ML20057B932

Summary of 930727 Meeting W/Util & Westinghouse Re Results Analysis of Three Tubes Pulled Form Unit 1 B SG During Last RF7.List of Meeting Attendees & Meeting Handouts Encl

ML20057B932
Person / Time
Site:	Summer
Issue date:	09/20/1993
From:	George Wunder Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9309240164
Download: ML20057B932 (22)

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,p W ASHINGTON. D.C. 205s5-0001 September 20, 1993 Docket No. 50-395 f i

LICENSEE: South Carolina Electric & Gas Company '

FACILITY: Virgil C. Summer Nuclear Station, Unit No.1

SUMMARY

OF JULY 27,1993, MEETING l

On July 27, 1993, the staff met with senior management of South Carolina Electric & Gas Company (SCE&G) and representatives of Westinghouse Electric ,

Corporation (Westinghouse) to discuss the results of the analysis of three tubes pulled from the Virgil C. Summer Nuclear Station, Unit No.1 (Summer Station) "B" steam generator during the last refueling outage 7 (RF7). ,

Leak and burst tests, crack morphology, causative mechanisms, steam generator chemistry, and remedial actions were addressed.

Five tubes indicated voltage growth rates greater than 2.5 volts. The three pulled tubes had the largest crack growth indications of the five tubes with voltage growth rates greater than the typical 2.5 volts expected during the cycle based on eddy current voltage amplitude measurements of 9 to 21. volts.

Two of the three tubes had burst strengths that were consistent with expected values from burst / crack length correlations. The third tube's burst strength was low. Based on metalographic evidence of the lack of tearing at the crack tip, presumably due to damage from the tube pulling operation and leak testing, the licensee considered this test not to be valid.

The licensee stated that, except for the five large eddy current indications that were all located at the O1H flow distribution baffle (FDB) support plate i location, voltage growth rates were consistent with previous cycle data and ,

with rates found in other domestic plants. The licensee's analysis of the  ;

crack surfaces showed evidence of copper and caustic chemistry. In addition, deposits on the surface of the tube indicated a packed crevice at the flow  :

distribution baffle (FDB) location and low frequency eddy current testing .

indicated tube offset misalignment and likely tube contact that would have  :

provided crevice corrosion conditions. The licensee's analysis concluded that the' combination of these factor.s was responsible for the rapid crack growth at this location and that these factors would not be present elsewhere in the steam generator.  !

l SCE&G pointed out that during RF7 tubes with any indication whatsoever were j plugged; this is more conservative than the criteria used during RF6. They

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also stressed. that they had made improvements in steam generator chemistry l control that would tend to mitigate the corrosion-inducing conditions that '

exist at the intersection of the U-tubes with the FDBs and the tube support plates. Finally, SCE&G committed to instituting administrative procedures to reduce the allowable leakage limit to 150 gallons per day (gpd), increasing operator training on steam generator transients, reducing allowable coolant activity to half of what is allowed in the Technical Specifications, and ensuring that one power-operated relief valve is unblocked at all times.

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'. 1 September 20, 1993 l The licensee reported that, according to its analysis, tube burst capability satisfied Regulatory Guide 1.121 (taking credit for FDB restraint) and calculated potential steam line break (SLB) leakage was within 10 CFR Part 100 limits. Based on the improved inspection and chemistry control, as well as on the remedial actions being instituted, SCE&G maintained that a mid-cycle inspection will not be necessary to ensure safe operation of Summer Station ,

for cycle 8.

A copy of the attendance list is attached as Enclosure 1. A non-proprietary copy of the meeting handout is attached as Enclosure 2.

ORIGINAL SIGNED BY:

l George F. Wunder, Project Manager Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1

1. Attendance list 1

2. Handout cc w/ enclosures:

See next page  !

DISTRIBUTION:

cc w/ enclosures 1 and 2: I Docket. File .

NRC/ Local PDRs .

PD II-l Reading File  !

