ML20217K255

Non-proprietary Evaluation of Pressurized Thermal Shock for Unit 1

ML20217K255
Person / Time
Site:	Callaway
Issue date:	07/31/1997
From:	Christopher Boyd, Howell D, Laubham T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20217K233	List: ML20217K228 ML20217K248 ML20217K255 ML20217K260 ULNRC-03664, Proposed Tech Specs 3/4.4.9 Re Pressure/Temp Limits
References
WCAP-14896, NUDOCS 9710240087
Download: ML20217K255 (19)

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WESTINGHOUSE NON PROPRIETARY CLASS 3 WCAP 14896-Evaluation of Pressurized Thermal Shock for

Callaway Unit 1 1

T. J. Laubham July 1997 l

Work Performed Under Shop Order URKP.108 l

Prepared by I Westinghouse Electric Corporation for the Union Electric Company Approved: , , Ouk C. H. Boyd, Manager 0 Engineering & Materials Techniogy Approved: , k D. A. Howell. Maliager Mechanical Systems integration WESTINGHOUSE ELECTRIC CORPORATION Nuclear Service Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355

@ 1997 Westinghouse Electric Corporation All Rights Reserved M.1997 4 * --

l o j l

PREFACE This report has been technically reviewed and verified.

Verified By: M E. Terek l

Evaluabon of PTS for Callaway Unit 1 July,1997

jj TABLE OF CONTENTS l

4 PREFACE-.-......................................................... 1 1 LI ST OF TABLE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii -

LI ST _ OF FIGU RE S . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . _ . . . . . . _ . . . . . . . . . . . . Iv -

1 I NTRODU CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 PRESSURIZED THERMAL SHOCK RULE . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . 2 3 METHOD FOR CALCULATION OF RTns.............. ,........,........ 3 4 VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES . . . . . . . . . . . . . . . . . 5 5 NEUTRON FLUENCE VALUES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8=

6 DETERMINATION OF RTn3 VALUES FOR ALL BELTLINE REGION MATERIALS . . . . . . 9 7 CON CLU SION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .................... 12 8 REFERENCES ...... ,, ......................... ........... 13 Evaluation of PTS for Callaway Unit 1 July, ,1997

lii LIST OF TABLES *

. Table 1 - Callaway Unit 1 Reactor Vessel Beltline Region Materials Properties . . , , . . . . 7 Table 2 . Fluence (E > 1.0 MeV) on the Pressure Vessel Clad / Base Metal Interface for Callaway Unit 1 @ 35 EFPY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 Table 3 -Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR Part 5^ 51. . .

_ 9 l

Table 4 Calculation of Chemistry Factors Using Surveillance Capsule Data Per 10CFR Part 50.61 , , . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Table 5 RTns Calculations for Callaway Unit 1 Beltline Region Materials at EOL  !

(35 E F PY) . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 i

4 Evaluation of PTS for Callaway Unit 1 July,1997

iv LIST OF FIGURES l

i j

Figure 1 Identifcation and Location of Beltiino Region Material for the Callaway Unit 1 Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . ................... ............. 6 Evaluadon of PTS for Callaway Unit 1 July,1997

3 1- lNTRODUCTION A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. - A PTS concem arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner wall surface, thereby potentially affecting the integrity .

of the vessel. I The purpose of this report is to determine the RTn3 values fu the Callaway Unit i reactor vessel using the results of the surveillance Capsule V evaluation. Section 2.0 discusses the PTS Rule and its requirements.

Sechon 3.0 provides the methodology for calculating RTns. Section 4.0 provides the reactor vessel beltline region material properties for the Callaway Unit 1 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0. The results of the RTn, calculations are presented in Section 6.0. The conclusion ano' references for the PTS evaluation follow in Sections 7.0 and 8.0, respectively, f

Evaluatior, of PTS for Callaway Unit 1 July.1997

8 2- PRESSURIZED THERMAL SHOCK RULE The Nuclear Regulatory Commission (NRC) recently amended its regulations for light-water cooled nuclear power i plants to clarih/ severalitems related to the fracture toughness requirements for reactor pressure vessels, )

including pressurized thermal shock requirements. The revised PTS RuleN ,10 CFR Part 50.61, was published in the Federal Register on December 19,1995, with an effective date of January 18,1996.

This amendment to the PTS Rule makes three changes:

1. The rule incorporates in total, and therefore makes binding by rule, the method for determining the reference terrperature, RT , including treatment of the unitradiated RT, value, the margin term, and the explicit definition of ' credible

• surveillance data, Weh is currently described in Regulatory Guide 1.99, Revision 2A,

2. Tht rule is reswetured to improve clarity, with the requirements section giving only the requirements for the vaw for the reference temperature for end of life fluence, RT,13

! 3. Thermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby reducing RT,13 The PTS Rule requirements consist of the following:

For each pressurized water nuclear power reactor for which an operating Icense has been issued, the licensee shall have projected values of RT,13, accepted by the NRC, for each reactor vescel beltline material for the EOL fluence of the material.

The assessment of RTp13 must use the calculation procedures given in the PTS Rule, and must specify the bases for the projected value of RT,73f or each vessel beltline material. The report must specify the copper and nickel contents and the fluence values used in the calculation for cach beltline material.

This assessment must be updated whenever there is a significant change in projected values of RT,33 or upon the request for a change in the expiration date for operation of the facility. Changes to RT,13 values are significant if either the previous value or the current value, or both values, exceed tre

. screening criterion prior to the expiration of the operating license, including any renewal term, if applicable for the plant.

The RT,13 screening criterion values for the beltline region are:

270*F for plates, forgings and axial weld materials, and 300*F for circumferential weld materials.

Evaluation of PTS for Callaway Unit 1 July.1997

3 L 3- METHOD FOR CALCULATION OF RTpy,_ d RTritmust be calculated for each vessel beltline material using a fluence value, f, which is the EOL fluence for ,

[ the material. Equation 1 must bc used to calculate values of RT,ci for each weld and plate or forging in the reactor vessel beltline.

l i

RTwo7 RTwoTru)*M+ ARTway 0)

-- RT.,g = reference temperature for a reactor vessel material in the pre-service or unirradiated condition M' =

Margin to be added to account for uncertainties in the values of RT,cin, copper and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2.

M 2/o u*o$ (2) o uis the standard deviation for RT,e3g.

o u= 0*F when RT,e,y is a measured value o u= 17'F when RT, erg is a generic value o, is the standard deviation for ART.1 For plates and forgings; o, = 17'F when surveillance capsule data is not used o, = 8.5'F when surveillance capsule data is used 4

For welds:

o, = 28'F when surveillance capsule data is not used o, = 14'F when surveillance capsule data is used o, not to exceed one-half of ART,ci.

ART,c, is the mean value of the transition temperature shift, or change in RT,ey, due to irradiation, and must be calculated using Equation 3.

ARTw07-(CF)=f m2e-o.iw (3)

Evaluation of PTS for Callaway Unit 1 July,1997

4 CF (*F) is the chemistry factor, which is a function of copper and nickel content. CF is determined from Tables 1 and 2 of the PTS Rule (10 CFR 50.61). When surveillance data is deemed credible it must be used to determine a material specific value of CF, A material-specific value of CF is determined using Equation 5.

In Equation 3 f is the best estimate neutron fluence, in units of 10" niem'(E > 1.0 MeV), at the clad base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest f:uence. The EOL fluence is used in calculating RT,,3 Equation 4 must be used for determining RT,13 using Equation 3 with EOL Clad / Base Metal fluence values to determining ARTris.

RTns RTwoT(u)*M+4RTns (4)

To verify that the RT uor value for each vessel beltline material is a bounding value for the speeific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. Results from the plant specife surveillance program must be integrated into the RTuo, estimate if the plant specife surveillance data has been deemed credible.

A material-specific value of CF is determined from Equation 5.

CF= E[A ii

.f""#] (5)

In Equation 5. 'Al is the measured value of ART,,01 and 'f' is the fluence for each surveillance data point. If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e., differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measure values of ART,or must be adjusted for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance weld.

Evaluabon of PTS for Callaway Unit 1 July,1997

5 4 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressunzed thermal shock evaluation, a review of the latest plant specific material

. properties for the Callaway Unit 1 vessel was performed. The beltline region of a reactor vessel, per the PTS Rule, is defined as 'the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience suffeient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage'. Figure 1 identifes and indicates the l_ ,

location of all beltline region materials for the Callaway Unit i reactor vessel.

The bem estimate copper and nickel contents of the beltline materials ware obtained from Reference 6 and documented in Table 1. These average copper and nickel values were calculated using all of the available material chemistry information. Initial RT values for Callaway Unit 1 Reactor Vessel Beltline Region Material Properties are also shown in Table 1.

Evaluation of PTS for Ca!!away Unit 1 July,1997

, 6 I.

N d

a m

E 0 90 l0 l j 1-124A' R2707-2 4 .2707-3 0 T ~~~

w E_

f 101-124C ' 101-124B

- N 27G7-1 2700 CORE 4 101-171 0 90

[ p 101-142A R2708-2 R2708-1 a

0 4 k 0 0 180 101-142B 101-142C R2708-3 N 270 0

Figure 1: Identification and Location of Beltline Region Material for the Callaway Unit 1 Reactor Vessel Evaluation of PTS for Callaway Unit 1

4 7

l Table 1 Cesilsway Unit 1 Reactor Vessel Beltline Region Material Properties Material Description Cu(%)* Ni (%) *

• RTwiui Intermediate Shell Pbte R27071 0.05 0.58 40'F Intermediate Shell Rate R2707 2 0.06 0.61 10 'F Intermediate Shell Rate R2707 3 0.06 0.62 10 'F Lower Shell Rate R2708-1 0.07 0.58 50 'F Lower Shell Rate R2708-2 0.06 0.57 10 'F Lower Shell Rate R2708-3 0.08 0.62 20 'F Intermediate and Lower Shell Longitudinal Weld Seams"' O.04 0.06 -60 'F Intermediate to Lower Shell Cecumferential

, Weld Seam"' O.04 0.06 -60 'F H.QEL (a) Average values of copper and nickel as indicated in reference 6.

(b) The RTag values for the plates and welds are measured values and were obtained from WCAP 129485 (c) All vessel beltiine weld seams were fabricated with weld wire heat number 90077, The intermedit,, to lower shell circumferential weld seam 101-171 was fabricated with Flux Type 124 Lot Number 1061. The intermediate and lower shell longitudinal weld seams were fabricated with Flux Type 0091 Lot Number 0842. The surveillance weld meta!

was fabricated with weld wire heat number 90077. Flux Type 124 Lot Number 1061. Per Regulatory Gide 1.99, Revision 2, ' weight percent copper' and ' weight-perc' 'icke!' are best-eshmate values for the matenal, which will normally be the mean of the measured values for a pite or forging or for the weld samples made with tne weld wire heat numtw that matches the entcal vessel weld. The surveillance weld metal was made wth the same weld wre heat as all of the vessel beltline weld seams and is representative of all of the beltiine weid seams.

Evaluaten of PTS for Callaway Unit 1 July,1997

8 5 NEUTRON FLUENCE VALUES l

The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the Callaway Unit i reactor vessel are shown in Table 2. These values were projected using the results of the Capsule V radiation er.aiysis. See Section 6.0 of the Capsule V analysis report, WCAP 148951 5 l

Table 2 Fluence (E > 1.0 MeV) on the Pressure Vessel Clad / Base Metalinterface for  !

