ML20035H304

Summary of 930407 Meeting W/Util in Bethesda,Md to Discuss PTS & Related Technical Issues Re Facility 5-yr Const Period Recovery.List of Attendees & Meeting Handouts Encl

ML20035H304
Person / Time
Site:	Fort Calhoun
Issue date:	04/28/1993
From:	Steven Bloom Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9305040153
Download: ML20035H304 (42)

Text

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April 28, 1993 1

Docket No. 50-285 LICENSEE: Omaha Public Power District j

f FACILITY:

Fort Calhoun Station, Unit 1 1

SUMMARY

OF MEETING HELD ON APRIL 7, 1993, WITH OMAHA PUBLIC POWER DISTRICT TO DISCUSS PRESSURIZED THERMAL SH0CK (PTS) AND RELATED i

TECHNICAL ISSUES FOR FORT CALHOUN STATION l

The staff met with Omaha Public Power District (0 PPD) on April 7,1993, at One White Flint North to discuss PTS and related technical issues as they pertain to Fort Calhoun's 5-year construction period recovery (CPR).

lists the meeting attendees.

The meeting consisted of OPPD discussing their reactor vessel embrittlement management program, and PTS issues which included discussion of material properties, surveillance program, thermal hydraulics, upper shelf energy and i

fuel management / fluence analyses. is the basic information that l

was presented by OPPD.

ORIGINAL SIGNED BY:

i Steven D. Bloom, Project Manager Project Directorate IV-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation

Enclosures:

I

1. List of attendees

2. Licensee's handout cc w/o enclosures:

t See next page DISTRIBUTION w/ Enclosure 2 Docket File NRC & LPDRs PDIV-1 r/f KWichman i

BElliot Llois SMalik w/o Enclosure 2:

i JPartlow TMurley/FMiraglia JRoe l

MVirgilio GHubbard PNoonan ABeach, RIV OGC EJordan 1

ACRS(10)

JMitchell, ED0 17G21 i

OFC LA/PD4 O Q PK/PD4-1 D/PD4-1 i

NAME PNoond h vw JPellet[}h S

DATE 4/T/93 N] If 93 k /th93 COPY YESk) kES)NO YESM 0,

j OFFICIAL RECORD COPY M cument Name: APR0793M. SUM 3

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April 28, 1993

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Docket No. 50-285 t

i LICENSEE: Omaha Public Power District FACILITY:

Fort Calhoun Station, Unit I

SUMMARY

OF MEETING HELD ON APRIL 7, 1993, WITH OMAHA PUBLIC POWER DISTRICT TO DISCUSS PRESSURIZED THERMAL SHOCK (PTS) AND RELATED TECHNICAL ISSUES FOR FORT CALHOUN STATION The staff met with Omaha Public Power District (OPPD) on April 7, 1993, at One White Flint North to discuss PTS and related technical issues as they pertain to Fort Calhoun's 5-year construction period recovery (CPR).

Enclosure I lists the meeting attendees.

The meeting consisted of OPPD discussing their reactor vessel embrittlement management program, and PTS issues which included discussion of material properties, surveillance program, thermal hydraulics, upper shelf energy and fuel management / fluence analyses. is the basic information that was presented by OPPD.

I fg c

Steven. Bloom, Project Manager Project Directorate IV-I Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation

Enclosures:

I. List isf attendees

2. Licer.see's handout cc w o enclosures:

See next page m

Y c

t Mr. Terry L. Patterson Omaha Public Power District Fort Calhoun Station, Unit No. 1 CC:

Harry H. Voigt, Esg.

LeBoeuf, Lamb, Leiby & MacRae 1875 Connecticut Avenue, NW Washington, D.C.

20009-5728 Mr. Jack Jensen, Chairman Washington County Board of Supervisors Blair, Nebraska 68008 Mr. Raymond P. Mullikin, Resident Inspector U.S. Nuclear Regulatory Commission Post Office Box 309 Fort Calhoun, Nebraska 68023 Mr. Charles B. Brinkman, Manager Washington Nuclear Operations Combustion Engineering, Inc.

