ML20057B632

Summary of 930826 Meeting W/Util in Rockville,Md Re Status of Activities Involving Unit 1 Outage Status,Cracks in Core Shrouds & Spray Sparger & Metal Turnings Found in Fuel Assemblies.List of Meeting Attendees & Slides Encl

ML20057B632
Person / Time
Site:	Brunswick
Issue date:	09/09/1993
From:	Milano P Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9309230038
Download: ML20057B632 (67)

Text

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• UNIT ED STATES NUCLEAR REGULATORY COMMISSION h

WASHINGTON, D. C. 20555

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Septster 9,1993 Docket No. 50-325 LICENSEE:

Carolina Power & Light Company FACILITIES:

Brunswick Steam Electric Plant, Unit 1

SUBJECT:

MEETING

SUMMARY

- SELECTED TECHNICAL ISSUES AFFECTING UNIT 1 RESTART 1

Representatives from Carolina Power & Light Company (the licensee) met with the Nuclear Regulatory Commission (NRC) staff on August 26, 1993, at the NRC offices in Rockville, Maryland.

The meeting was requested by the NRC to l

discuss the status of activities associated with the restart of Unit 1

{

following its present refueling outage. The topics discussed were (1) Unit 1 outage status, (2) cracks in the core shroud, (3) cracks in the core spray sparger, and (4) metal turnings found in the fuel assemblies.

The attendees at the meeting are listed in Enclosure 1 and a copy of the slides presented at the meeting are given in Enclosure 2.

The licensee stated that the Unit 1 outage status was on track, but that the fuel reload is being held up pending the completion of the inspection of the core shroud, and the final determination of the safety of operation for another cycle.

The licensee and representatives of General Electric Nuclear Engineering (GENE) then discussed the findings to date of their examinations of the core shroud cracking. A sample of the cracked area has been removed from the shroud and is undergoing various tests to determine the mechanism of the cracking and the severity.

The licensee also has a third party review underway of the crack to assure correct characterization.

The licensee is reviewing mitigation and repair options; developing future inspection criteria / scope; and determining the scope of future activities as the inspection and analysis results develop.

Also discussed were the cracks found in core spray sparger and the corrective actions undertaken. The licensee plans further inspections of this crack during the next refueling outage.

Finally, the licensee presented its findings on the metal turnings found in the fuel assemblies. All of the fuel bundles were inspected and 212 out of i

453 had these turnings. No other turnings were found and no damage to the i

fuel, internals or associated equipment is expected.

The staff has no further concern with this issue.

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/lI 9309230038 930909 PDR ADOCK 0500 5

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The meeting concluded with a commitment to meet again when further information on'the core shroud cracks is available.

Original signed by:

Patrick D. Milano, Senior Project Manager Project Directorate II-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1.

List of Attendees 2.

Slides Presented cc w/ enclosures:

See next page DISTRIBUTION: & 2 Docket File NRC/ Local PDRs PD 11-1 Reading P. Milano C. E. Carpenter E. Merschoff, R-II J. Johnson, R-II T. Murley/F. Miraglia 12-G-18 J. Partlow 12-G-18 W. Russell 12-G-18 J. Richardson S. Varga G. Lainas R. Hermann 14-C-7 T. Collins 8-E-23 L. Phillips 8-E-23 G. Hornseth 7-D-4 G. Hubbard 8-D-1 P. Anderson ACRS (10)

OGC L. Plisco 17-G-21 0FFICE L M '1(M PE PE:PD21;DRP.E PM:PD21:QRPE AD:PD21:DRPE NAME PAN CECarpYn5 D PDMilanoN SSBajwaIk DATE 9/9/93 9 / 9 /93 3/3/93 6 / 4 /93 Document Name:

BR827.MTS

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8 Brunswick Steam Electric Plant Units 1 and 2 cc:

t Mr. Mark S. Calvert Karen E. Long Associate General Counsel Assistant Attorney General Post Carolina Power & Light Company State of North Carolina Post Office Box 1551 Post Office Box 629 Raleigh, North Carolina 27602 Raleigh, North Carolina 27602 l

Mr. Kelly Holden, Chairman Mr. Robert P. Gruber Board of Commissioners Executive Director l

