ML20096E794
| ML20096E794 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 03/31/1992 |
| From: | Chicots J, Meyer T, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20096E784 | List: |
| References | |
| WCAP-13251, NUDOCS 9205200011 | |
| Download: ML20096E794 (16) | |
Text
I 1
~
WESTINGHOUSE PROPRIETARY CLASS 3 WCAP-13251 EVALUATION OF PRESSURIZED THERMAL SHOCK FOR SOUTH TEXAS UNIT 2 J. M. Chicots March 1992 Technically Reviewed by:
E. Terek 31-Approved by:
T. A, Meyer, (hnager Structural Reliability & Plant Life Optimization Work Performed Under Shop Order H00P-106 Prepared by Westinghottse Electric Corporation for Houston Lighting and Power Company WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
- 1992 Westinghouse Electric Corporation All Rights Reserved 9205200011 920504 AcoCK0500g9 paa
TABLE OF CONTENTS PEtt
. Table of Contents-i i
List of Tables 11 List of Figures 11 1.
Introduction 1
-2.
Pressurized Thermal Shock 2
3.
Method for Calculation of RTPTS 4
4.
Verification of Plant-Specific Material Properties 5
5.
Neutron Fluence Values 7
6.
Determination of RTPTS Values for _All Beltline 9
Region Materials 7.
Conclusions
-11
~ 8.
References 12 l
p-LIST OF TABLES-Table Title Egge 1.
South Texas Unit 2 Reactor Vessel Beltline Region Material 7
Properties 2.
Neutron Exposure Projections at Key Locations on the 8
Reactor Vessel Clad / Base Metal Interface
-3.
RTPTS Values for South Texas Unit 2 10 1
LIST OF FIGURES
_ Lioure Iltle Eage 1.
Identification and Location of Beltline Region 6
Material for the South Texas-Ur.it '. Rcat. tor Vessel 2.
RTPTS_ versus Fluence Curves for South Texas Unit 2 11 Limiting Material - Intermediate Shell Pitta R2507-2 ii
1,-
INTRODUCTION--
A limiting-condition on reactor vessel integrity known as pressurized thermal shock (PTS) may occur 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.
Fracture mechanics analysis can be used to evaluate reactor vessel integrity under severe transient conditions.
i l-In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on
(
pressurized thermal shock.
It established-screening criterion on pressurized water reactor (PWR) vessel embrittlement as measured by the nil-ductility ill.
RTPTS semening values were set reference temperature, termed RTPTS
[
for beltline axial welds, forgings and plates and for beltline circumferential weld seams for end-of-life plant 3peration. 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 these criteria through end-of-life. The Nuclear Regulatory Commission has amended its regulations for light water nuclear power plants to change the procedure for calculating radiation embrittlement.
The revised PTS Rule was published in the Federal Register, May 15, 1991 with an effective date of June 14, 1991 I23 This amendment makes the procedure for calculating RTPTS values consistent with the methods given L
in Regulatory Guide 1.99, Revision 2I33 l-
_1
The purpose of this report is to determine the RTPTS values for_the South Texas Unit' 2 reactor vessel and address the revised Pressurized Thermal Shock
_(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 South Texas Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.
The results of the RTpy3 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 values.
The Rule outlines ~ regulations to address the potential for PTS events on
. pressurized water reactor (PWR) 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 produce the propagation of flaws postulated to exist near the inner wall surface, thereby_ potentially affecting the integrity of the vessel.
The Rule establishes the following requirements for all domestic, operating PWRs:
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 l
or renewal has been requested.
This assessment must be r
i
~
submitted within six months after the efective 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),
- 2) a.opper 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 must be provided.)
- The RTPTS (measu',e of fracture resistancc) Screening Criteria for the reactor vessel beltline region are 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 PTS - (CF)f(0.28-0.10 log f)
Equation 2: tiRT
- All valu s 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 embrittlement.
~
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;-
3.
METHOD FOR CALCULATION OF RTPTS Y
In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTPTS at a given time.
For the purpose of comparison with the Screening Critieria, the value of RTPTS for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as given below.