G. F. Wunder E. W. Merschoff, R-II cc w/ enclosure 1 T. Murley/F. Miraglia 0GC .

J. Partlow E. Jordan  !

S. Varga J. Partlow G. Lainas J. Strosnider S. Bajwa ACRS (10)

P. Anderson L. Plisco, EDO 17-G-21 T. Reed T. Farnholtz .

K. Karwoski E. Hackett i

0FFICE LA:PUNID5PE PM;p[2h0kPE DE:EMCB* AD:PD21:DRPE NAME PAnbr n Gddr:dt JStrosnider SBajwa Mb ,

DATE

/ /93 N/ f(/ /93 08/19/93 b/ M /93 Document Name: SUM 930727. MIG i

i The licensee reported that, according to its analysis, tube burst capability satisfied Regulatory Guide 1.121 (taking credit for FDB restraint) and calculated potential steam line break (SLB) leakage was within 10 CFR Part 100 limits. Based on the improved inspection and chemistry control, as well as on the remedial actions being instituted, SCE&G maintained that a mid-cycle inspection will not be necessary to ensure safe operation of Summer Station for cycle 8.

A copy of the attendance list is attached as Enclosure 1. A non-proprietary copy of the meeting handout is attached as Enclosure 2.

[ f4 George F. Wunder, Project Manager Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1. Attendance list

2. Handout cc w/ enclosures:

See next page O

p.,

Virgil C. Summer Nuclear Station cc:

Mr. R. J. White Nuclear Coordinator S.C. Public Service Authority c/o Virgil C. Summer Nuclear Station Post Office Box 88, Mail Code 802 Jenkinsville, South Carolina 29065 J. B. Knotts, Jr., Esquire Winston-& Strawn Law Firm 1400 L Street, N.W.

Washington, D. C. 20005-3502 Resident Inspector / Summer NPS c/o U.S. Nuclear Regulatory Commission Route 1, Box 64 Jenkinsville, South Carolina 29065 Regional Administrator, Regi,on II U.S. l4uclear Regulatory Commission 101 Marietta St., N.W., Ste. 2900 Atlanta, Georgia 30323 Chairman,_Fairfield County Council Drawer 60 Winnsboro, South Carolina 29180 Mr. Heyward G. Shealy, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 Mr. R. M. Fowlkes, Manager Nuclear Licensing & Operating Experience South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station Post Office Box 88 Jenkinsville, South Carolina 29065 Mr. John L. Skolds, Vice President Nuclear Operations South Carolina flectric & Gas Company Virgil C. Summer Nuclear Station Post Office Box 88' Jenkinsville, South Carolina 29065

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Enclosure 1 ,

ATTENDANCE LIST Staff SCE&G Other J. Partlow J. Skolds T. Pitterle (Westinghouse)

G. Lainas R. Clary J. Barkich (Westinghouse)

J. Strosnider R. Kelso B. Cullen (Westinghouse)

S. Bajwa J. Frick D. Mayes (Duke Power)

M. Caruso R. White (SCPSA) L. Z1rr (STS)

E. Murphy J. Hopenfeld T. Reed G. Wunder T. Farnholtz K. Karwoski E. Hackett l

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ENCLOSURE 2 '!

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l South Carolina Electric & Gas Company V. C. Summer Nuclear Station  ;

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NRC / SCE&G / Westinghouse  ;

Meeting  ;

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Steam Generator Pulled Tubes l

l Examination Results N / July 27,1993

V C Summer Nuclear Station

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, , Agenda' ..

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s w , .,3 Introduction & Purpose Jack Skolds Recap of April 27 Meeting Ron Clary Update of NDE Results Rollin Kelso Overview of Evaluation & John Frick Remedial Actions Detailed Presentations Pulled Tube Examination Summary Tom Pitterle  !