Post Office Box 249 Public Staff - NCUC i

Southport, North Carolina 28422 Post Office Box 29520 Raleigh, North Carolina 27626-0520 Resident Inspector U.S. Nuclear Regulatory Commission Mr. H. W. Habermeyer, Jr.

l Star Route 1, PO Box 208 Vice President Southport, North Carolina 28461 Nuclear Services Department Carolina Power & Light Company Regional Administrator, Region Il Post Office Box 1551 - Mail OHS 7 U.S. Nuclear Regulatory Commission Raleigh, North Carolina 27602 101 Marietta St., N.W.,

Ste. 2900 Atlanta, Georgia 30323 j

Mr. Dayne H. Brown, Director Division of Radiation Protection i

N. C. Department of Environmental, i

Commerce and Natural Resources Post Office Box 27687 Raleigh, North Carolina 27611-7687 Mr. J. M. Brown Plant Manager - Unit 1 Carolina Power & Light Company Brunswick Steam Electric Plant Post Office Box 10429 Southport, North Carolina 28461 Public Service Commission State of South Carolina Post Office Drawer 11649 Columbia, South Carolina 29211 Mr. C. C. Warren Plant Manager - Unit 2 i

Brunswick Steam Electric Plant Post Office Box 10429 Southport, North Carolina 28461 Mr. R. A. Anderson, Vice President Carolira Power & Light Company i

Brunswick Steam Electric Plant i

Post Office Box 10429 Southport, North Carolina 28461

i i

1 MEETING ATTENDANCE LIST CAROLINA POWER & LIGHT COMPANY BRUNSWICK, HARRIS, AND ROBINSON i

NAME TITLE ORGANIZATION fjiQE I

S. S. Bajwa Acting Project Director NRC/NRR/PD21 301-554-1466 l

P. D. Milano Project Manager NRC/NRR/PD21 301-504-1457 l

C. E. Carpenter Project Engineer NRC/NRR/PD21 301-504-1457 G. D. Miller Manager - Technical Support CP&L - BRUNSWICK 919-457-2352 Larry Phillips Chief Core Performance SRXB NRC/NRR/DSSA 301-504-3232 G. Hornseth Materials Engineer NRR/EMCB 301-504-2756 Tony Harris Sr. Specialist CP&L/ Licensing 919-457-3312 i

John Langdon NDE Supervisor CP&L/NDE Services 919-457-2072 Robert Grazio Mgr Eng. Support CP&L/NED 919-457-3470 Vauhan Waeoner Mgr Mech Unit CP&L/NED 919-546-7959 l

l Ray Hanford Chief Materials Engineer CP&L/NED 919-546-7603 Bill Campbell V. P., Nuc. Engineer CP&L/NED 919-546-3780 Carl R. Osman Principal NDE Specialist CP&L/ Tech Dept.

919-362-3375 Gus Lainas AD RII Reactors NRR/DRPE 301-504-1453 l

Gary Vine EPRI Wash. DC Rep.

EPRI 202-293-6347 i

Morris Brown Plant Manger CP&L 919-457-2210 William Koo Sr. Materials Engr.

EMCB/0E/NRR 301-504-2706

.l M. Banie Acting Project Director NRR 301-504-2771 R. Dermann Section Chief EMCB NRR 301-504-2768 R. Frahm Sr. Syst. Engr NRR 301-504-2886 i

S. L. Wu Reactor Engr NRR 301-504-3284 M. B. Sims Nuclear Specialist SNC/ Hatch 205-877-7473 Sam Ranganath Manager Struce Mech.

GENE 408-925-6825 Gerry Gordon Chief Materials GENE 408-925-6421 1

Ed Black SRNDE Specialist CP&L 919-457-3639 Steven Bertz NED Engineer CP&L 919-546-2658 James Medoff Chem. Eng. Mat. E.BR/NRR NRR 301-504-2715 l

Paul M. Byron Resident Inspector NRC/RII 919-547-9531 i

James L. Cokyely Reactor Inspector NRC/RII 404-331-5584 Michael Markley OPS. Engineer NRR/RPEB 301-504-1011 1

