ML20097H797

Evaluation of PTS for McGuire Unit 2

ML20097H797
Person / Time
Site:	Mcguire
Issue date:	10/31/1992
From:	Chicots J, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20097H795	List: ML20097H792 ML20097H797
References
WCAP-13518, NUDOCS 9302030012
Download: ML20097H797 (20)

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WESTING 110USE CLASS 3 (Non-Proprietary) 1 WCAP-13518 EVALUATION OF PRESSURIZED THERMAL SH0CK FOR MCGUIRE UNIT 2 J. M. Chiccts October 1992 Work Performed Under Shop Order DXBP-108B Prepared by Westinghouse Electric Corporation for Duke Power Company Approved by:

• D4 '

T.A.Meyer,Manaher Structural Reliability & Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania' 15230-0355 0 1992 Westinghouse Electric Corporation All Rights Reserved

PREFACE This report nas been technically reviewed and verified.

q

-l m, / ,

Reviewer: M. A. Ramirez '}' t A.l b, N 9

1

TABLE OF CONTENTJ Eaat Table of Contents 11 List of Tables iii List of Figures 111

1. Introduction 1

2. Pressurized Thermal Shock 3

3. Methods of Calculation of RTFTS 5 _

4. Verification of Plant-Specific Material Properties 6

5. Neutron Fluence Values 9

6. Determination of RTPTS Values for All Beltline 10 Region Materials

7. Conclusions 14

8. References 15 11

LIST OF TABLES IAhlg. Title Eaga

1. McGuire Unit 2 Reactor Vessel Beltline Region Material- 8 Properties

2. Heutron Exposure Projections at Key-Locations on the 9 McGuire Unit 2 Pressure Vessel Clad / Base Metal

-Interface for 6.05 and 32 EFPY

3. Calculation of Chemistry. Factors Using McGuire Unit 2 11 Surveillance Capsule Data

4. RTPTS Values for McGuire Unit 2 for 6.05 EFPY 12:

5. RTPTS Values for McGuire Unit 2 for 32 EFPY 13 l

l-

~

LIST OF FIGURES ,

Fiaure Title EAER

.1. Identification and Location of Beltline Region 7~

-Materials for the McGuire Unit 2 Reactor Vessel l-

-2. RTpy3 versus Fluence _ Curves for McGuire Unit 2 14 Limiting Material - Intermediate Shell Forging 05 l

iii

l 1. INTRODUCTION A limiting condition on reactor vessel integrity known as Pressurized Thermal Shock (PTS) may occer during a severe system transient such as a loss-Of-Coolant-Accident (LOCA) or a steam line break. Such transients may challenge the integrity of a reactor vessel under the following conditions:

severe overcooling of the inside surface of the vessel wall followed by high repressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect in the vessel wall.

In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on PTS. It established screening criteria on pressurized water reactor (PWR) vessel embrittlement as measured by the nil-ductility reference temperature, termed RTPTS ill. RTPTS screening values were set for beltline axial welds, forgings or plates and for beltline circumferential weld seams for the end-of-license plant operation. The screening criteria were determined using conservative fracture mechanics analysis techniques. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the criteria through end-of-license. The NRC has amended its regulations for light water nuclear power plants to change the procedure for calculating radiation embrittlesent. The revised PTS Rule was published in the Federal Register, May 15, 1991 with an effective date of June 14,1991[2] This .

amendment makes the procedure for calculating RTPTS values consistent with the methods given in Regulatory Guide 1.99, Revision 2I33

_1_

The purpose of this report is to determine the RTPTS values for-the McGuire Unit 2 reactor vessel to address the revised PTS Rule. 'Section 2 discusses the-Rule and its requirements. Section 3 provides. the methodology for calculating RTPTS. Section 4 provides the reactor vessel beltline region material properties for the McGuire Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5. The results of the RT PTS calculations are presented in Section 6. The conclusions and references for-the PTS evaluation follow in Sections 7 and 8, respectively.

2. PRESSURIZED THERMAL SH0CK The PTS Rule requires that the PTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other information that indicates a significant change in projected RTPTS values.

