ML16162A470

Summary of 830512 Meeting W/Util & B&W in Bethesda,Md Re Neutron Dosimetry & Uppershelf Matl Properties Provided in Topical Repts BAW-1697 & BAW-1699.List of Attendees,Agenda & Viewgraphs Encl

ML16162A470
Person / Time
Site:	Oconee
Issue date:	06/06/1983
From:	Suermann J Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8306200185
Download: ML16162A470 (35)

Text

JUNE 6 V0 Dockets Nos. 50-269, 50-270 and 50-287 LICENSEE:

Duke Power Company (DPC)

FACILITY:

Oconee Nuclear Station, Units Nos. 1, 2 and 3

SUBJECT:

SUMPMARY OF MEETING HELD ON NAY 12, 1983 WITH REPRESENTATIVES OF DPC AND BABCOCK & WILCOX TO DISCUSS NEUTRON DOSIiETRY AND UPPERSHELF MATERIAL PROPERTIES The meeting was held to clarify the additional information requested by NRC letter dated March 22, 1983 in regard to the NRC review of the Oconee neutron dosimetry provided in 8AW-1697 and BAW-1699. The attendees list and a copy of the meeting agenda utilized is enclosed.

Discussion Bob Gill (DPC) announced at the beginning of the meeting his tentative schedule to submit the written responses rquested by NRC in the March 22, 1983 letter by May-23, 1983.

[Project Manager's Note:

This submittal date will probably be delayed one week.]

In addressing the four areas of concern to NRC, the presentation made by Whitrarsh and Lowe (B&W) presented the case that the analysis made to date took into sufficient account the plant specific similarities for neutron fluence and that due to their method of measurement the rotation of the capsule does not affect the fluence values obtained.

Although B&W say they can support their uncertainty values used in the Oconee fluence analysis, NRC questioned the precision of the individual elements that comprise the overall cumulative uncertainty values. NRC inquired of B&W on the advisability of updating the predictions used in uper shelf life. Updating is being considered by B&W for future work.

NRC still questions the predictive value of B&W's model in addressing weld metals.

Origth lli John F. Suermann, Project Manager Operating Reactors Branch #4, DL

Enclosures:

As Stated cc w/enclosures:

See next page 6306200185 830606.

PDR ADOCK 05000269 P

PDR ORB#4:DL JFSuermann cI 6/3 /83 DAE I..................

NR RM318(10-80)RCM0240.

OFFIC. AL.RE.CO.........

RDnc CO P.

.SGPO:.1981-3 NRC FORM 318 (10-80) NRCM 0240 OFF IC IAL RECORD CO PY LJSGPO: 1981-335-960

MEETING

SUMMARY

DISTRIBUTION Licensee:

• Copies also sent to those people on service (cc) list for subject plant(s).

Docket File NRC POR L PDR ORB#4 Rdg Project Manager-JSuermann JStolz aGrimes (Emerg.

Preparedness only)

OELD NSIC ELJordan, IE JITaylor, IE ACRS (10)

NRC Meeting

Participants:

LLois BElliot PRandall

MEETING OF MAY 12, 1983 OCONEE NEUTRON DOSIMETRY NRC DPC B&W JSuermann RGill CWhitnarsh LLois JPetty ALowe, Jr.

BE11iot CHudson PRandall HBehnke CChagnon

Acenda Duke PowerNRC/B&W a:0C a.m. May 12, 193 Room P110 Philliis BuildCnC Bethesda, Maryland Introduction B.J. Elliot -

NRC R.L. Gill - DPCo Objective of B&W Presentation C.J. Hudson - B&W

,Review of &W Fluence Analysis Calculational C.L. Whitmarsh -B&W Droceoure

-cI Sources of Uncertainty C.L. Whitmarsh Status of Fluence Analysis at B&W C.L. Whitmarsh NRC Concerns for Oconee Fluence C.L. Whitmarsh/A.L. Lowe Plant Similarity Capsuie Rotation Uncertainty Values Use Limit Fuzure Action C.L. Whitmarsh/A.L. Lowe NRC Procedures Utility/Vendor Effort Summary R.L. Gill -

DPCo

DPCO/NRC/B&W iEETING OBJECTIVE c

REVIEW FLUENCE ANALYSIS COMPLETED FOR OCONEE o

ANSWER THE QUESTIONS ASKED OF DUKE POWER BY 3/22/83, LETTER FROM J. F, STOLZ TO-H. B. TUCKER, DOCKET NOS.