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AGENDA a

OPENING REMARKS J. M. Brown a

UNIT 1 OUTAGE STATUS J. M. Brown a

CORE SHROUD R. E. Grazio/GE a

CORE SPRAY SPARGER R. E. Grazio a

METAL TURNINGS IN FUEL ASSEMBLIES G. D. Miller a

CLOSING REMARKS J. M. Brown

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UNIT 1 OUTAGE STATUS u

CRD SYSTEM e

NEUTRON MONITORING SYSTEM i

a NUCLEAR SERVICE WATER SYSTEM u

CORE SHROUD INSPECTION s

CORE RELOAD u

VESSEL ASSEMBLY / HYDRO u

REACTOR STARTUP a

POWER ASCENSION

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NED LEADING EVALUATIONS m

KEY ISSUE - CRACK CHARACTERIZATION Best Available Examination Technique e

Best Available Examination Expertise o

EPRI - Third Party Review a

ENGINEERING REVIEWS CONTINUING e

CP&L - NDE, Metallurgy, Structural e

GE - Metallurgical / Structural Detailed Analyses e

SIA - Third Party Review

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CORE SHROUD CRACKING (CONT.)

a COMPLETE BOAT SAMPLE ANALYSES u

. OTHER CONTINUING ACTIVITES e

Review Mitigation / Repair Options e

Develop Future inspection Criteria / Scope Scope of Future Actions Dependent Upon inspection / Analysis e

Results a

SUMMARY

e Outage Progressing Well e

Further Evaluation Necessary Prior To Fuel Reload

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i CORE SPRAY SPARGER CRACKING 1

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CRACKS FOUND AT TWO NEW LOCATIONS (B-LOOP) e Upper Tee Box Circ Weld HAZ,3" in Length e

Lower Sparger Arm,18" from Tee Box in HAZ of Circ Weld, 2" in Length a

EVALUATION FOR CONTINUED OPERATION COMPLETED e

Submitted to NRC July 26,1993 e

Significant Structural Margin Exists Possible Loose Parts Will Not Adversely Affect Safe Operation e

Similar To Unit 2. Experience e

Acceptable For Cycle 9 Operation s

FUTURE PLANS i-e inspect During Next RFO e

implement Repair Similar to Unit 2 i

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o METAL TURNINGS IN FUEL ASSEMBLIES a

FUEL LEAKAGE IDENTIFIED DURING CYCLE 8 OPERATION m

FUEL. SIPPING IDENTIFIED TWO FAILED ASSEMBLIES LYL815 - Small Hole identified During UT With Metal Turning in Place - Assembly Repaired LYG577 - Not Being Reloaded, No Repair Made a

FUEL BUNDLE INSPECTIONS GE SIL 552 Recommendation initially inspected 60 Assembiies - 60% With Turnings e

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100% inspected - 47% With Turnings (212 Out of 453) i i

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METAL TURNINGS IN FUEL ASSEMBLIES

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METAL TURNING ORIGIN INVESTIGATION 1

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Reviewed pre-Cycle 8 outage scope e

Sample Analyzed Typical Of Type 304 Austenitic Stainless Filing Sample Sent To independent Lab For-Further Analysis e

a ADDITIONAL ACTIONS TAKEN 4

e System Engineers Reviewed Job Scope Of Current Outage To Determine System Flush Needs e

Fuel Debris Prevention. Control Program - Audit Of Cleanliness-Control Procedures Training Provided By Nuclear Fuels and Tec.hnical Support o

Lessons Learned From Hatch Experience

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UNIT 1 OUTAGE

SUMMARY

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METAL TURNINGS IN FUEL ASSEMBLIES e

Inspections Complete And Assemblies Are Ready For Reload m

CORE SPRAY SPARGER e

Acceptable Structural Margin For Cycle 9 Operation m

CORE SHROUD e

Resolve Crack Characterization issue Core Reload Delayed Until Analyses / Evaluations Completed e

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Overview of Brunswick Units 1,2 Shroud SCC Status Presented to US NRC in Support of Carolina Power & Light G.M. Gordon, S. Ranganath GE Nuclear Energy I

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1 Description of Shroud Weld HAZ Cracking Observed at Brunswick Unit 1 4

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j introduction - CP&L Overview of BNPS Shroud SCC Status - G.M. Gordon, GENE i

Description of Shroud Weld HAZ Cracking Observed at BNPS Basis for Understanding

+ Fabrication History l

Effect of Plate Orientation on IGSCC Susceptibility

+

Expected Effect of Neutron Fluence on SCC Susceptibility

+

Predicted Effect of Water Chemistry on SCC

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Current Understanding of BNPS Shroud Weld Indications

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Shroud Weld HAZ SCC Observed at Overseas BWR-4 Shroud indication Boat Sample Status Structural Margin Analysis - S. Ranganath, GENE Summary.and Review of Mitigation Options - G.M. Gordon, GENE i

CP&L Planned Actions - CP&L

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TYPE H3 50 46-43 Circ ID 70 RL (.300) IN.

