ML20116C982
| ML20116C982 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/31/1992 |
| From: | Meyer T, Strauch P, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20116C965 | List: |
| References | |
| WCAP-13437, NUDOCS 9211050121 | |
| Download: ML20116C982 (14) | |
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WESTINGHOUSE PROPRIETARY CLASS 3 WCAP-13437 1
EVALUATION GF PRESSURIZED THERMAL SHOCK FOR COMANCHE PEAK UNIT 1 _
e P. L. Strauch E. Terek July 1992 Work Performed Under Shop Order WCTP-6620B Prepared by- Westinghouse Electric Corporation for Texas Utilities Approved/T.A./Meyer,Mahager
[4c by:/] d -
Structural Reliability & Plant Life OptimEzation WESTING;;0USE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1992 Westinghouse Electric Corporation All Rights Reserved l
- u. . . . .. . - . .
D L , ,
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TABLE OF CONTENTS EiLqt Table of Contents i List of Tables 11 List of Figures ii
- 1. Introduction 1
- 2. Pressurized Thermal Shock 2
- 3. 4 Method for Calculation of RTPTS
- 4. Verification of Plant-Specific Material Properties 5 e
- 5. Neutron Fluence Values 7 Determination of RTPTS Values for All Beltline
- 6. 8 Region Materials 7, Conclusions 9
- 8. References 11 e
i
.~
g-LIST OF TABLES Table Title' Pace 1.- Comanche Peak Unit:1-- Reactor Vessel Beltline Region Material 5 Properties
.2.- Neutron Expcsure Projections at Key locations on the 7 Comanchs Peak Unit l' Pressure Vessel Clad / Base Metal r -Interface for 0.91,-32 and 48-EFPY
- 3. . RTPTS Values for Comanche Peak Unit 1 for 0.91 EFPY 8
_ RTPTS Values for Comanche Peak Unit I for 32 EFPY
- 4. 8
- 5. RTPTS Values for Comanche Peak Unit I for 48 EFPY 9 LIST OF FIGURES Fiaure' Title Paae
- 1. Identification and Location of Beltline Region d
' Materials- for the Comanche' Peak Unit 1 Reactor Vessel-
- 2. 10 RTPTS versus Fluence Curves' for Comanche Peak Unit 1 Limiting Materials - Plates R1108-1 and R1108-2 11 a
y f .
+: - 1. . ' INTrt0 DUCTION
??
l Aflimiting-condition.oq reactor vessel integrity known as' Pressurized-S 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
_ simultaneous _ 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 Pressurized-Thermal Shock. It established the screening criteria for pressurized-wat'er 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 and plates and for beltline circumferential weld seams 'for end-of-license plant oper_ation. The screening. criteria were determined using conservative
- fracture mechanics- analysis techniques. _ All PWR vessels in the United States have- been_ required to be evaluated for vessel embrittlement in accordance with-the criteria through end-of-license. 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, 1991I43 This amendment.makes the procedure for calculating RTPTS' values consistent with the methods given in Regulatory Guide 1.99, Revision 2[2] ,
1
O ;
- Thk purpose of this report is to determine the Pressurized Thermal Shock reference temperature RTPTS values for the Comanche Peak UMt I reactor vessel to address the 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 Comanche Peak Unit I reactor 1 vessel. The neutron fluence values used in this analysis are presented in !
Section 5. The results of the RTPTS calculations are presented in Section G. The conclusions and references for the PTS evaluation follow in Sections 7_and.8, respectively.
- 2. PRESSURIZED THERMAL SH0CK The PTS Rule requiras that the PTS submittal be upiated whenever there are changes in core ".oadings, surveillance measurements or other information that indicatek.a significant change in projected RTPTS 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 result in 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 or renewal has been requested. This assessment must be submitted by six months after the effective date of this Rule if the value of RTPTS f r any material is projected to exceed the e
c _ .. w
- 59
(% y
.c ,
screening criteria. Otherwise, it should be submitted with the next' update of the pressure-temperature limits, or the next reactor vessel material surveillance report, or. 5 years from the effective date of this Rule, whichever comes first. These' values must be calculated based _on the-methodology ~ specified in this rule. The submittal =must include the fol_ lowing:
I) the bases _for the-projection ( Mcluding any assumptions regarding core loading patterr.),
- 2) copper and nickel content and fluence values used in the
- calculations for each beltline material. (If these valuer, differ from those previously submitted to the NRC, justification _must be' provided.)
'* The RTPTS (measure of fracture resistance) acresr.ir,9 criteria for the reactor vessel beltline recien are 270*F for plates, forgirgs, axial welds 300'F for circumferential weld materials
- .The following equations should be used to calculate the RTpTS
= values for each weld,-plate- or forging in the reactor vessel bel tline.
' Ec,oation 1: RTPTS = I + M + ARTPTS Equation 2: ARTPTS - (CF)f(0.28-0.10 log f)
- All values of RTpys 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. This information includes but is not . limited to the reactor vessel operating temperature an; surveillance results.