50-269, 50-270 AND 50-237.

o OBTAIN NRC REGULATORY POSITION AND PLANS FOR REACTOR

~II).E T TP~C 10 N

CALCULATIONAL MODEL INPUTS GEOMETRY MATERIALS POWER DISTRIBUTION TRANSPORT CODE PARAMETER DOSIMETER ACTIVITY CALCULATION DOSIMETER ACTIVITY MEASUREMENT

. REAWOR VESSEL ARRANSMlNT

.1

.;AGING sur~

CAPSLE FLUx CAPSULE LOCATIO IVs I0

Sur ee Sect Shoving Location C77 Crstcal RierUit3,Cps CR,3-T 2HgEA:"R LOCAT0I ON OF CAPSULE CR3-B SUREILLANCE

'UBEES CAPUEHOLDER TB c

lI

\\e al c

I CI I le a

0 F e I oeilic I 0 col!c 1 6

SURVEILLANCE CAPSULE HOLDER

-SURVEILLANCE CAPSULETUE HOLDER TUBE 3-4 Eb ok L W..ilIco0x

Figure 3-2.

Survo1iatice-Capntale Arresigeitict 11tl Comjtwnct I'cnoion Specimens CHRI lisp.'~ m.s

Transport Dosim-iecter Cross Section Cross Sections AI ity Activity.

Fission Spectrum StredCapsule DOT Saturated Normal-Axial R-C Activities ization Shape Geometry Factor Average Power Power Factor Distribution Fluence Vessel For Axial Capsule &

Shape

'Vessel Factor Core Fluence Escape Vessel Flux Long Term

PRIMARY sliataD PRESSURE VESSEL INLET COOLANT CAVITY CORER IE1 S

CAPSULE LOCATIONS LINER(RELATIVE)

CORE REFLECTED BOUNDARY 6 00 MAJOR THEtIMAL SHIELO XI INLET COOLANT MOE DAIREL WYASS COOLANT

ioldor Tube Wator Al Fillor Specimens CI Core

CALCULATIONAL MODEL INPUTS GEOMETRY MATERIALS POWER DISTRIBUTION TRANSPORT CODE PARAMETER DOSIMETER ACTIVITY CALCULATION DOSIMETER ACTIVITY MEASUREMENT

GEOMETRY INPUTS CAPSULE LOCATION Radio a <15%/cm Azimuthal %l%/deg (4 cm)

SIMULATION OF NON-CYLINDRICAL BOUNDARIES St reaming Attenuation EFFECT OF TEMPERATURE

• Operating condition ntE'S0Nom TOLEdANCES Nominal dimensions

MATERIALS INPUTS METAL COMPOSITION Nominal values CORE COMPOSITION Mid-cycle values COOLANT TEMPERATURE IN BYPASS REGION Between inlet and core avg. temp.

6%/240F INLET COOLANT TEMPERATURE

.2%/*F at capsule 4/OF at vessel inner surfcce Function of reactor rower

POWER DISTRIBUTION INPUTS SENSITIVITY NEAR CORE EDGE Linear proportionality Difficult to calculate BOC TO EOC

~15% variation Avg, static value CYCLE TO CYCLE

,20% in early cycles Dependent on fuel management FISSION SPECTRUM Fresh fuel primarily 2 Multiburned fuel 50-70% 239iu

TRANSPORT CODE PARAMETERS P3 Cross Sections S8 Quadroture Mesh Spacing Smaller is more accurate Smaller is more costly Microscopic Cross Sections Fundamental point data Broad group averaging Energy groups Macroscopic Cross Sections Material density

DOSIMETER ACTIVITY CALCULATION A = K Ia(E)4(E)

F (1-e

-)e NON SATURATION Reactor power vs. time Isotope half life AXIAL FLUX SHAPE RPD in nearby fuel assembly SHORT LIVED ISOTOPES Insensitive-to flux early in irradiation ENERGY RESPONSE Only fission dosimeters responsive in 2.5 > E > 1,0 MeV

>60% of fcst flux in cosule in 2.5 > E > 1.0 MeV DOSIMETER REACTION CROSS SECTIONS ENDF/BV Data Broad group averaging FISSION YIELDS ENDF/BV Data