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H3 90 50-27 Circ ID 70 RL (.350) IN.

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j H3 110 50-19 Circ ID 70 RL (.330) IN.

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Brunswick Unit-1 Shroud Fabrication History Shroud fabricated in 1971 by Sun Shipbuilding and Drydock-Top guide support ring (3" thick x 6" wide x 174.5".lD x 189.5" OD) consisted of six curved segments cut out of 3" plate stock L

+ Welded into ring and machined to final size Ring and rolled 1-1/2" plate materials produced by Eastern Stainless Steel RING PLATE

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+ Yield Strength, ksi 42-49 39-43

~+ UTS, ksi 82 - 86 82 Ring welded to shroud (H-3 weld) probably by submerged arc weld (procedure 75)

+ Weld Prep was single J backed up by a 3/4 inch fillet

+ Weld procedure - backchipping to sound metal; full penetration likely Post we!d stress relief heat treatment performed (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> @ 750 F). -

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FIGURE 5 - Stress dependency for SCC of sensitized AISI Type 304 in 288 C water with 0.2 ppm oxygen.

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IRRADIATED ANNEALED TYPE 304SS (288"C Pure H,0 with 32 ppm O )

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Severity of IASCC-Observsd Field

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Summary of Neutron Fluence Effects on 304 SS SCC Yield Strength increases rapidly (from ~20ksi to ~60ksi by 2 x E20)

Weld residual stress relaxes (~35% by 2 x E20)

IASCC initiation fluence threshold ~ 3 x E 20 under constant load and CERT and ~ 1 x E 21 under fixed deflection loading Large increase in SCC susceptibility observed due to synergistic interaction of neutron irradiation and thermal sensitization Neutron (and gamma) produced radiolysis of water leads to highly

-oxidizing environment (ECP ~ 200mV, SHE) inside upper shroud area

+ High ECP can shorten initiation time and increase crack growth rate t

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GENE PLEDGE Model Prediction for.

Sensitized Type 304 Crack-Growth Rate

~

4 4

Crack Growth Rate, mpy O

FOl 1000 Early-t-ife y-Operation 9.4 5 250LmV I

- 150 mV

/

Current Operation

/

' ~ -

}

O mV 10 j -100 mV '

Z-200 mV

]-300 mV -

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0.1 -

0.2 0.3 0.4 0.5.

.0.6 Conductivity, pS/cm PLEDGE: 15 C/cm2, 25ksi/in i

Y

Understanding of BNPS Shroud Cracking Observations

. Top guide support ring H-3 weld HAZ cracking likely accelerated by

+ Relatively high carbon content (~.063%)

+ Short transverse orientation of surface (can lead to early initiation and extensive circumferential extent)

+ Highly oxidizing environment (ECP ~ 200mV,SHE) and high conductivity 4

Vertical shroud crack (~1 inch long) at H-4 girth weld (~36 inches below top guide support ring) may-have initiated because of

+ Weld sensitization (~.05

.056% C)

+ Local surface cold work from shroud plate cold rolling during fabrication

+ Highly oxidizing environment (ECP ~ 150mV,SHE) and high conductivity 9

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Overseas BWR Shroud SCC Update First incident occurred in beltline region at one of the oldest BWRs in 1990 (found after ~190 on-line months)

Visual indications discovered during in-vessel TV inspection

- Several circumferential indications on inside surface

- One area of axial indications on outside surface Subsequent refined UT exams showed cracks up to ~0.6 in.

depth in 1992 and ~0.8 in.in 1993: ~0.15 in. avg crack growth Metallography confirmed IASCC mechanism

- Fluence ~8xE20 nyt; >. threshold of 3-5xE20

+ Grain boundary impurity segregation / distribution typical of IASCC

+ No apparent weld sensitization present; cracking appears driven.by oxide wedging stresses