The Rule outlines regulations to address the potential for PTS events on pressurized water reactor vessels in nuclear power plants that are operated-with a license from the United States Nuclear Regulatory Commission (USNRC).

PTS events have been shown from operating experience to be transients that-result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The PTS concern 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 result in the propagation of flaws postulated to exist near tne inner wall surface, thereby potentially affecting the integrity of the vessel.

The Rule establishes the following requirements for all domestic, operating DWRs:

All plants must submit projected values of RTPTS for reactor vessel beltline materials by giving values for time of submittal, the expiration date of the operating license, and the projected expiration date if a change in the operating license or renewal has been requested. This-assessment must be submitted within six months after the effective date of this Rule if the value of RTPTS for any material is projected to exceed the screening criteria. Otherwise, it must be submitted with the next update of the pressure-temperature limits, or the next reactor vessel surveillance capsule report, or within 5 years from the effective date of this Rule change, whichever comes first. These values must be calculated based on the methodology specified in this rule. The submittal must include the following:

1) the bases for the projection (including any assumptions regarding core loading patterns), and

2) copper and nickel content and fluence values used in the calculations for each beltline material. (If these values differ from those previously submitted to the NRC, justification mJst be provided.)

The RTPTS (measure of fracture resistance) screening criteria for the reactor vessel beltline region is 270*F for plates, forgings, axial welds; and, 300*F for circumferential weld materials.

The following equations must be used to calculate the RTPTS values for each weld, plate or forging in the reactor vessel beltline:

Equation 1: RTPTS - I + M + ARTPTS .

Equation 2: ARTPTS - (CF)f(0.28-0.10 log f)

All values of RTPTS must be verified to be bounding values for the specific reactor vessel. In doing this each plant should consider plant-specific information that could -affect the level of ehibrittlement.

• Plant-specific PTS safety analyses are required before a plant is within 3 years of reaching the screening criteria, including ,

analyses of alternatives to minimize the PTS concern.

NRC approval for operation beyond the screening criteria is required.

4

3. METHOD FOR CALCULATION OF RTPTS-In the PTS Rule, the NRC Staff has selected a conservativo and uniform method for determining plant-specific values of RTPTS at a given time.

For the purpose of comparison with the screening criteria, the value of RTPTS for the reactor vessel must be calculated f r each weld and plate or forging in the beltline region as follows.

RTPTS - I + M + ARTPTS, where ARTPTS = (CF)f(0.28-0.10 log f)

I= Initial reference temperature (RTNDT) in 'F of the unirradiated r material M- Margin to be added to cover uncertainties in the values of initial RTNDT, copper and nickel contents, fluence and calculational procedures. M = 66'F for welds and 48'F for base metal if generic values of I are used.

M = 56'F for welds and 34*F for base metal if measured values of I are used.

f= Neutron fluence, n/cm2 (E > 1MeV at the clad / base metal interface),

divided by 10 19 CF = Chemistry factor in 'F from tables [2] for welds and for base metal (plates and forgings). If plant-specific surveillance data has been deemed credible per Reg. Guide 1.99, Rev. 2, it may be considered in the calculation of the chemistry factor.

4. VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of- the latest plant-specific material properties was performed. -

The beltline region is defined by the PTS Rule [2] to be "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 sufficient neutron irradiation damage to be considered in the sele. tion of the most limiting material with regard to radiation damage." Figure t identifies and indicates the location of all beltline region materials for the McGuire Unit 2 reactor vessel.

Material property values were obtained from material test certifications from the original fabrication as well as the additional material chemistry tests performed as part of the surveillance capsule testing program [4,5]. The average copper and nickel values were calculated for each of the beltline region mater'ials using all the available material chemistry information.

A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the McGuire Unit 2 reactor vessel are given in Table 1. All of the initial RTNDT values (I-RTNDT) are also presented in Table 1.

l U

N -- INTERMEDIATE SHELL FORGING 05, HL. 526840 CORE g -( .

CIRCUHFERENTIAL WELD SEAM, WO5 Wire Ht. 895075, flux Grau Lo, Flux Lot No.P46

_ _ _ LOWER SHELL FORGING 04, Ht. 4 M337.