1 9 f}
RTPTS =
I + M + ARTPTS, where ARTPTS * (CI)f( '
' ' Initial reference temperature (RTNOT) c' the uMrradiated material M'-
Margin to be added to cover uncertair.th in the values of initial RTNDT, copper and nickel contents, fluence and calculational procedures. M = 66*F-for welds ano 48*F ft.- bue metal if generic values of I are used.
M = 56*F-for welds and 34*F for base metd if measured values of I: are used.
f-=
Neutron fluence, n/cm2 (E > IMeV at the: clad / base metal interface),
19 divided by 10 CF -
Chemi stry. facto) from tablesI23 for welds and for base metal (plates and forgi:.gs).
If plant-specific -surveillance data has been deemed credibla pe' Reg. Guide 1.99, Rev. 2 and two or more surveillance capsules have been tested, surveillance capsule data should be considered in the calculation of the chemistry factor.,
4.
VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES i
Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties was perf0rmed.
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 selection of the most limiting material with regard to radiation damage," Figure 1 identifies and indicates the location of beltline region materials for the South Texas Unit 2 reactor vessel, Material property values were derived from vessel fabrication material certifications (5].
Fast neutron irradiation-induced changes in the tension, fracture and impact properties or reactor vessel materials are largely dependent on chemical composition, particularly in the copper concentration. The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration with the weldments.
A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the South Texas Unit 2 reactor vessel are given in Table 1 [5],
The initial RTNDT values (1-RTNOT) are also presented in Table 1, e
1 a
e 0
d 90 0
1200 E
/
/
0 0
0 180 Wa W5 S
\\
0 2400 2?0 E
0 CORE 90 4
0 0
pp 330 0_
m 5
33a0 2100 0
270 4
Figure 1.
Identification and Location of Beltline Region Material for the South Texas Unit 2 Reactor Vessel TABLE 1
+
SOUTH TEXAS UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES [5]
CU NI I-RTNDT Material' Description
(%)
(%)
(*F)
Intermediate Shell, R2507-1 0.04 0.65
-10 Intermediate Shell, R2507-2 0.05 0.64
-10 Intermediate Shell, R2507-3 0.05 0.61
-40 Lower Shell, R3022-1 0.03 0.63
-30 Lower Shell, R3022-2 0.04 0.61
-40 Lower Shell, R3022-3 0.0" 0.60
-40 Inter. Shell Longitudinal Welds 0.0b 0.10
-70 Lower Shell Longitudinal Welds 0.05 0.07
-70 Circumferential Weld 0.05 0.07
-70 5.
NEUTRON FLUENCE VALUES The calculated-fast neutron fluence (E>l MeV) at the clad / base metal interface of the South Texas Unit 2 reactor vessel for 1.30 (April 1992), 32 and 48 EFPY are-shown in Table 2.
These values were projected using the results 'of the Capsule V radiation surveillance programI43 In the evaluation of the future exposure of the reactor pressure vessel the desian basis exposure rates were employed.
Since the South Texas Unit 2 reactor has operated for only one fuel cycle and equilibrium fuel management has not been fully established, the use of these design basis values is still appropriate.
The use of-the design basis values should result in conservative predictions of future vessel exposure that can be refined as additional dosimetry becomes available,-
4 TABLE 2 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE REACTOR VESSEL CLAD / BASE METAL INTERFACE (E>1.g]MeV)
Haterial EFPY
[n/cm 19 Intermediate Shell 1.3 1.08 x 10 3
Basemetal 32 2.89 x 1019 dB 4.34 x 1019 Intermediate Shell 1.3 6.42 x 1017 19 Long. Weld 32 1.70 x 10 At 0* Azimuth 48 2.55 x 10l9 l7 Intermediate Shell 1.3 7.33 x 10 Long. Weld 32 1.94 x 1019 At 120' Azimuth 48 2.91 x 1019 Intermediate Shell 1.3 6.75 x 1017 19 Long. Weld 32 1.78 x 10 At 240' Azimuth 48 2.67 x 1019 Intermediate / Lower 1.3 1.08 x 1018 4
19 Shell Circ. Weld 32 2.89 x 10 19 48 4.34 x 10 Lower Shell 1.3 1.37 x 1018 Basemetal 32 3.53 x 10l9l9 48 5.20 x 10 Lower Shell 1.3 6.55 x 1017 Long. Weld 32 1.70 x 1019 At 90' Azimuth 48 2.55 x 1019 18 Lower Shell Weld 1.3 1.06 x 10 At 210' and 32 2.77 x 1019 19 330* Azimuths 48 4.16 x 10 _ _ _ _ - - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - -
6.