+ Sample UT crack depth of 0.5 in. verified

BWR Core Shroud Cracking Comparison OVERSEAS PLANT SITUATION l

Cracking attributed to IASCC (Shroud had low carbon - not sensitized)

GENE RICSIL revised to recommend inspection of both high and low carbon stainless steel shrouds Crack growth analyses performed - continued operation approved Automated UT equipment developed and used in 1993 to obtain full baseline Program to better understand cause of cracking and generic implications carried out - 2 in. core sample with shroud crack fully analyzed BNPS UNIT 1 SITUATION All BNPS-1 high fluence girth welds visually inspected

- Single ~1 inch long vertical crack observed on H-4 weld OD

+ Believed to have initiated in sensitized HAZ and may have propagated by lASCC

- Circumferential H-3 weld HAZ has high carbon, low fluence - IGSCC likely mechanism - Possible cracking acceleration from neutron fluence

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Boat Sample Status Boat samples containing crack sections from H-3 and H-4 welds taken 8/20-21/93 H-3 boat sample located at Cell 42-07,142 degree azimuth H-4 boat sample located at Cell 10-47,316 degree azimuth Samples arrived at Vallecitos Nuclear Center 8/23/93 and metallographic evaluation now underway i

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-_..-__.-~.._--_,._._.._.-.._._._.--.__.__.._.._.,....-_..__-.._..._..._..._.._-...-.._.-..,-._.,.-,.-_._s.

Summary of Shroud' Weld UT Results for H-3 and H-4 Welds-WELD NO AZIMUTH CELL ID +

INDICATION ID UT DEPTH +

OD UT DEPTH + -

TYPE H3 50 46-43 Circ ID 70 RL (.300) IN.

60 RL (.200) IN.'

H3 90 50-27 Circ ID 70 RL (.350) IN.

60 RL (.280) IN.

H3 110 50-19 Circ ID 70 RL (.330) IN.

60 RL (.300) IN.

H3 140" 42-07 Circ ID 70 RL (.280) IN.

60 RL (.250) IN.

H3 230 06-11 CircID 70 RL (.180) IN.

60 RL (.180) IN.

H3 320 10-47 Circ ID 70 RL (.400) IN.*

60 RL (.300) IN.

H4 315 "

10-47 Axial OD None 70 RL (.250CCW)

(.250 CW)

• • BOAT SAMPLES TAKEN AT THESE LOCATIONS -

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=

H3 WELD H4 WELD FRACTURE MECHANICS EVALUATION a

SUMMARY

AND CONCLUSION a

f'

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POTENTIAL CRACKING MECHANISMS H3 WELD FATIGUE - UNLIKELY a

+

SYSTEM CYCLING NEGLIGIBLE

+

THERMAL CYCLING OR VIBRATING UNLIKELY

+

FATIGUE NOT LIKELY TO BE A CONTRIBUTING FACTOR IGSCC - LIKELY MECHANISM a

HIGH CARBON CONTENT IN THE FLANGE (H3 WELD)

+

PROVIDES SENSITIZATION

+

ORIENTATION EXACERBATES IGSCC-SUSCEPTIBILITY HIGHLY OXIDIZING IN-CORE ENVIRONMENT (~200mV ECP)

+

+

WELD RESIDUAL STRESS

+

POSSIBLE SYNERGISTIC EFFECT OF NEUTRON FLUENCE-

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APPLIED STRESSES AT THE H3 WELD Primary Stresses

+

Pressure

+

Weight

+

Seismic Secondary Stresses

+

Weld Residual Applied Stresses Are Small Crack Growth Mainly Driven By Weld Residual Stresses

~

SCC SUSCEPTIBILITY OF H4 WELD k

Iligh Carbon Contents 0.056% max. (but less than flange) That Could Cause Sensitization. Crack Appears To Be In The Weld HAZ.

Residual Stress Initially High But Decreases With Huence.

Outside Surface Of Shroud Exposed To Slightly Lower ECP Environment Compared To That Inside The Core.

Conservative Current Ruence Estimate: 3.1 X 10 0n/cm2.

2 Muence Based On Dosimetry: 1.6 X 1020n/cm2, Ruence Below IASCC Initiation Threshold But Could Combine Synergistically With Thermal Sensitization In Crack Growth.