12300 Twinbrook Parkway, Suite 330 Rockville, MD 20852 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Harold Borchert, Director Division of Radiological Health Nebraska Department of Health 301 Centennial Mall, South Post Office Box 95007 Lincoln, Nebraska 68509 Mr. James W. Chase, Manager Fort Calhoun Station Post Office Box 399 Fort Calhoun, Nebraska 68023

l ENCLOSURE I j

MEETING BETWEEN NRC STAFF AND OHAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT 1 i

PTS ISSUES r

April 7, 1993 MEETING PARTICIPANTS NAME ORGANIZATION M. Virgilio NRR/DRPW G. Hubbard NRR/DRPW S. Bloom NRR/DRPW K. Wichman NRR/DE B. Elliot NRR/DE L. Lois NRR/DSSA l

S. Malik RES/MEB B. Jones OPPD/ Senior Vice President

5. Gambhir OPPD/ Division Manager - Production Engineering l

R. Short OPPD/ Manager - Nuclear Licensing / Industry Affairs R. Phelps OPPD/ Manager - Design Engineering K. Holthaus OPPD/ Manager - Nuclear Engineering

^

B. Weber OPPD/ Supervisor - Reactor Performance Analysis l

M. Guinn OPPD/ Supervisor - Reactor Physics l

G. Cavanaugh ABB/CE S. Byrne ABB/CE L. Connor Southern Technical Services (STS) i I

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ENCLOSURE 2 FORT CALHOEN STATION M

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OMAHA PUBLIC POWER DISTRICT l

.NRC/OPPD MEETING l

APRIL 7,1993 t

1 Omaha Public Power District l

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ENCLOSURE 2 i

l FORT CALHOUN STATION t

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j OMAHA PUBLIC POWER DISTRICT i

NRC/OPPD MEETING j

APRIL 7,1993 a

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Omaha Public Power Distric!

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Omaha Public Power District 1

AGENDA Introductions / Overview W.C. Jones Reactor Vessel Embrittlement t

Management Program K. C. Holthaus l

Pressurized Thermal Shock (PTS) Issues i

e Material Properties W. O. Weber

• Surveillance Program W. O. Weber e Thermal Hydraulics W. O. Weber

• Upper Shelf Energy W. O. Weber l

l e Fuel Management /

l Fluence Analyses M. J. Guinn Summary S. K. Gambhir.

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Omaha Public Power District 2

MEETING PURPOSE:

Discuss PTS and Related Technical Issues for the FCS 5-Year Construction Period Recovery (CPR)

CURRENT OPERATING LICENSE EXPIRATION

- June 2008 HISTORY:

July 1986:

OPPD Submits CPR Request to NRC 1986-1988:

OPPD Responds to NRC Request for AdditionalInformation July 1987:

NRC issues EA/ Finding of No Significant impact July 1988:

GL 88-11 on RV Embrittlement issued Dec.1988:

PTS Resolution Required for CPR Approval-NRC/OPPD Agree f

a May 1992:

OPPD PTS Rule Submittal July 1992:

GL 92-01 Response k

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Omaha Publit Power District 3

J REACTOR VESSEL ?.RV.

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EMBRITTLEMENT MANAGEMENT PROGRAM l

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ELEMENTS:

1 eMaterial Composition f

• Surveillance Program I

• Fuel Management i

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4 EORT CALHOUN REACTOR PRESSURE VESSEL INLET OUTLET INLET 1 - 410 8-410 2 - 410

.......S-90RE 9 - 410 3 - 410 2hD 0

90 1E O 360 AZlMUTHAL LOCATION DEGREES

l,I Omaha Public Power District 5

I MATERIAL COMPOSITION OBJECTIVES:

eDefine the RV Materials (Weld and Plate Material) oldentify the Weld / Plate Chemistry Factors i