Results from the plant-specific surveillance program shall be integrated'into the-embrittlement estimate if, (i)- The plant-specific credible surveillance as defined data in Regulatory has been Guide deemed [2],
1.99, Rev.2 and (ii) The 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.)
m
- Plant-specific safety analysis is 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.
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,i
[ 3. , 'HETil0D FOR CALCULATION OF RTPTS fri the PTS Rule, the NRC Staff has selected a conservative and uniform n . method for determining plant-specific values of RTPTS at a given time, 1
For the purpose of comparison with the screening criteria, the value of for the reactor vessel must be calculated for each weld and plate RTPTS or forging in the beltline region as shown below.
RTPTS - I + H + ARTPTS$
where ART PTS - (CF)f(0.28-0.10 log f)
I= Initial reference temperature in *F (RTNDT) f the unirradiated material. Measured values must be used if
-credible values are available.
M- Margin to be added to cover uncertainties in the values of initial RTNOT, 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. 2 f= Best estimate neutron fluence, in units of n/cm 2 divided by 1019 (E>lMeV), at the clad / base metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period in question.
CF = Chemistry factor in *F from Tables I43 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.
04 -VERIFICATION OF PLANT-SPEC 1FIC 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 Rulel43 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:suffi. lent neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage."
Figure;l . identifies the location of all beltline region materials for the Comanche Peak Unit I reactor vessel.
A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the Comanche Peak Unit I reactor vessel are provided.in-Table 1.-
TABLE 1 LCOMANCHE PEAK UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES
- Cu Ni I-RTNDT Material Description (%) (%) ('F)
' Intermediate Shell' R1107-1 0.06 0.65 10 Intermediate 1Shell R1107-2 0.06 0.64 -10 Intermediate Shell.R1107-3 0.05 - 0.68 10 Longitudinal Weld 101-124A 0.04- 0.195 -70
- . Longitudinal Weld!101-124Bi- - 0.04 0.195 -70 Longitudinal, Weld 101-124C 0.04 0.195 -70 Circumferential Weld 101-171 _ 0.04 0.195 -70 Lower Shell' R1108-1 0.08 0.64 0 Lower Shell R1108-2 0.05 0.59 20 Lower Shell R1108-3 0.07- 0.64 0 E Longitudinal Weld 101-142A 0.04 0.195 -70 K Longitudinal-Weld 101-142B 0.04 0.195 1 Longitudinal Weld 101-142C- 0.04- 0.195 -70 L
- All' %Cu-and.%Ni values are obtained from Combustion Engineering material certifications,- except for the weld %Ni value, which is taken as the average of the the CE material certification and the surveillance capsule material analysis' [3).- Note that the -surveillance capsule weld metal
- represents all. welds in the reactor vessel.
1
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o 0 E
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Figure 1. Identification and Location of Beltline Region Materials for the Comanche Peak Unit 1 Reactor Vessel l ,
(
- 5. NEUTRON FLUENCE VALUES The calculated fast neutron fluences (E>l MeV) at the inner surface of the Comanche Peak Unit I reactor vessel are shown in Table 2 for 0.91, 32 and 48 EFPY.
TABLE 2 NEUTRON EXPOSURE PROJECTIONS
- AT KEY LOCATIONS ON THE COMANCHE PEAX UNIT 1 PRESSURE VESSL*. CLAD / BASE HETAL INTERFACE FOR 0.91, 32 AND 48 EFPY EFPY O' 25' 30' ** 45' O.91 0.0523 0.0917 0.0850 0.0921
-c 32 1.80 3.04 -2.76 2.85 48 2.70 4.56 4.14 4.29
- Fluence units are 10 l9 n/cm2 (E>1.0 MeV).
Fluence values at 0.91 EFPY are obtained from Capsule "U" dosimetry analysis [5).
Fluence values at 32 and 48 EFPY are based on design basis projections beyond 0.91 EfPY.
- 30' is the azimuthal location of longitudinal welds 101-124B,101-124C, 101-142B, and 101-142C (see figure 1).
All fluence values shown_ at the 30* azimuthal lccation were projected .
using linear interpolation between the fluence \;1ues at the 25' and 45' azimuthal locations, g.
, . . .- . ~- , - - . . . ~
- [6 (DETERMINATIONOFRTPTS VALUES FOR-ALL BELTLINE-REGION MATERIALS e
,.t. ,
T ;;Usingithe; prescribed ' PTS Rule methodology, RTPTS values were generated for Lall-beltline region materials of the. comanche Poak Unit I reactor vessel-as a Lfunction of. 0.91, l32 nand 48 EFPYLf'uence values, as shown-in Tables 3, 4-and l 5,; r.espectively. -
-TABLE 3 -
i RTPTSiVALUES FOR COMANCHE PEAK UNIT 1-FOR-0.91 EFPY
-ARTNDT(*F)-.'+ Initial RTNDT + Margin. -
RTPTS Material. .-(CF' x FF.)- (*F) (*F) (*F)
. Plate.R1107-1 15- 10 34 59 Plate R1107-f '15 -10 34 39 P1 ate R1107-3' 12 -10 34 56 Weld 101-124A 9-0.. '13 -70 56 -1 Weld 101-124B 9'30 16 56 2 Weld 101-124C 9-30 16- -70 56 2 Weld 101-171 17 * -70 56 3 P1 ate R1108 -
0 34 54
- P1 ate R1108 '12 20 34 66 P1 ate R1108-3-~ 18 0 34 52 Weld;101-142A 0.0' 13 -70 56 -l Weld:101-142B 9 30 -70 56 2 Weld 101-142C 9 30 16' -70 56 2 i*' Peak' fluence at .91' EFPY- occurs at the 45' azimuthal location.