L I

- Crack Initiation Due To IGSCC But Crack Growth Could Have Been Influenced By IASCC

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o CRITICAL FLAW SIZE l

FOR THE H3 WELD Depends Upon The Plane Of Cracking 1

Over 4. Sin. Parallel To The Plane Of the Ring Over 2.25in. If Crack Follows Weld HAZ Even If One Assumes Cracking Around The Circumference, The Crack Is Expected To Arrest Near The Weld Metal Sufficient Area.In The Weld Ligaments To Carry The Upset Core Delta P m

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6 Places - 60 Apart I

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i FRACTURE MECHANICS EVALUATION OF THE H4 WELD t

Applied Stresses Due To Pressure Only. Other Stresses'( weight, seismic) Not Expected To Affect i

The Axial Crack Residual Stresses Could Be High Initially. However This Could Have Been Reduced By Initial Stress 1

Relief And Subsequent Neutron Irradiation.

1 Current Fluence On The OD Surface Conservatively Estimated To Be 3X1020n/cm2 Value Based On Dosimetry: 1.6X1020n/cm2 1

}

h ALLOWABLE FLAW SIZE FOR H4 WELD.

4 Flaw Conservatively Assumed To Be A Through Wall Crack, One Inch Long Safety Factor Of 3 On Stress Allowable Flaw Size Based On LEFM = 50 in.

Higher Allowable Flaw Size Based On-Limit Analysis

_ ________.__..___._______________________ _ ___ _ __. _.___ _____._ _. _,--,__,... _..-.-..,, _.. ~,,_._ -,,.....

. [

CRACK GROWTH ANALYSIS FOR THE.H4 WELD Assuming Conservatively That The Weld Residual Stress Remains Constant With Irradiation, Bounding Crack Growth Rate For NWC Estimated To Be 8X10-5in/hr

+

Consistent With Test Data For Highly Irradiated Stainless Steel Material i

1

. Predicted Crack Length At The End Of The Next Fuel Cycle = 3 in.

Final Crack Well Below The Allowable Flaw Size

CONCLUSION Conservative Assumptions On Stress, Flaw Configuration And Crack Growth Models Used In The Analysis t

Predicted Final Crack Size At The End Of The Next Fuel Cycle.Is Well Within Allowable Values Even After Including ASME Code Safety Factor Boat Sample Provides Confirmation Of Cracking e

Mechanism l

Continued-Operation Justified For The Next Fuel Cycle 6

m

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1 Crack Mitigation Strategy Increase in Operating Dose Rates & Hydrogen Injection Normalized Steam Line Activity 6.0 5.0 -

I 1

4.0 -

I I

l I

I 3.0 -

I I

1 Data Band from

' intemals 2.0 -

.n

' Protection 22 Operating Plants l*

I i

1.0 -

8 Piping Protection 'i I

i g'

0.0 0.0 0.4 0.8 1.2 1.6 2.0 2.4 Feedwater Hydrogen Concentration (ppm)

HWC Reactor Location Dependence L

The amount ofIGSCC protection realized by HWC is f

l dependent on electrochemical potential (ECP)

- Full protection is achieved for ECPs -230 mV SHE ECP is reduced by increasing hydrogen addition rate ECP varies throughout the reactor coolant system due to radiolysis caused by high irradiation fields 1

- Higher hydrogen addition rates help counteract radiolysis effects Computerized radiolysis model developed to estimate amount of protection for various reactor regions and

, ydrogen addition rates h

- Results are highly dependent on plant configuration and operating parameters

Potential for Added HWC Protection at H-3 Weld Conservatively, H-3 weld was dispositioned using measured ECP response vs. hydrogen addition rate ECP determined with electrodes located just below top guide at FitzPatrick in close to mid-core LPRM Because H-3 crack located at about same vertical location but at periphery of core (lower neutron / gamma flux), the GENE radiolysis model used to estimate H-3 HWC benefit Model oxygen plus peroxide predictions benchmarked against ECP measured in LPRM Then, using the model oxidant predictions at the H-3 location, the ECP vs Hydrogen addition was estimated h

i

~

Comparison Between Stainless Steel Core Shroud and Inner Core Bypass ECPs 300 s

i 200 100--

S 0--

U N

Dper f -100 -

8.Vpass (y,asu d) re g

0 0

-200 --

-300 -

-400 --

\\NNhx\\\\(Projecte\\\\\\\\ ?\\

a p

r Core Shroud

-500 l

l 0

20 40.