DOSIMETER ACTIVITY MEASURE MENT COUNTING Statistics Efficiency IMPURITY REACTIONS Competing obsorption n f in fission dosimeters MATERIAL COMPOSITION Weight % of target isotope

SOURCES OF UNCERTAINTY TO VESSEL FLUENCE EFFECT OF CAPSULE NORMALIZATION ELIMINATED OR REDUCED SIGNIFICANTLY Core boundary simulation Temperature effect on dimeRsions Coolant temperature in bypass region Core and SS304 compositions Power Axial flux distribution UNAFFECTED OR PARTIALLY REDUCED Capsule location Inlet coolant temperature Power distribution Fission spectrum Microscopic cross sections (PCA)

(P)

Macroscopic cross sections (PCA)

(P)

Dimensional tolerances Mesh spacing (PCA)

(P)

Order of scattering (PCA)

(P)

Quadrature (PCA)

(P)

ADDED EFFECTS Activity measurement (P)

Activity to fluence conversion (PCA)

(P)

Flux perturbation (P)

Al A2 A3 A4 A5 AZ

/\\

1.1 LSO 0.9 0.8 0.7 0.6 0.5 0.4 0

10 20 30 40 50 Distance From PCA Core, cm

TABLE

..4 REEVALUATED EXPOSURE VALUES AND THEIR UNCERTAINTY FOP LWR PRESSURE VESSE SURVE ILL AN AP (Revision of Reference 40 cata)

?hnc

(*t p

a ow' 21 d~i mf I\\*1 rt I in 4 ft10 (1.1 ewr C m.

A M" 3.

18 (If ) " 1. 2. 3 s* o3 1 7 ) 1.54 4 1 9. 15. 11 5. 23 2

• 07 lam*.

126) 1.3 9.70*03 177) 1.57-71 1.27*10 7.651

• 01

• . n U

.79

• 9

. U

• 19 1 5) 1.15

3. 28* C 123) 1.64-1 1.42-I0

2. 90

• 0 A

1.20

• 19 2.93

• 19 (22) 2.4 6.04-02 (P7) 1.77

.s*I0 5.84

• 0)

Ls 2.2

• 19

• 19 (4 2.4 t.S1*. (9) 1.441 1.0-4 1.51

• 07 S.14

• 19

.51

• 19 (14) 1.13 9.71-M (2t) 1.66-21 4:C-3 2.438

• 1 S

1.41 ftS 1.6 * ?9 (75 1.It 2.M" (77) l.64-21 2.42*Fe 1.M

• U M "*1.51

• i.

1.U..

(17 i.114 4.;4.g

?.

• 17

1.

Ine I

3,g

• 11 2.9

• 18 (24) 1.2 1.99-43 (27) 1.75-21 1.5*Il 4.1D9

• 97 SB.

ta T

4.51

• 14 7.43

• 1 (22) 1.65 1.1-C (2.)bZ.*

1

.14.0 1.au

• 13 SE97 1

T 2.0

• Is 2.8

• 1s 9) 1.15 4.54-42 (12) 1.54-1 1.35-10 2.3/3

• S9 Lorr.

• 8.

• 1 (11) 1.01 4.143 (13) 1.4-21 1.15*10

.34

• 7 I 2.

9

• Is 2.74

• I8(

91 1.10 4.17.43 (II1 1.5.2.21 1.17.10 3.70

• 67 L-121 Is4 6.91 *is (1ll 1.17 I.01-M (16) 1.72-211 Z44 4.244 6 11?

Pi.11

.4 Is1 6.3

• Is (1#

1.?

1.1. (13) 1.71-v Z.71-10 4.3 NO 67

.Soe I & 7.0

)0 1.17

• 19 (10) 1.14 Z.1SC (14) 1.M-1t 2.82-10 5.MS * @1 C6

4.

• 1S 5.3 *

(181 7 1.4 02-4 (21 1.7141 2.221-0 4.612

• U?

e 1.t9

• 18 4.4

• 18 (10)

1. 4 1.16 4 2 ( 1 )

1.8

- 1 2.5

• 1 4.067

• 07 1.51

• I (10) 1.41-.

(3) 1.74.1 Wa.I0 a

1 1

2.22

• 19 2.17

• )9 (10)

0.

4.41-4?