Leak & Burst Tests Morphology l

Causative Mechanisms John Barkich j l

Mechanisms '

Comparisons with Another Plant Corrective Action Tube Integrity Analysis Tom Pitterle i

Burst Margins SLB Leakage Assessment Remedial Actions Summary Jack Skolds Discussion All

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- Purp.ose of Meeting. . -~oM9m un ' . ag  ;

h Inform NRC of: -

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9 Evaluations performed on Pulled Tubes  !

i 9 Results of Leak & Burst Tests l

9 Identification of Causative Fa tors  ;

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I 9 Tube Integrity Analysis i

9 Remedial Measures

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• gg? -cqRecap vgp;;%yg7 of April 27 Meeting with NRC R1,ff:s w

j e VCS Steam Generators Inspection Techniques & Methodology RF-6 Results RF-7 Inspection Plan & Results l

1 9 Review of Cycle 7 Operation 9 Tube Pulls Reason / Tube Selection Basis Tube Pull Process )

Examinations / Schedule Preliminary Findings e Cycle 8 Operation Basis for Start-up/ Initial Operation Future Activities l

1 9 Steam Generators to be Replaced in Fall 1994 i

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sr ,, Key Points of April 27 Meeting -2 gig.

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i e There were Inspection Improvements Between RF-6 and RF-7 l

1) Data Analyzed per TSP-IPC WCAP #13522

2) Probe Wear and IPC Transfer Standards used in RF-7

3) Industry Events Between RF-6 and RF-7 Raised Analyst Sensitivity for Recognition of TSP Flaws e No Flawlike Indications Left in Service Regardless of Indicated Depth l

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• We were Searching for Causative Factor (s) l

1) No Correlation Found Between Plant Operating Parameters and the 5 Atypical Voltage Indications

2) Pulled Tube Testing / Analysis Required to Determine Cause e Tubes Damaged During Tube Pull Process

- Expected Results were Conservative Compared to Insitu Tube Behavior

1. Population of Atypical Voltage Indications was Small - 5 Indications out of 99,592 Hot Leg Intersections

• Interim Operating Period of 111 EFPD Based I on Conservative Safety Evaluation

1. We informed NRC we Would Return to Discuss Results of the Pulled Tube Testing

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D-3 S/G AV 2 AV3 Support Number AVI - - - AV4 Scheme 7  ;

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12H . _ . . _ _ _ _ . . . . _ ___ _ 12C l

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l Refuel 7 Results  !

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.;, khe 5 Indications Exceeded 3.7 Volts Out of 12,449 Tubes or 99,592 Hot Leg TSP /FDB Intersections l

S/G Row Col  ; Volts ! Location i

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I I B 42 43 22.32 1 01H + 0.00 l ]

1  : i B 28 41 11.59 01H + 0.00 l [

4 B 33 l 20 l 9.84 ; 01H + 0.00

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B 31 45 7.72 01H + 0.00 i

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45 6.02 01H + 0.00 l OlH is the Flow Distribution Baf fle Plate  !

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. . . m.mu+mndummm.e4Lscu N El#!%s ECT Probe Uncertainty ,

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- All 5 Atypical Large Voltage Indications '

Found During RF-7 were Inspected with the Same Probe During RF-6 l

- Review of RF-6 Data Indicates that Voltage Response of this Probe was Deficient with Respect to the Requirements in WCAP-13522

- Actual Voltage Growth less Than Indicated During Cycle 7

- Implementation of Probe Wear Standard Prevented Reoccurrence in RF-7

V C Summer Nuclear Station Pulled .,.. Tubes FDB Region NDE Comparison, R28 C41 R33 C20 R42 C43 '

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Bobbin: 11.6 V Bobbin: 9.8 V  ! Bobbin: 22.3 V 550/130 550/130 550/130 i Field 93% 88% 91% j ECT RPC: M AI - 7.7 V R PC: M AI - 6.4 V R PC: M AI 10.8 V i

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Bobbin

4 3.6 V Bobbin: 3 2.8 V . Bobbin: 4 2.4 V 94% 95%  ! 89%

Lab ECT R PC: M AI - 14.7 V R PC: M AI - 13.4 V R PC: M AI - 15.0 V l 0.7 5" Lo ng 0.5 4" Long 0.75" i

i M AI - over 40 i M AI - over 90 , M AI - over 60 Degrees, within Degrees, within Degrees, within l Lab j crevice , crevice , crevice j 100% 100% 100 %

UT l max length 0.76" { max length 0.46" . max length 0.68" l

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- - - . - -- D V = Vol ts M Al= Multiple Axial Indications

• Lab ECT did not include use of TSP-IPC Transfer Standard

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. R..;s' M. .> lle'd Tubes -;Lsb Examination Results@n

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100% for o.41" Depth 100% for o.45" '

100% for 0.33" l

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Depth 95% l 96% .

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• Macrocracks 1 & 2 are interconnected near the centers of t the macrocracks by a "V-shaped

• 100% TW crack that is ,

approximately 0.1" long on each side of the "V"

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Evaluation  !

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L . . . ,e ,t e Causative Mechanisms for Cycle 7 Crack Growth Rates 1 l

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9 Tu be/FDB Misalignment  !

- Contact Region Maintained During Shutdown and Operating Conditions

- Misalignment Configuration Permits more Rapid Concentration of Contaminants than Standard FDB or TSP Intersections

- Highest Crevice Superheat Conditions at FDB Crevice which Enhances Contaminant Concentration

- High Crevice Temperature

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Causative Mechanisms for Cycle 7 Crack Growth Rates 9 Dissolved Copper Transported to Crevice Locations

- Reducible Cu Compounds Acting as Oxidants to Accelerate SCC in Alkaline Environment 9 Caustic S/G Crevice Chemistry Over Most of Cycle 7

- High Cation / Anion Ratios Present to Accelerate ODSCC in Cu Oxidizing Environment

- Alkaline Conditions Associated with Ammonia Breakthrough of S/G Blowdown Demineralizers

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Rem.edial Actions Chemistry Actions to Reduce Crack Growth 9 Reduced Copper Transport to S/G MSR Tube Bundles Replaced in RF-7 (Initial Reduction of Copper Concentration in FW by a Factor of 3 and Still Declining)

1. Reduced Potential for Caustic Crevice Environment

- S/G Blowdown /Demineralizer Operation Modified to Reduce Potential for Na Entry to S/G

- Monitoring Na/Cl Molar Ratio on a Daily Basis

- Goal to Control Molar Ratio to <;0.7 in Bulk Water to Reduce Crevice Caustic Corrosion Potential

- Existing Procedures (Response to Ratio Above Goal) to be Modified for Ratio Control

- Method (Chemical Addition or Alternatives) to Control Molar Ratio to be Implemented n ...........

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9 Even Though large Crack Growths Occurred on a Few  ;

Tubes in Cycle 7, Insitu Tube Integrity was Maintained at End of Cycle 7

- Burst Capability Satisfying R.G.1.121 Guidelines with FDB Restraint

- Potential'SLB Leakage less than that Resulting in a Small Fraction of 10 CFR 100 Dose Limits using SRP e M et hodology {

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1 9 Remedial Actions Implemented at V.C. Summer for Cycle 8 Can Be Expected to Result In:

- Reduced Crack Growth Comparable to that from Prior Operation and Typical Domestic Plant Experience

- Enhanced Margins Against Radiological Limits even if it is Postulated that a SLB Event Occurs Subsequent to Reoccurrences of Large Growth Rates l

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9 Full Cycle 8 Operation to Scheduled Refueling  !

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Remedial Actions Plant Controls to Enhance Safety Marg. ins 9 Operating Leak Limit Reduced to 150 GPD

- Increase Leak Before Break Capability 9 Increased Operator Training / Sensitivity to S/G Transien ts

- Further Enhance Likelihood of Appropriate / Timely Response to S/G Transients such as Secondary Pipe Breaks O Reduction in Coolant Activity Limit by Factor of Two

- Reduce Potential Radiological Consequences of a Secondary Pipe Break 9 Maintain 1 PORV Available (Unblocked) at all Times During Cycle 8

- Increase Confidence that S/G Tube ^_. P in a Transient will be 2335 psid with Associated Reduction in Radiological Consequences 9 Enhanced Plant Response Criteria Based on Radiation Monitors Sensing S/G Tube Leakage 9 All Flaw Indications Found in 1993 were Plugged

- Reduced Likelihood of significant indications lef t in Service Subject to Potential for Large Growth than n ....,..,,,, Following 1991 Inspection