RESULTS:

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• Well Defined Materials Database e3-410 Weld is Most Limiting l

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I Omaha Public Power District 6

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SURVEILLANCE PROGRAM OBJECTIVES:

• Confirm Material Condition and i

Validate Analytical Methods e Monitor RV Condition Using the ISI Program RESULTS:

e Good Agreement Between Analytical Predictions and Surveillance Capsule Results l

• 1992 RV ISI Confirmed "Very Clean Vessel" i

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'I FUEL MANAGEMENT OBJECTIVES:

• Accurately. Predict Limiting Weld Fluence i

with Benchmark to FCS Measurements e Utilize Fuel Management to Minimize Flux i

RESULTS:

e Accurate Fluence Predictions e Substantial Flux Reduction Achieved

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Omeha Publid Poder District i

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NRC PTS RULE: LIMITING WELD RESULTS AS OF MAY 1992 Chemistry Factor 234.5 F Estimated Fluence at EOC 14 1.0 x 1019 n/cm2 Fluence to Reach 270 F Screening Criteria:

1.45 x 1019 n/cm2 Remaining Fluence 4.5 x 1018 n/cm2 Rate of Cycle 14 Fluence Accumulation 0.34 x 1018 2

n/cm /EFPY l

l Remaining EFPY 13.2 Estimated Date to Reach Screening Criterion December 2010/

January 2011

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Omaha Public Power District 9

PTS SCREENING CRITERION AS OF APRIL 1993 e Cycle 14 Patterns December 2010/

(May 1992 PTS Submittal)

January 2011 e Cycle 15 Patterns with September 2013 Flux Reduction Only e Cycle 14 Patterns with December 2017 Expanded Chemistry Data Only i

e Cycle 15 Patterns with February 2020 Flux Reduction and i

Expanded Chemistry Data I

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l Smaha Public Power District 10 FUTURE ACTIVITIES P

e Continue Extreme Low Leakage Fuel Management Strategies

• Replace W--275 Surveillance Capsule to Contain Limiting Weld Material during 1993 Refueling Outage i

e Integrated Surveillance Program with Diablo Canyon Unit One 4

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I Omaha Public Power District 11 PRESSURIZED THERMAL SHOCK ISSUES 4

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. MATERIAL PROPERTIES f

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. SURVEILLANCE PROGRAM l

. THERMAL HYDRAULICS 1

1

. UPPER SHELF ENERGY l

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Omaha Public Power District 12 MATERIAL PROPERTIES OBJECTIVE: Identify the Limiting RV Material and Chemistry Factors.

. Extensive Record Search and Weld Sampling

-Identified from CE manufacturing records i

-Weld sample taken from FCS RV head in 1985

. Limiting Material

-Composite weld (12008,

13253, 27204)

-Weld heat #27204 is most limiting

-Chemistry Factors improved through expanded database in 1993 1

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ll,((l Omaha Public Power District 13 ART for Beltline Materials Cycle 14 -May 1992 Weld Cu Ni-Chemistry ART 'F ART'F t

Seam w/o w/o Factor F 2008 2013 2 - 410 0.17 0.17 89.45 107.66 109.85 (longitudinal) 3 - 410-0.22 1.02?

234.50?

266.02 L 271.78 (longitudinal):

s 9 - 410 0.21 0.74 187.10 214.28 218.86 (circumferential)

D-4802 0.12 -

0.56 82.20 125.75 127.76 (Intermediate shell-plate) l D-4812 0.12 0.60 83.00 126.62 128.65 (Lower shell-plate)

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Smaha Public Power District 14 i

t BELTLINE WELD CHEMISTRY FACTORS May 1992 to April 1993 Comparison-Limiting Weld Weld Wire Heat No.