I ,

Figure 1. Identification aH I_ocation cf Beltline Region Materials for the 6c. ire Unit 2 Reactor Verrel

~7-

1 1

. 1 I

TABLE 1 M*.GUIRE UNIT 2 REACTOR VESSEL DELTLINE REGION HiTERIAL PROPERTIES CU NI I-RTNDT i Mater!al (m cription (%) ('f)

(%)

')

Intermediate Shell Forging 05* l 0.153 0.793 -4 i Lower Shell forging 04 0.15 0.88 -30  :

Circumferential Weld

• 0.039 0.73 -68

• Hean values of copper and nickel as indicated below Copper Nickel Material Q3ta Sourig (wt. %1 (wt. %)

Forging 05 Original Mill Test Report 0.16 0.85 Surveillance Program (6) 0.16 0.79 Capsule V Report (7) 0.14 0.71 Capsule U Report (41 0.151 __0.82_

Mean Average Value 0.153 0.793 Forging 04 _ Original Mill Test Report 0.15 0.88 Weld Original Hill Test Report 0.05 0.70

-Surveillance Program [6] 0.031 0.73.

Capsule V Report (7) 0 03 0.66

,1psule U Report (4) 0.039 0.765 Caps 11e U Report [4] 0.036 0,747-l Capsule U Report (4) 0.045 0.776-Mean - Average . Value- 0.039- 0.730

~

5. NEUTRON FLUENCE VALUES i

The calculated fast neutron fluence (E>l MeV) et the inner surface of the McGuire Unit 2 reactor vessel is shown in Table 2. These values were projected using thn results of the Capsule U radiation surveillance programI43. The RTPTS calculations were performed using the peak fluence value, which occurs at the 45' azimuth in the McGuire Unit 2 recctor vessel.

TABLE 2 NEUIRON EXPOSURE PROJECTIONS

• AT KEY LOCATIONS ON THE MCGUlRE UNIT 2 PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 6.05 AND 32 EFPYI43 EFPY O' 15' 30* 45' 6.05 0.252 0.376 0.279 0.385 32 1.33 1.99 1.48 2.04

• Fluence x 1019 n/cm2 (E>1.0 MeV)

.g.

6. DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE REGION MATERIALS ,

Using the prescribed PTS Rule methodology, RTPTS values were generated for all beltline region materials of the McGuire Unit 2 reactor vessel as a function of present time (6.05 EFPY per Capsule U analysis) and end-of-life (32 EFPY) fluence values. The fluence data were generated based on the most recent surveillance capsule program resultsI43 The PTA Rule requires that each plant assess the RTpys values based on plant specific surveillance capsule data under certain conditions. These conditions are:

Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99, Revision 2, and RTPTS values change significantly. (Changes to RTPTS values are considered significent if the value determined with RTPTS equations (1) and (2), or that using capsule data, or both, exceed i the screening criteria prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.) .

For McGuire Unit 2, the use of p' int specific surveillance capsule data arises for the intermediate shell forging 05 and circumftrential weld because of the following reasons:

1) There have been three capsules removed from the reactor vessel, and ,

the deta is deemed credible per Regulatory Guide 1.99, Revision 2.

2) The surveillance capsule materials are representative of the actual vessel intermediate shell forging and weld materials.

The chemistry factors for the intermediate shell_ for9 tng 05 and circumferential-L weld were calculated using the surveillance capsule data as shown in Table 3.

H The chemistry factor value for the lower shell forging 04 was calculated using the Table 2 from 10 CFR 50.61[2],

+

• - . ,, -_v. -

TABLE 3 CALCULATION OF CHEMISTRY FACTORS USING HCGulRE UNIT 2 SURVEILLANCE CAPSULE DATAI43 Conporent Capsule fluence f7 DRIND1 if*DRfWD1 (ff)*2 f onctwc 05 (Long) 1 ANG V 0.337 0.701 65 45.554 0.491 N 1.45 1.103 100 110.301 1.217 U 2.02 1.192 90 107.255 1.420 fDRG1WG 05 (trans) AxlAL V 0.337 0.701 70 49.037 0.491 X 1.45 1.103 1 05 115.816 1.217 U 2.02 1.192 B5 101.?96 1.420 529.239 6.255 Chemintry f actor = $29.239 /6.255 e 64.61 Weld Metet V 0.337 0.701 45 31.524 0.4V1 x 1.45 1.103 35 38.605 1.217 U  ?.02 1.192 20 23.814 1.420 93.963 3.128 Chemistry factor a 93.963 /3.128 3 0. 04 I1m - . . .. . .