Callaway Unit 1 @ 35 EFPY MATERIAL FLUENCE Intermediate Shell Plate R27071,2 and 3 2.074 x 10" n/cm' intermediate and Lower Shell Longitudinal Weld Seams 101124A and 101142A 1.167 x 10" n/cm' ,

I (90' Azimuth)

Intermediate and Lower Shell Longitudinal Weld Seams 2.042 x 10 n/cm' 101124B&C and 101142B&C (210' & 330' Azimuth)

Intermediate to Lower Shell Circumferential Weld Seam 2.074 x 10" n/cm' 101 171 Lower Shell Plate R2708-1,2 and 3 2.074 x 10" nlcm' Evaluation of PTS for Callaway Unit 1 July,1997

R 9

I

! 6 DETERMINATION OF RTers VALUES FOR ALL BELTLINE REGION I MATERIALS i

Using the prescribed PTS Rule methodology, RT,r3 values we,e generated for all belthne region materials of the Callaway Unit 1 reactor vessel for fluence values at the EOL (35 EFPY).

Each plant shall assess the RT,33 values based on plant-specific surveillance capsule data. For Callaway Unit 1, the related surveillance program results have been included in ..as PTS evaluation. Specifically, the Callaway Unit i plant-specific surveillance capsule data for the lower shell plate R27081 and weld meta! is provided for the following reasons:

1) There have been three capsules removed from the reactor vessel, and the data is deemed credible per Regulatory Guide 1.99, Revision 2.

2) The surveillance capsule program is credible (See Appendix D of Reference 5).

As presented in Table 3, chemistry factor values for Callaway Unit i based on average copper and nickel weight percent were calculated using Tables 1 and 2 from 10 CFR 50.6114 Additionally, chemistry factor values based on credible surveillance capsule data are calculated in Table 4. Table 5 zntains the RT,13 calculations for all beltline region materials at 35 EFPY.

Table 3 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR Part 50.61 Material Ni, wt % Chemistry Factor. 'F IDLtDna< hate Shell Plate R27071 G=en cu wt% = 0 05 W 310 IntennematelhcL21 ate R2707 2 Gwen cu wt% = 0 06 00 3M JDitDDfddle Shell Piate R2707J 0.62 37.0 Gwen cu wt% = 0 06 Lower Shell Pl ate R27081 Gwen Cu wt % = 0.07 038 40 Lower Snell Plate R27QU 0.57 37.0 G=en cu wt % = 0.06 LQ*tdibt!!Eate R2708 3 0 62 51 0 Gwen Cu wt % = 0 08 Sepne Recon wetd Meta!

Gwen Cu wt % = 0 D4 OM 29 7 Surveeance Cgwe Pmoram WelGt;a' Gwen Cu wt % = 0 045 O.M5 3t8 survemanc:: Program Maienal Evaluation of PTS for Callaway Unit 1 July,1997

10 l

I Table 4 Calculation of Chemistry Factors UsinD Surveillance Capsule Data Per 10CFR Part 50.01 l

Malenal Capsule Capsule t FF" ART,c," FF' ART c FF' Lower Shell Plate U 0.3342 0.698 7.33 5.12 04B7 R2708-1 (Longitudinal) Y 1.237 1.059 25.15 26.63 1.121 V 2.359 1.232 16 45 20.27 1.518 Lower Shell Plate U 0.3342 0.698 25.66 18 05 0.487 R2708-1 (Transverse) Y 1.237 1.059 46.39 49 13 1.121 V 2.359 1.232 44.82 55.22 1.518 SUM 164.18 6.252 CF,, , = E(FF ' ART,c,) + E(FF') = (164.18) + (6.252) = 26.3*F Veuel Weld Metal U 0.3342 0.698 64.01 44.68 0.487 Based on Surveillance Program Weld Metal Y 1.237 1.059 34.48 36.51 1.121 Results*

V 2.359 1.232 45.03 55.48 1.518 SUM 136.67 3.126 CF = E(FF ' ART,c,) + E(FF') = (136.67) + (3.126) = 43.7'F MQIES (a) f a fluence (10 nicm' E > 1.0 MeV). M updated fluence values were taken from the Capsule V analysis (Table 8

612 of WCAP.14895 ).