(14) 2.0341 2.79-10 1.62

• 0 I

• .S

• 11 9.47

• 18 (10) 1.0 1.50-2 (13) 1.6141 1.46*10 1.al

• Ci S 4.74

• IS 7.3J

• 13 (11) 1.54 L.Z3-47 (11) 1.54.1

.5610 4.C$

• 07

• 1 211 1.5.6

• 19 (10) 1.21 4.M*-

(14) 1.54*C 2.45-10 1.40

• 0s C. c1.

• It 2.73

• tS (2!)

1.54 441-43 (7%)

1-.1 i.1 g

2.p1

. 07 loce i.

7.8

• it.)

• i (22) 1.65 5.4-.4 (Z71 1.6-1 1.23.10 4.473

• a7 meam 1t.

3 T 2.5

• 1S 2.3

• 18 (22) 1.12 5.38-02 (1) 1.3*41 1.25.10 4.211

• 47 I*7 1.8

• 18 J.06

• 1 (10)

(.70 4.51-0 (171 1.L241 1.31-IC

2. 7g

• 07 Zo I

1.52

• 11 I.C

• 19 (10) 1.14 1.647 1131 1.6141 1.41.10 1.123

• 01 lis 2.0

• Is 2.12

• i8 ( 9) 1.41 4.54-03 :12) 1.61.21 1.13*10 4.07

• 07 S41 T 2.

• IS 2.91

• Ia (2) 1.14 4.77-C (26) 1.11 1.3910 3.42

• al AM1j 1

4.40 19 6.10

• it (23) 1.29

9. 77-M (2n) 1.60421 1.3 7-U

7. 1.

• @7 Cm~

S.10

• Is 6.22

• 1s !IS) 1.22z 1.ZO-43 (191 1.464C 1

.12-C 6.191 07 oi 1

1.3

• 19 1.79

• 19 (19) 2.2 2.4-(221 1.4

.777 07 Im*24

• 19 7.26

• 19 (13) 0.89 1.25.4 1 !181 1.59-21 0.61-10 1.44

• D i" ?woo I

6.S

• 18 6.12

• 13 (121 0.9 9.1 3 I IM 1.50421 6.27-11

(.444 g

C.at t.supes 1

9 5.70

• 17 7.10

• 17 (21t) 0.22 9.t2at (201 1.26*FT 3.74*It 2.625

• @F ft 1.5

• 1.54

• tS (10) 1.0 2.11-43 (10) 1.41-21 4.07-1I

.1a

• @f 2

C 9.43

• 17 1.02

• I (101 1.01 1.50.43 (11) 1.41.-1 3.9-11 J.AM *w?

&an A

7.29

• 17 6.10

• 1I (10) 1.10 1.1s43 (11) 1.42-11 3.5-1 2.13

• 07

~ m i t. I B

1.01

• 18 1.

39

• s 9) 1.02 1.34 3

( 9) 1.40-4 1

J. -It 4.02b

• 6 1.11

• i w

.17s 1.t? t*

wto

• tm1

(..

2.6 1

I REACTOR 2.4 0

0 WESTINGHOUSE O COMBUSTION ENGINEERING 2.2

-A BABCOCK AND WILCOX ME 2.0 A

C 0

U) 00

3 1.6 L-0 0

00 0

1.4

[O

.0 f

1.2 O

O0 0

B&W ANALYSIS.

1.0 I

A 0.8 II AI0liln 70 1 12 7P 74 15 7M 77 78 19 00 81 02 03 04 05 06 07 0e 09 90 Year Original Fluence Reporten

PROGRAM TO IMPROVE FLUENCE ANALYSIS -

B&W OWNERS GROUP TO DATE SSC-1 Test PCA Blind Test ASTM E10.05 IN PROGRESS Cosule Perturbation Test DPA PLANNED PSF Blind Test ENDF/BIV or BV Transport Cross Sections Cavity Dosimetry In-Reactor Benchmark Test

REACTOR PROPERTIES PERTINENT TO VESSEL FLUENCE

1.

CORE SHAPE

2. FUEL

3. INTERNALS CONFIGURATION

4. MATERIALS OF CONSTRUCTION 5,

COOLANT DENSITY

6. POWER 7, POWER DISTRIBUTION

C) 0

Fast flux, n/cm2 sec E > I MeV ItoI I ~ l I

i I I I IFs, THERMAL SHIELD CD3 I

,E Cl Ili C")

C)

C)

/

THOLDER TUDE mn C3 1120 0)P CLAD r-m CAPSLE CLAD CI/C AL FILLER SPECIMENS O3 o

.'2 AL FILLER cu CAPSULE CLAD re3.

C3

-s o3 1120 HOLDER TUBF ncZ3 um e-

-1 ilCC3 CID C3

10 r"

C3) ru-1 PRESSURE VESSEL

IN FLUENCE VALUES

1. MATERIAL SPECIMENS IN CAPSULES, +/- 14%

SSC1 measurement comparison HEDL reevaluation of capsule data

2. VESSEL DURING CAPSULE EXPOSURE, +/- 20%

CAPSULE measurement CAPSULE location Radial extropolation from capsule to vessel Azimuthal extrapolotion to maximum location, 3,

VESSEL AT EOL (MAX, LOCATION), +/- 22%

Vessel fluence from capsule measurement Time extropolotion to EOL

4.

WELDS OUTSIDE BELTLINE REGIONS,

+/- 33%

Maximum vessel location at EOL Spatial extrapolation in azimuthal and axial directions.

Oconee Plants Controlling Weld Metals Sources of Irradiated Data Res. Prog.

RVSP Sim.W.W.

Plant Weld I.D.

Cv CF Cv CF Cv CF OC-1 SA1229(c)

X X

SA1430(l)

OC-2 WF25(c)

X X

X X

X X

WF154(c)

X X

X X

OC-3 WF67 X

X

Projecteo Oconee Plant Fluences EOL Fluence T/4 Fluenc&

Plant Weld I.D.

L.S.

T/4 to 50 ft-lbs OC-1 SA1229 9.4E18 5.2E18 8.9E18 SA1585 1.2E19 6.8E18 4.9E19 SA1493 9.0E18 5.0E18 1.5E19 SA1430 1.1E19 6.1E18 9.7E18 OC-2 WF154 9.1E18 5.1E18 7.4E18 WF25 1.2E19 6.7E18 1.1E19 OC-3 WF67 1.6E19 8.7E18

>2.0E19

- Per BAW-1511P

Oconee D lan. Weld Metals RVSP-WF25 1.1E18 81/24 30%

(30%)

WF154 9.0E18 88/39 44%

(42%)

2.4E19 88/46 52%

(48%)

7.2E18 72/20 28%

(41%)

2.4E19 72/23 32%

(48%)

Res.Data -

WF25 3.9E18 78/10 13%

(37%)

8.5E18 78/20 26%

(42%)

WF67 3.6E18 74/7 9%

(32%)

7.6E18 74/20 27%

(37%)

SA1585 4.2E18 80/3 4%

(29%)

9.5E18 80/17 21%

(35%)

Predicted Oconee Plant A Use (Per BAW-1511P)

Calendar Plant Wld I.D.

Fluence Years OC-1 SA1229(c)

> 5E18

> 16 SA1585(c)

> 5E18

> 16 SA1493(l)

> 5E18

> 16 SAl 430(l)

> 5E18

> 16 OC-2 WF154(c)

> 5E18

> 16 WF25(c)

> 5E18

> 16 OC-3 WF67(c)

> 5E18

> 16

Surveillance Capsul Data Observed A Use Vs Predicted A Use

% A Use Drop Weld I.D.

Fluence Predicted*

Observed WF193 4.0E18 22 25 WF193 6.6E18 23 25 WF209-1A 1.0E18 13 21 WF209-1A 3.4E18 17 28 WF209-1B 8.2E17 12 12 WF209-1B 3.1E18 17 26 WF182 2.3E18 22 14

• Per BAW-1511P

H SU0E PLANT SIMILARITY Equivalent specimen damage Per unit fluence, Vessel fluence is accounted for in calculation, CAPSULE ROTATION No effect on specimen overage fluence.

No effect on dosimetry overage activity, FLUENCE UNCERTAINTY Estimated values are cOoliccble.

USE LIMIT Predictions in BAW 1511P of > 16 years. (>13 EFPY) are still valid.

Based on current fluence analysis, USE is > 50 ft.-lbs, through 32 EFPY,

NRC PLANS WITH REGARD TO FLUENCE UNCERTAINTY What effect will quantification of fluence uncertainty have on vessel service life?

Will reported uncertainty values be treated individually or collectively?

CAVITY DOSIMETRY Is there any NRC interest in the use of dosimeters in the cavity of B&W reactors?

SPECTRAL EFFECTS Will DPA be used to evaluate vessel service life?