DETERMINATION OF RTPTS VALUES FOR ALL' BELTLINE REGION HATERIALS
-Using the prescribed PTS Rule methodology, RTPTS values were generated for beltline region materials.of the South Texas Unit 2 reactor ves:el for
- 1.30 EFPY_(April 1992),32 EFPY (end-of-license EFPY) and 48 EFPY, The PTS Rule requires that each plant assess the RTPTS values based on plant specific surveillance capsule data undar certain conditions. These conditions are:
Plant specific surveillancs data has bo n deemed credible as defined in Regulatory Guide 1.99, Revision 2, and RTPTS values change significantly, (Changes to RTPTS values are considered significar.t if the value determined with RTPTS eqt'ations, er that using capsule data, ce both, exceed the screening criteria prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.)
For South Texas Unit 2, the use of plant specific surveillance capsule data does not arise because there has been only one capsule removed from the reactor vessel, hence there is insufficient data at this time.
Table 3 provides a summary of the RTPTS values for beltline region materials for-1.30 EFPY, 32 EFPY and 48 EFPY, respectively, using the PTS Rule.
l TABLE 3 RTPTS VALUES FOR SOUTH TEXAS UNIT 2 1.30 EFPY 32 EFPY 48 EFPY*
Material
(*F)
(*F)
(*F)
Intermediate Shell Plate, R2507-1 35 57 60 Intermediate Shell Plate, R2507-2 37 64 67 Intermediate Shell Plate, R2507-3 7
34 37 Lower Shell Plate, R3022-1 14 31 32 Lower Shell Plate, R3022-2 7
29 31 Lower Shell Plate, R3022-3 7
29 31 Inter Shell Longitudiani Weld 0 0*
-1 29 33 Inter, Shtli Longitudianl Weld 0 120*
-1 30 34 inter. Shell Longitudianl Weld 0 240'
-1 29 33 Lower Shell Longitudinal Weld 0 90'
-3 25 29 Lower Shell Longitudinal Weld 0 210*
1 29 32 Lower Shell Longitudinal Weld 0 330*
i 29 32 Circumferential Weld 1
30 33 The values for 48 EFPY are presented for information only. -
i-7.
CONCU1!ONS As shown in Table 3, the RTPTS values remain below the NRC screening values ist ?>*F using the projected fluence vai"et for both the end-of-life (32 EFPY) and 48 EFPY, A plot of the Rip 7$ values versus the fluence are shown in Figure 2 for the most limiting material, the intermediate shell plate, R2507-2, in the South Texas Unit 2 reactor vessel blitline region.
300 r>CREENING CRITERLA 250
+
200 LT 150 e
cc 100
$0 /
0 2E+19 4E+19 6E+19 BE+19 1E + 2n FLUENCE (n/cnf)
Figure 2.
RTPTS versus Flucace Curves for South Terds Unit 2 Limiting Material - Intermediate Shel! Plate. R2507-2. _.
8.
REFERENCES
[1]
10CfR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23, 1985.
[2]
10CfR Part 50, *Fracturo Toughness Requirements for Protection Against Pressurized Thermal Shock Events, May 15, 1991.
(PTS Rule) t
[3]
Regulltory Guide 1.99, Revision 2
" Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
[4]
WCAP-13182, " Analysis of Capsule V from the %';ston Lighting and Power Company South Texas Unit 2 Reactor Vessel Radiation Surveillance Program," J. H. Chico't., February 1992.
(Westinghouse PropH etary Class 3)
(5)
Comtustion Engineering Material Chemistry Test Records (on file at West inghouse, NATD).
_ _ - _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - -. - _ - _ - _ _ - - - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _.