ML20097G281
| ML20097G281 | |
| 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 | |
| ML20097G280 | List: |
| References | |
| WCAP-13252, NUDOCS 9206170004 | |
| Download: ML20097G281 (16) | |
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WESTINGHOUSE PROPRIETARY CLASS 3 WCAP-13252 t
I l
EVALUATION Of PRESSURIZED THERMAL SH0CK FOR SOUTH TEXAS UNIT 1 l
J. M. Chicots March 1992 7
Technically Reviewed by: k hv E, Terek b% N Approved by:
T A. Meyer lMan'ager Structural Reliability & Plant Life Optimization Work Performed Under Shop Order H00P-106 Prepared by Westinghouse Electric Corporation for Houston Lighting and Power Company WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Olvision P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 0 1992 Westinghouse Electric Corporation All Rights Reserved
\\
1
. TABLE OF CONTENTS han Table of Contents 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
))
8.
References 12 i
LIST Of TABLES Table Title 119.9 1.
South Texas Unit 1 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 1 10 LLST OF flGURES Fiaure lills Ejtqg 1.
Identification and Location of Beltline Region 6
Material for the South Texas Unit 1 Reactor Vessel 2.
RTPTS versus Fluence Curves for South Texas Unit -1 11 Limiting Material - Intermediate Shell Plate, R1606-3 l
l i
l l
l ii
^
1.
INTRODUCTION i
A limiting condition on reactor vessel integrity known as pressurized thermal shock (PTS) may occur during a severe system transient such as a l
loss-of-coolant-accident (LOCA) or a steam line break.
Such transients may challenge the integrity of a reactor vessel under the following conditions:
f severe overcooling of the inside surface of the vessel wall followed r
by high repressurization; significant degradation of vessel material toughness caused by radiation embrittivment; 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.
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 i ll.
RTPTS screening 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 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 i
embrittlement in accordance with these'eriteria.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[2].
This amendment makes the procedure for calculating RTPTS. values consistent -with the methods given in Regulatory Guide 1.99, Revision 2I33 f
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f The purpose of this report is to determine the RTPTS values for the South Texas Unit I reactor vessel and address the revised Pressurized Thermal Shock l
(PTS) Rule.
Section 2 discusses the Rule and its requirements.
Section 3 provides the methodology for calculating RTPTS.
Section 4 provides the reactor vessel heltline region material properties for the South Texas Unit 1 reactor vessei.
The neutron fluence values used in this analysis are presented in Section 5.
The results of the RTPTS calculations are presented in Section 6.
The conclusions and references for the PTS evaluation follow in Sections 7 and 8, respectively, i
i i
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.
l The Rule establishes the following requirements for all domestic, operating l
PWRs:
l 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 l !
submitted within six months after the effective date of this Rule if the value of RTPTS for any material is projected to l
exceed the screening criteria. Otherwise, it must be submitted i
with the next update of the pressure-temperature limits, or the l
next reactor vessel surveillance capsule report, or within 5 l
years from the effective date of this Rule change, whichever j
comes first.
These values must be calculated based on the l
methodology specified in this rule.
The submittal must include the following:
l
- 1) the bases for the projection (including any assumptions l
regarding core loading patterns),
l
calculations for each beltline material. (If these values
{
differ from those previously submitted to the NRC, justification must be provided.)
f
- The RTPTS (measure of fracture resistance) Screening Criteria for the reactor vessel beltline region are j
270'F for plates, forgings, axial welds; and, 300*F for circumferential weld materials.
The following equations must be used to calculate the RTpjg 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 embrittlement.
a
-3 l
fJ Plant-specific PTS safety analyses are required before a plant is o
within 3 years of reaching the Screening Criteria, including analyses of alternatives to minimite the PTS concern m
NRC approval for operation beycad the Screening Criteria is required.
3.
McThsD FOR CALCULATION OF RTPTS
~
In the PTS Rule, the NRC Staff has selected a conservative and uniform method for datermining plant-specific values of RTPTS at a given time.
Tc-af comparison with the Screening Critieria, the,:;ue of "Tp mactor vessel must be calc 91ated for each weld and plate or c
forc.,g n
aeltline region as given below.
1 9 }
Rh,,
' + M + ARTp33, where ARTPTS " ICI)f( '
~'
+
1-Initial reference temperature (RTNDT) of the unirradiateJ material M=
Hargin to be added to t.over 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.
c M = 56** hr n 'ds and 34*F for base metal if measured values
)
of I are u cc.
f=
Neutron luence, n/cm2 (E > )MeV at the clad / base metal interface),
19 divided by 10 i
CF =
Chemisti) factor from tablesI2} for welds ud for base metal (plates and forgings).
If plant-specific surveillai.ce data has been, deemed credible per Reg. Guide 1.99, Rev. 2[3] and two or_ more surveillance capsules have been tested, surveillance capsule da'. should be considered in the calculation of the chemistry factor.
I
! l l
4.
VERIFICATION OF PLANT-SPECIFIC MATERIAL'PkOPERTIES I
Before performing the pressurized thermal shock evaluation, a review of the
~
f latest plant-specific material properties was performed.
[
[
The beltline region is defined by the PTS Rule (23 to b'e "the region of the I
reactor. vessel (shell material including welds, heat;affected zones and plates
[
or forgings) that directly surrounds t.
9ffective height of the active core.
l and adjacent regions of the reactor vessei that are-predicted to experience:
=
sufficient neutron irradiation damage to be. considered in the selection of the-1 i
i most limiting material with regard to radiation damage."
Figure T identifies' i
and indicates the location of beltline-region materials-for the South Texas
{:
Unit I reactor vessel.
l Material property values were derived from vessel fabrication material =
4-certifications [5].
Fast neutron irradiation-induced changes in'the tension, I
j fracture and impact properties of reactor vessel materials are largely:
dependent on chemical composition,. particularly in -the copper concentration.
j
'The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration within the.
1
-l A summary of the' pertinent chemical and mechanical properties of the beltline
{
. region plate and. weld materials of the South Texas Unit I reactor vessel r o-given in Table 1 [5]. The initial RTNDT values (I-RTNDT) are also presenteo i-in Table 1.
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ad 5
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2 4
0 0-1800 2
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2700 2400 5
0 CORE 90 d
0 o
180 N
0 3300 210 m
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0
-270 4
Figure 1, Identification and Location of Beltline Region Material for the South Texas Unit 1 Reat.ier Vessel.
6
' TABLE 11 SOUTH TEXAS UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES [5):
CU NI i-RTNDT Material Description
_(%)
(%)
('F)
Intermediate Shell, R1606-1 0.04 0.63 10 Intermediate Shell, R1606 0.04 0.61-0 Intermediate Shell, R1606-3 0.05 0.62 10 Lower Shell, R1622-1 0.05 0.61
-30 Lower Shell, R1622-2 0.07 0.64
-30 Lower Shell,.R1622 0.05 0.66
' 30-Inter. Shell Longitudinal Welds 0.03-0.05
-50 Lower Shell Longitud _inal Welds 0.03
-0.05
-50 circumferential-Weld 0.03l 0.04 -
-70 5.
NEUTRON FLUENCE VALUES-1 The calculated fast neutron fluence (E>l MeV) at the clad / base metal-interface of the South Texas Unit I reactor _ vessel for 2.5 (April 1992),
32 and 48 EFPY are shown in Table 2.
These values were projected using the results of the Capsule U radiation surveillance programI43. _ In th'e evaluation of the future exposure-of.the. reactor pressure vessel the design basis exposure rates were employed. Since.the South Toxas= Unit:1 reactor has aperated for only_ o_ne fuel cycle:and equilibrium fuel management has not-been fully established.: the use of:these design basis u
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.
, L.
__ ___ i - _ _ _ _ _ -. - _ - _ _
TABLE 2 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE REACTOR YESSEL CLAD / BASE METAL INTERFACE (E>1.g]HeV) l Material EFPY
[n/cm 18 Intermediate Shell 2.5 2.27 x-10 Basemetal 32 2.90 x'1019 48 4.35 x 1019 18 Intermediate Shell 2.5 1.33 x 10 Long. Weld 32 1.71 x 1019 At O' Azimuth 4u 2.56 x 1019 i
Intermediate Shell 2.5
- 1. v.' x 1018 Long. Weld 32 1.95 a 1019 At 120' Azimuth 48 2.92 x 10l9 Intermediate Shell 2.5 l'.39 x 10I8 19 Long. Weld 32 1.79 x 10 At 240' Azimuth 48 2.68 x 1019 Intermediate / Lower 2.5 2.27 x 1018 Shell Circ Weld 32 2.90 x 10l9 19 48 4.35 x 10 18 i
Lower Shell 2.5
-2.84 x 10 Basemetal 32 3.54 x 1019 48 5,31 x 1019 Lower Shell 2.5 1.35 x 1018 Long. Weld 32 1.71 x 1019 i
At 90' Azimuth 48 2.56 x 1019 Lower Shell Weld 2.5 2.21'x 1018 At 210' and 32 2.78 x 10l9 330' Azimuths 48 4.17 x 1019
'N
6.
DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS '"le methodology, RTPTS values were generated for beltline region materials of the South Texas Unit I reactor-vessel for 2.S EFPY (April 1992) 12 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 under certain conditions.
These conditions are:
Plant specific surveillance data ha; been deemed credible as defined in Regulatory Guide 1.99,- Revision 2, and RTPTS values change significantly.
(Changes to RTPTS values are considered significant if the value determined with RTPTS equations (1) and (2), or that using capsule data,-or 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 1, 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 2.5 EFPY, 32 EFPY and 48 EFPY, respectively, using the PTS Rule.
O.
TABLE 3 RTPTS VALUES FOR 50VIH TEXAS UNIT 1 2.5 EFPY 32 EFPY-48 EFPY*
Material
(*F)
-('F)
(*F)
Intermediate Shell Plate, R1606-1 60 77 80 Intermediate Shell Plate, R1606-2 50 67 70 Intermediate Shell Plate, R1606-3 63 84 87 Lower Shell Plate, R1622-1 24 45 48 Lower Shell Plate, R1622-2 33 62 66 Lower Shell Plate, R1622-3 24 45 48 Inter. Shell Longitudianl Weld 0 0*
18 35 38 Inter. Shell Longitudianl Weld 0120' 19 36 38 Inter. Shell Longitudian1 Weld 0 240' 18 35 38 Lower Shell Longitudinal Weld 0 90' 18 35 38 Lower Shell Longitudinal Weld 0 210' 21 38 40 Lower Shell Longitudinal Weld 0 330' 21 38 40 4
Circumferential Weld 1
18 20 The values for 48 EFPY are presented for information only. -..
j
i l
7.
CONCLUSIONS As shown in Table 3, the RTPTS values remain below the NRC screening values for PTS using the projected fluence values for both the end-of-life (32 EFPY) and 48 EFPY. A plot of the RTPTS values versus the fluence are shown in Figure 2 for the most limiting material, the intermediate shell plate, R1606-3, in the South Texas Unit I reactor vessel beltline region.
- n SCREENING CRITERM 2M 3%
GT
$ 150 Elx ICO 50 0
1E+18 2E+18 3E+18 SE+18 1E+19-2E+19 3E+19
$E+19 1E + 20 2
FLUENCE (n/cm )
Figure 2.
RTPTS versus Fluence Curves for South Texas Unit 1 Limiting Material - Intermediate Shell Plate, R1606-3.
8.
REFERENCES
[1]
10CFR Part 50, " Analysis of Potential Pressurired Thermal Shock Events," July 23, 1985.
[2]
10CFR Part 50,'" Fracture Toughness Requirements for Protection Against Pressurized Thermal-Shock Events," May 15, 1991.
(PTS Rule)
[3] Regulatory Guide 1.99, Revisio.n 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
[4] WCAP-12629, " Analysis of Capsule U from the Houston Lighting and Power Company South Texas Unit 1 Reactor Vessel Radiation-x
-h Surveillance Program," E. Terek, et al., August 1990.
(Westinghouse Proprietary Class 3)
[5] Combustion Engineering Material Chemistry Test Records (on file at Westinghouse, NATO).
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