. - TABLE 4 RTPTS VALUES FOR. COMANCHE: PEAK, UNIT 1-FOR 32 EFPY.
- ARTNDT(*F) +J -Initial RTNDT + LMargjn -
RTPTS
~
Material' (CF --:: x . . FF ) . .(.F) ('F) (*F)
Plate' R1107-1l -. 48 10- 34 92 P1 ate R1107-24 r48 -10' 34 72 P1 ate R1107-3: - 40 -10 34 84 Weld 1101-124A 9I O' 491 -70 56 35 Weld 101-124B 9 30.
- 54 --70 56- 40 Weld 101-1240 9 30. 54E -70 56 40
- Weld 101-171 55 **- -70 -56 41 P1 ate R1108 '66 0 34 100 Plate R1108-2 ' 40- 20 34- 94 P1 ate RI108-3' . 57J 0 34- 91 Weld 101-142A 9.0 35
- 49 ~70 56 L > Weld 101-1428 9 30 54 -70 56 40 Weld 101-142C 9 30: :54 -70 56 40 6** .
Peak flu'ence at 32 EFPY. occurs at the 25' azimuthal location.
~_ .,
3:
, ) ,
p" . D ,
l 1 -
TABLE 5. ';
4 1
, Il RTPTS'VALUESLFOR COMANCHE. PEAK = UNIT 1 FOR 48 EFPY- d J
f,I .
ARTNDT(=F)1+ Initial' RTNDT 1+ Margin. -
RTPTS lI Material (CFKx?FF) (*F) -- (
- F) ;'F) if P1 ate R1107-1 51- -10 34 95- 4 m -
P1 ate ' R1107-2 -' 51 -10L 34 75 i P1 ate'R1107-3 =43- 10 34 87:
.56 40
~
Weld 101-124A 9'O2 - 54' -70 '
Weld- 101-124Bl9 30'- 58 -70 56 44-Weld;101-124C 9 30 58 -70. 56 44 '
Weld 101-171 59 56 45
~
Plate R1108-1 -71. 0 34 105 Plate . R1108 43 20 -34 97 Pl ate = R1108-3 61- 0 34 95 a Weld 101-142A 0-0 '54 -70 56 40 Weld 101-1428 9 30" 58 - -70 56 - 44 z Weld,-101-142C 9 30' -58 -70 56 44
-7.- : CONCLUSIONS-1
~
g -.As shown in Tables 3, 4 and 5, .all RTPTS values remain below the NRC
- : screening criteriatfor PTS using the Capsule'"U" fluences for 0.91 EFPY, and .
_ the; projected design fluences for 32 and 48 EFPY.
LA plotiof th'e: RTPTS values versus' the: fluence is- shown in Figure 2 for the
'most limiting materials-inithe Comanche ; Peak Unit I reactor vessel beltline region,~ plates 1R? led-1 and R1108-2.
w t
)
c u -
-9
~
- , i . _
+
, 3 ,
e pa a; .,
-?/O 4
p n:
300-SCREENING CRITERIA 270*F for plates. forgings axial welds eso -
e 200 --
..C o
.v:
e : g:1So g;
a- 7
- 1oo -.
. r r.r.r.$ . .k .~.~. . . .
........................z.u.a.,..
So O
1E+20-
.1E + 18 - 2E+18 3E+18 SE+18 : ' 1E+ 19 2E+19 3E+10 SE+19 2
FLUENCE (NEOTRONS/cm )
R1108-1 R1108 ! 32 EFPY 48 EFPY
. . . . ........ 6-- A ,
e < ,
P Figure ?.. 'RT versus Fluence Curves for Comanche Peak Unit 1 LimIk:ing _ Materials -- Plates R1108-1 and R1108-2
,, [ y -
C
'h. - . . , ,
L
+
! 8. REFERENCES
[1] 10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events", July 23, 1985.
_ [2]l Regulatory Guide 1.99, Revisien 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Conmission, May 1988.
r
[3] WCAP-9475,:" Texas Utilities Comanche Peak Unit.No. 1 Reactor Vessel Radiation Surveillance Program", W. T. Kaiser, et al.,
April 1979. (Westinghouse Proprietary Class III)
[4] 10CFR Part 50.61, " Fracture Toughness Requirements for j Protection Against Pressurized Thermal Shock Events", -ay 15, i
1991. (PTE Rule)
[5]- Westinghouse Report FSE/P.EA-174/92, " Comanche Peak Unit 1 -
Capsule U Neutron Dosimetry", S. L. Anderson, June 17,1992.
l t
5