s

4

., Tables 4 and 5 provide a summary of the RTPTS values for all beltline region materials fo. 6.05 EFPY and end-of-1-icense (32 EFPY), respectively, using the PTS Rule.

TABLE 4 RTPTS VALUES FOR MCGUIRE UNIT 2 FOR 6.05 EFPY ARTNDT('F) + Initial RTNDT + Margin -

RTP15 Material (CF x FF*) ('F) (*F) (*F)

Intermediate Shell 117.2 0.7358 -4 34 116 ,

Forging 05 (84.6) 0. 7 ', ~

34 (92)

Lower Shell 115.8 0.7358 -30 34 89- 1 Forging 04  :

Circumferential 52.7 0.7358 -08 56 '27 ,

Weld Seam (30) 0.7358 -68 56 (10).

() Indicates numbers were calculated using surveillance capsule data.

• Fluence factor based upon peak inner surface neutron fluence of 3.85 x 1018 n/cm[4),

2

[

.I 12 :

b m a -

TABLE 5 RTPTS VALUES FOR HCGUIRE UNIT 2 FOR 32 EfPY 1

ARTNDT('f) + Initial RTNDT + Margin -

RTPTS Haterial (Cf x ff*) ('f) ('f) ('f)

Intermediate Shell 117.2 1.1942 -4 34 170 forging 05 (84.6) 1.1942 -4 34 (131)

Lower Shell 115.8 1.1942 -30 34 142 forging 04 Circumferential 52.7 1.1942 -68 56 51 Weld Seam (30) 1.1942 -68 56 (24)

() Indicates numbers were calculated using surveillance capsule data.

• fluence factor based upon peak inner surface neutron fluence of 2,04 x 2

1019 n/cm [4).

7. CONCLUSIONS As shown in Tables 4 and 5, all the RTPTS values remain below the NRC screening values for PTS using the fluence values for the present time (6.05 EFPY) and the projected fluence values for the end-of-license (32 EfPY). A plot of the RTPTS values versus the fluence is shown in figure 2 for the most limiting material, the intermediate shell forging 05, in the McGuira Unit 2 reacter vessel beltline region.

300 SCREENING CRITERIA 250 -

C 200 -

o. ,,,.....

3c. . ,, ,..... -

l- $$o ._ ',,... **... , . **

CC A ~~,,.

9 "',,.... *, .....~,,,. ~~~,,,

.. **,&*~~

100

~~,,.- $ 6.05 EFPY

,,. A 32 EFPY

! g i i I I I i 1 1E+18 2E+ 18 3E+18 SE+18 1E+ 19 2E+19 3E+19 SE+19 1E+ 20

FLUENCE (NEUTRONS / CM )

l INTER. SHELL FOROING j (NTER. SHELL FORGING US'NG SURV. CAPSULE DATA Figure 2. RTPTS versus Fluence Curves for McGuire Unit 2 Limiting l Material - Intermediate Shel' forging 05 l

l t

8. REFERENCES (1) 10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23, 1985.

[2] 10CFR Part 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)

[3] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.

[4] WCAP-13516, " Analysis of Capsule U from the Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program " J. H. Chicots, et al., October 1992. (Westinghouse Proprietary Class 3)

(5) WCAP-12556, " Analysis of Capsule X from the Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program," E. Terek, et al., April 1390. (Westinghouse Proprietary Class 3)

(6) wJAP-9489, " Duke Power Company William B. McGuire Unit No. 2 Reactor Vessel Radiation Surveillance Program," K. Koyama and J. A. Davidson, Hay 1979.

(7) WCAP-11029, "Ana'iysis of Capsule V from the Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program",

S. ". Yanichko, et al., January 1986.

. . . . . .