Flux / Lot Cu Ni Chemistry Seam No.

w/o w/o Factor F

[

2-410 A/B/C 51989 Unde 124/3S87 0.17 0.17 89.45 3-410 A/B/C 12008 Unde 0.23 0.95 227.75 1092/3774 13253 Unde 0.21 0.73 185.45 1092/3774 3-410 A/B/C.

27204.

Unde:..

0.223.

1.02; 234.50 c 1092/3774 REVISED W 27204-:

Unde ~:

0.214 1.00 ::

229.00i 1092/3774 9 - 410 20291 Unde 0.21 0.74 187.10 1092/3833 i

Note: (1) Revision is based on 27204 wire heat data for Palisades and j

Diablo Canyon.

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REVISED CHEMISTRY FACTOR 1

IMPACT ON ART for BELTLINE MATERIALS e

May 1992 to April 1993 Comparison -Limiting Weld Weld Cu Ni Chemistry ART F ART F C

Seam w/o w/o Factor F 2008 2013 2 - 410 0.17 0.17 89.45 107.66 109.85 l

(longitudinal) i 3 - 410 0.22 1.02-234.50:

266.02 271.78.-:

(longitudinal) l REVISED O.21 1.00 229.00 m 260.02 W 265.64 Ch l

9 - 410 0.21 0.74 187.10 214.28 218.86 (circumferential)

D-4802 0.12 0.56 82.20 125.75 127.76 (Intermediate j

shell-plate) q D-4812 0.12 0.60 83.00 126.62 128.65 (Lower shell-plate)

Note (1): Does not include credit for Cycle 15 flux reduction CONCLUSION: PTS Screening Criteria Reached December 2017 4

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.Smaha Public Power District 16 SURVEILLANCE PROGRAM OBJECTIVE: To Monitor the RV Material Condition Through inservice inspection and Surveillance Capsule Evaluation.

. Reactor Vessel inservice Inspections

-1983 Inspection

-1992 Inspection-100%

. Surveillance Capsules

. Integrated Surveillance Program r

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Omaha Public Power District 17 REACTOR VESSEL INSERVICE INSPECTION

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1992 Inspection

-100% of the vessel examined

-50/70 degree search unit for vessel clad-to-base interface

-0,45 and 60 degree examinations

.Results

-8 Indications

. Largest indication -5/32" 7 had no measurable length

-Very Clean Vessel b

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FCS REACTOR VESSEL (j'

CLOSURE HEAD ASSEMBLY o...-o

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INTERMEDIATE SHELL i

ASSEMBLY i

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4maha Public Power District 19 i

SURVEILLANCE CAPSULES i

. Total of Six (6) Capsules

-ASTM -E185-1966

.W-225/W--265/W-275

-W-225 Removed in 1977 I

-W-265 Removed in 1983

-W-275 to be Removed in Fall 1993

.W-275 Capsule Replacement

-Limiting Weld Material

-Will Meet 10CFR 50, Appendix H

-Fabricated to ASTM E-185-1982 i

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((l Omaha Public Power Distrl+i 20 SURVEILLANCE CAPSULE _ LOCATIONS _

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COMPARISON OF MEASURED VERSUS PREDICTED REFERENCE TEMPERATURES FOR SURVEILLANCE CAPSULE MATERIALS Survei!!ance Material Chemistry ARTNDT ARTNOT Difference:

Capsule No.

Factor #F, RG Measured F Predicted F Measured-1.99, Rev. 2 Predicted F*

W-225 Weld 212 205 180 25 W-265 Weld 212 221 199 22 W-225

, Plate Long 65 60 55 5

W-265 Plate Long 65 74 62 12 W-265 Plate Trans 65 70 62 8

• This result is bounded by the 10 value of 28 F for welds and 17 F for plato material t

CONCLUSION: Good Agreement Between Analytical Predictions and Surveillance Capsule Results CQldPARISOlN OF MEASURED VERSUS PREDICTED CHARPY UPPER SHELF ENERGY DECREASE FOR SURVEILLANCE CAPSULE MATERIALS i

Surveillance Material Copper Upper Shelf Upper Shet!