60 80 100 H, injection (SCFM)

J i

Summary of GENE Radiolysis Model Predictions for H-3 Crack Location In-core calibration curve from radiolysis model predictions vs ECP used to predict benefit of HWC at H-3 location Model predicts significant benefit at relatively low hydrogen additions compared to hydrogen requirement needed to protect top guide

+ Up to full SCC protection (ECP < -230 mV, SHE) predicted with between 10 and 20 SCFM hydrogen addition rate compared with

~90 SCFM at in-core LPRM ECP electrode location Because of uncertainties in empirical correlation used, results must be considered tentative and only indicitive of potential added benefit of smaller HWC additions

\\

Program to benchmark actual inner shroud ECP values vs model predictions being evaluated D

g m

4 Optimum Water Chemistry Benefits SCC protection achieved with low hydrogen additions Radiation dose rate increases minimized Noble Metal Coatings (1994)

~

- Localized protection achieved even for components located in highly irradiated regions, e.g., shroud beltline and top guide Depleted Zinc Oxide / HWC (1994)

- Focused on protecting lower plenum /downcomer area.

Internal Catalytic Recombiner (Later)

- Hardware installation results in downcomer and lower plenum protection Fast Track Qualification Programs Underway

_ _ _ _. _. _ _ _ _ _ _ _ _. _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ ~ _. _ _ _ _. _ _ - _ _ _ _ _ _ _ _ _. _ _., - _

.c il Summary e

m SCC incidence in BWR fleet internals is increasing as service time accrues g

1 Brunswick Units have been significantly affected by SCC - Early life A

conductivity history has likely accelerated situation

?,

BNPS internals ISI program and current implementation status compares j

favorably with best of BWR fleet

.p Shroud weld cracking at H-3 and H-4 welds appears consistent with current E'

understanding.

Significant structural margin verified for H-4 weld Once extent of cracking is fully characterized for H-3 weld, significant crack tolerance calculated should lead to justification for operation an additional 1

fuel cycle

+ High measured shroud toughness, recent excellent plant water chemistry history plus planned HWC operation provide added margin Operation with pure water plus HWC will also significantly slow growth rate of any undetected incipient cracking and extend future crack initiation times i

I

w

.,w

SUMMARY

(C0nt'd)

. Based on existing LPRM ECP results, hydrogen addition rate sufficient for full BNPS internals component protection will require costly shielding

+ Recent radiolysis model results promising for potential reduction in H2 i

demand but not yet verified

. Optimum Water Chemistry will allow SCC protection of key internals with currently acceptable hydrogen addition rates.

+ Availability of noble metal coatings in 1994 will provide viable option for full protection of otherwise SCC susceptible shroud welds

._1

+

i F.

9 i$

8

5 E

!h

..a....

' i The meeting concluded with a commitment to meet again when further information

. ' on the core shroud cracks is available.

1 Original signed by:

l Patrick D. Milano, Senior Project Manager Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1.

List of Attendees 2.

Slides Presented 1

cc w/ enclosures:

See next page i

DISTRIBUTION: & 2 Docket File NRC/ Local PDRs PD 11-1 Reading P. Milano C. E. Carpenter E. Merschoff, R-II

}

J. Johnson, R-II

?

l T. Murley/F. Miraglia 12-G-18 J. Partlow 12-G-18 i

W. Russell 12-G-18

'i J. Richardson S. Varga G. Lainas R. Hermann 14-C-7 T. Collins 8-E-23 L. Phillips 8-E-23 l

G. Hornseth 7-D-4 i

G. Hubbard 8-D-1 P. Anderson ACRS (10) j OGC L.-Plisco 17-G-21, 0FFICE LATPkk1(NRPEPE:PD21;DRP,E PM:PD21:DRPE AD:PD21:DRPE NAME PA N M CECarpenthb PDMilanob

.SSBajwa,dk DATE Cf/9/93 9/9/93 9 / T /93 0 / 4 /93 Document Name:

BR827.MTS