(b) FF = fluence factor = f ""'"*

(c) ART,c, values were obtained from the Capsule V analysis" (d) The surveillance weld metal ART,c, values have been adjusted by a ratio of 0.934 (CF + CF, , = 29.7 - 31.8 = 0.934).

Evaluat:on cf PTS for Callaway Unit 1 July.1997

11 Table 5 RTm Calculations for Callaway Unit 1 Beltline Region Meterials at EOL (35 EFPY) umummusummmmmmmmma summmumusumm-Ma'ena: f' FF*' CF RT,oi,"' Margin ART,/ RT,,,

(*F) (*F) (*F) (*F) (*F)

Inter. SheB Rate R27071 2.074 1.20 31 0 40 34.0 37 2 tit inter. She# Rate R2707 2 2.074 1.20 37.0 10 34 0 44 4 84 Inte. Shell Rate R2707-3 2.074 1.20 37.0 10 34.0 44 4 68 Lower Shen Rate R27081 2.074 1 20 44 0 50

- . . . . . 52 8 137

- . . . . . . . . . . . .........~ . . . ... .... ............. . . 34. 0 . . . . . ........

Ush1 S/C Data 2.074 1.20 26.3 50 17.0 31.6 90 Lower Shen Rate R2708-2 2.074 1.20 37.0 10 34 0 44 4 88 Lower Shel PWe R2708-3 2.074 1.20 51.0 20 34 0 61.2 115 Intermediale and Lower Shen 1.167 1.04 29.7 40 30.9 30.9 2 Longitud nal Weld Seams 101124A & 101142A (90' Arimutn)

Using S/C Data 1,167 1.04 43.7 40 28.0 45 4 13 Intermedete and Lower Shen 2.042 1.19 29.7 -60 35.3 35.3 11 Longitudinal Weld Seams 1011248&C & 101142B&C (210' & 330' Azirnuth)

Usmg SIC Data 2.042 1.19 43.7 60 28.0 52.0 20 Intermediate to Lower Snea 2.074 1.20 29.7 40 35.6 35.6 11 Circumferenbal Weld Seam 101 171 Using SIC Data 2.074 1.20 43.7 60 28.0 52.4 20 bOTES:

(a) f = peak dad! base metal interface fluence (10 nicm', E > 1.0 MeV). See Tabe 2.

(d) FF = f 8 ' ' "

(O RT,or, values are measured values.

(d) .6RT,13 : CF'FF l

Evaluation of PTS for Callaway Unit 1 July,1997

m 1a L

7 CONCLUSIONS -

As shown in Table 5, all of the beltline region materials in the Callaway Unit i reactor vessel have EOL RT,33-values well below the screening cnteria values of 270*F for plates or forgings and _ longitudinal welds and 300*F -

for cucumferential welds at EOL (35 EFPY). .

Evaluaton of PTS for Callaway Unit 1 July,1997 r .

I

Lt. 13

8. - REFERENCES

[

1. 10 CFR Part 50.61, ' Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events', Federal Register, Volume 60, No. 243, dated December 19,1995.

2. Regulatory Guide 1.99, Revision 2,

• Radiation Embrittlement of Reactor Vessel Materials,' U.S. Nuclear Regulatory Commission, May 1988.

3. WCAP 9842, ' Union Electric Company Callaway Unit No.1 Peactor Vessel Radiation Surveillance Program *, LR. Singer, May 1981.

4. WCAP 11374, Revision 1 ' Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program', S. E. Yanichke, et. al., June 1987.

5. WCAP 14895, ' Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program", E. Terek, et. al., July 1997.

6. EDRE-EMT 121, 'Best Estimate Copper / Nickel for Reactor Vessel Beltline Materials @ Callaway Unit 1*,

T. J. Laubham, March 17,1997.

7. WCAP412946, ' Analysis of Capsule Y from the Union Electric Company Callaway Unit i Reactor Vessel Radiation Surveillance Program *, E. Terek, et. al., June 1991.

8. WCAP 12948, ' Evaluation of Pressurized Thermal Shock for Callaway', J.M. Chicots, et. al., May 1991.

Evaluation of PTS for CaHaway Unit 1 July,1997