Difference %

Capsule No.

Content %

Energy Energy Decrease Decrease I

Measured *4 Predicted %

W-225 Weld 0.35 37 38 1

W-265 Weld 0.35 43 41 2

W-225 Plate Long 0.10 13 17 4

W-265 Plate Long 0.10 23 18 5

W-265 Plate Trans 0.10 23 18 5

Reference:

Response to GL 92-01, Revision 1 (Letter (LIC-92-203R) from W.G. Gates (OPPD) to Document Control Desk (NRC), dated July 6,1992.

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=,iii l~L[l Omaha Public Power District 22 INTEGRATED 1

SURVEILLANCE PROGRAM

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OBJECTIVE: Provide a Method for Adjusting Surveillance Data from a Host Reactor to j

Provide Material Property Data Point (s) for i

the Subject (FCS) Reactor.

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.Diablo Canyon Unit 1 Host Vessel I

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.CEN-405 Program i

-RG 1.99, Rev. 2, Regulatory r

l Position 2.1 i

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Omaha Public Power District 23 i

l THERMAL HYDRAULICS

. Operating Temperatures Greater Than 525 F Except for Part of Cycle 2 i

. Partial Cycle 2 Reduced Temperature 4

Operation ~

-Approximately 522 F

-0.3 EFPY

-Reflected in W-225 and W-265 Capsule Results i

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!l [(l Omaha Public Power District 24 UPPER SHELF ENERGY OBJECTIVE: To Maintain an Upper Shelf Energy Greater than 50 ft-Ibs at the End of Vessel Life.

4

. Capsule Evaluation Concludes > 50

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ft-lbs Beyond 2013

-ABB/CE Provided Analysis

-Will be Updated With W-275 Results

-Includes Fluence Reduction Efforts i

. Generic CE Equivalent Margins Analysis

-CEN-604 Evaluation

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FUEL MANAGEMENT / FLUENCE ANALYSES OBJECTIVES:

)

. Accurately arecict limiting weld "luence 3enchmar<ec to FCS measurements Utilize state-of-the-art fuel management to minimize core aeriaheral" lux j

ELEMENTS:

. Core Loacing Patterns

. Flux Recuction Analyses

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l,I Omaha Public Power District 26

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I FUEL MANAGEMENT Low Radial Leakage

- Commenced Spring 1983 with Cycle 8

- Cycles 8,9,11,12 and 13 Extreme Low Radial Leakage

- Cycles 10,14 and 15 Extreme Low Leakage to Low Leakage (Cycle 10 to Cycle 11) j

- Non-optimized: Assumed 2-410 weld j

most limiting

- 1986 PTS submittal showed >40 EFPY of operation

- Cycle 10 resulted in significant loss of operating margin t

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FUEL MANAGEMENT { Continued?

. Cycle 14 Fuel Management

- Integral Fuel Burnable Absorber (IFBA) fuel rods

- Full-length Hafnium flux suppression rods

- Natural Uranium assemblies

. Cple 15 Optimized Fuel Management

- Reduced fluence from Cycle 14 by approximately 8%

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EXTREME LOW RADIAL LEAKAGE LOADING PATTERN 300*

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glf Twice-burned fuel with I

full-length Hafnium lh:hl hhkh l l

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'dk Natural Uranium fuel kh 1 l H

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FAST NEUTRON FLUENCE: 3-410 WELD MAY 1992 Fluence increment Fast Neutron 18 2

18 2

(10 n/cm )

Fluence (10 n/cm )

ART ( F)

Time 60 /300 180 60 /300 180 60 /300 180 Interval (EFPY)

Welds Weld Welds Weld Welds Weld 1C9clei1.0;

' N 041 LO.424,

40.255

' g 41) 9 40!4 1224;8P (224i7s i

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(Cyclelif

i1;23i

[0.423 10.27?

