 material remaining in service in Steam Generator 1 until 1993.
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4.0 REFERENCES
1.
90 5M45-SEQYA M1, " Examination of TVA Sequoyah Unit 1 Steam, Generator Tube Plugs-Phase I", August 14, 1990.
2.
90 5M45-SEQYA M2, " Examination of TVA Sequoyah Unit 1 Steam Generator Tube Plugs-Phase II", August 14, 1990.
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