j 9.991

?8.97i '

F244;5)

1. 23714d Traditional "Out-in"

.A Low Radial Leakage Extreme Low Radial Leakage Fuel Management N

Fuel Management Fuel Management

Reference:

May 15,1992 PTS Rule submittal [ Letter LIC-92-175R from W. G. Gates (OPPD) to USNRC Document Control Desk)

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RATE OF FLUENCE ACCUMULATION AT 60%00 3-410 WELDS 18 (n/cm )/EFPY) 2 (x10 0.94 0.74 0.69 0.68 0.63 0.60 0.40 i

0.31 0.20 Cycles 1-7 Cycle 8 Cycle 9 Cycle 10 Cycle 11 Cycle 12 Cycle 13 Cycle 14 Cycle 15*

• Estimate j

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Omaha Public Power District 31 CYCLE 15 Expected Fluence Reduction of 8%

Compared to Cycle 14

. ABB/CE Weld Deposit Analysis

- Compared to May 15,1992 PTS Rule submittal, equivalent chemistry factor margin gain is 5.5 F

. PTS Screening Criterion Using Cycle 15 Fuel Management and Expanded Weld Chemistry Data: February 2020 l

A

!I I] I. Ornaha PQbRc PoWbr Didridt 32 PROJECTED ART FOR LIMITING WELD J.3-410)

APRIL 1993 280 RTigor Sc reening Criterion 270 y

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Feb. 20;t0_

260 Projected

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(CF = 229 F) 200 I

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I 7 8 9 101112131415161718192021222324252627282930313233343536 Cycle

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Omaha Public Power District 33

SUMMARY

r

• Aggressive, Proactive PTS Management

-Aggressive Flux Reduction

-Chemistry Data for FCS Limiting Weld is Well Documented

-Conservative Approach in Use of Uncertainties

-Methodology Well Documented /

Accepted

-100% RV 181

-Trained in-house Staff

~

• PTS Screening Criterion Reached Well After 2013

-February 2020 Using Optimized Flux Reduction and Expanded Chemistry Data 4

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SUMMARY

(continued?

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e PTS Issue Addressed for 5-Year Construction Period Recovery e Continue Initiatives Aimed at Adding Margin 1

l e NRC Submittal Planned Summer 1993 l

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I APPENDIX I

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ll,,,1 Omaha Public Power District I

I FORT CALHOUN STATION JFCSh PTS HISTORY 1967-1968 FCS Reactor Vessel (RV) Fabricated by Combustion Engineering 1973 Commercial Operation 1983 Implementation of Low Radial Leakage Fuel Management-Cycle 8 1984 RV Materials Database Search to identify Weld Wire Heats and Fluxes.

1985 RV Weld Samples Taken to identify Chemical Composition of 2-410 Weld (Wire Heat 51989) and Wire Heat 13253 (Component of 3-410 Weld).

1986 Cycle 10 Use of Non-Optimized Extreme Low Radial Leakage Fuel Management 1987 Cycle 11 Low Radial Leakage Fuel Management Resumed i

1988 eR. G.1.99, Rev. 02, More Limiting Chemistry Factor

• CEOG Task 578/632 for R. G.1.154 Begins i

t-l,I,,l Omaha Public Power District I

FORT CALHOUN STATION J.FCSS.

PTS HISTORY 1991 Cycle.14 Optimized Extreme Low Radial Leakage Design 1992

• RV 10 Year ISI-Very Clean Vessel

• 10 CFR 50.61 Screening Criterion Reached in 2010/2011 Per DOT 4 Analysis Projections e Generic Letter 92-01 Response 1993 Cycle 15 Design with Further Flux Reduction-10 CFR 50.61 Screening Criterion Reached September,2013 4

Expansion of Materials Database for 3-410 Weld / Wire Heat 27204 Shows 5.5 F Improvement in Chemistry Factor.

Data Evaluation Underway. Use of this Data Permits Operation to February 2020.