ML13324A794
| ML13324A794 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/23/1986 |
| From: | Medford M Southern California Edison Co |
| To: | Lear G Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TAC-59980, TASK-A-49, TASK-OR NUDOCS 8601270004 | |
| Download: ML13324A794 (7) | |
Text
Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 M.O.MEDFORD TELEPHONE MANAGER. NUCLEAR LICENSING January 23, 1986 (818) 302-1749 Director of Nuclear Reactor Regulation Attention:
Mr. George E. Lear, Director PWR Project Directorate No. 1 Division of PWR Licensing A U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:
Subject:
Docket No.
50-206 Pressurized Thermal Shock San Onofre Nuclear Generating Station Unit 1 On August 22, 1985 a new 10CFR5O.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, went into effect.
The new regulation (10 CFR 50.61(b)(1)) requires that each PWR licensee submit projected values of RTPTS of reactor vessel beltline materials from the time of submittal to the expiration date of the operating license. The submittal must specify the bases for the projected values and include assumptions regarding core loading patterns. The regulation (10 CFR 50.61(b)(2)) also provides the method on how the value of RTPTS is to be calculated.
The screening criteria of RTPTS is 270oF for plates, forgings and axial weld material or 300OF for circumferential weld material.
Provided as an enclosure is the assessment of the San Onofre Unit 1 reactor vessel beltline materials. Current values and values at the operating license expiration date of RTPTS are provided. The enclosure describes the bases for the projected values and discusses the assumptions made. As indicated in the enclosure the projected values of RTPTS do not exceed the 10 CFR 50.61(b)(2) screening criteria. Consequently, no further action regarding 10CFR50.61 is required for the San Onofre Unit 1 reactor vessel.
This submittal is required to be updated in the event that changes in core loadings, surveillance measurements, or other informafion indicate a significant change in projected values.
If you have any questions regarding this information, please let me know.
2601270004 860123 Sincerely, PDR ADOCK 05000206 P
PDR Enclosure AD - J. Knight (1tr only) cc:
R. Dudley, NRC Project Manager EB (BALLARD)
F. R. Huey, NRC Site Inspector EICSB (ROSA)
PSB (GAMMILL)
RSB (BERLINGER)
FOB (BENAROYA)
Enclosure SCE RESPONSE FOR THE SAN ONOFRE UNIT 1 REACTOR VESSEL TO 10 CFR 50.61 PTS RULE REQUIREMENTS 10 CFR 50.61 describes the Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events. Specifically, Paragraph (b)(1) requires the licensee of each PWR for which an operating license has been issued to submit projected values of RTPTS at the inner vessel surface of the reactor vessel beltline materials from the time of submittal to the expiration of the operating license.
Paragraph (b)(2) provides the PTS screening criterion and prescribes the method by which the values of RTPTS must be calculated. The method specified is as follows.
For each weld, plate or forging in the reactor vessel beltline, RTPTS is the lower of the results given by Equations 1 and 2.
Equation 1:
RTPTS = I + M + [-10 + 470 Cu + 350 Cu Ni]f 0.27 Equation 2:
RTPTS = I + M + 283f0.194 where:
I =
initial reference temperature of the unirradiated material measured as defined in the ASME Code Paragraph NB-2331.
M = margin to be added to cover uncertainties in the values of initial RTNDT, copper and nickel content, fluence and the calculational procedures Cu, Ni = best estimate weight percent of copper and nickel in the material f = best estimate neutron fluence in units of 1019 n/cm 2 (E greater than or equal to 1 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 fluence for the period of service in question.
The projected values of RTPTS are tabulated in Table 1. These values were calculated utilizing Equation (1) since it produced the lower RTPTS. The identification and location of reactor vessel beltline material of the San Onofre Unit 1 vessel are shown in Figure 2-1.
The bases for the projections are discussed in more detail below.
-2 I and M Values of the initial reference temperature (I) of reactor vessel beltline materials are documented in Reference 2. Measured values of initial reference temperature were available for all shell plates. Measured data was not available for welds. Therefore, a generic mean value of -560F was used as prescribed by 10 CFR 50.61(b)(2)(1) for welds made with ARCOSB-5 weld flux.
San Onofre Unit 1 reactor vessel beltline welds were made with ARCOSB-5 flux per Reference 1. The margin to be added to cover uncertainties (M) was 480F where a measured value of I wasavailable and 590F where a generic mean value of I was used as prescribed by 10 CFR 50.61(b)(2)(ii).
Copper and Nickel Content The copper content of the reactor vessel shell plates is given in Reference 2. The nickel content of the shell plates, while not directly measured, is known to be less than 0.20 weight percent since the SA302B steel utilized for the plates is known to contain less than 0.20 weight percent nickel.
Records of the Cu and Ni content of the reactor vessel beltline welds are not available. However, on the basis of information provided in References 1 and 3 values of 0.27% copper and 0.20% nickel are conservatively estimated for the San Onofre Unit 1 reactor vessel beltline welds.
Also, the use of 0.27%
copper and 0.20% nickel in Equation 1 results in a trend curve which bounds all known surveillance data for vessels made in the same time period as the San Onofre Unit 1 reactor vessel.
Hence, these values were used in the determination of the projected RTPTS values for the reactor vessel beltline welds and are considered conservative.
Fluence f The analytical methodology and the design basis used to predict time averaged fast neutron flux and fluence levels at the vessel inside surface have been discussed in some detail in Reference 3. Plant specific results of neutron transport calculations for the San Onofre Unit 1 reactor vessel are presented in Reference 1 and are fully consistent with the methodology outlined in Reference 3. Using the information in Reference 1, the neutron flux on the inside vessel surface at each weld and plate location in the reactor vessel beltline region can be determined. The fluence per EFPY in Table 1 is determined by time averaging the neutron flux over one effective full power year. As discussed in Reference 1 the design basis core power distributions used in the computation of time averaged neutron flux and long term neutron fluence levels are derived from a statistical analysis of calculated distributions from all available independent cycles of a given reactor type and have proven to yield satisfactory results for long term fluence predictions. Credit was not taken for the flux reduction effect of low leakage loading patterns implemented in Cycles 7 and 8 of San Onofre Unit 1 which would have the effect, if continued, of lowering the long term neutron fluence predictions.
-3 Prolected Values of RTPTS Values of RTPTS in Table 1 are given from the time of submittal to the expiration date of the operating license. The San Onofre Unit 1 reactor vessel as of January 23, 1986 has 10.12 EFPY. San Onofre Unit 1 is currently operating on a Provisional Operating License issued March 27, 1967.
Based on an equilibrium cycle length of 484 EFPD and a production factor of 75% with an assumed 90 day refueling outage, the EFPY on March 27, 2007 is 24.05 EFPY.
Conclusion The projected values of RTPTS for all beltline materials as indicated in Table 1 do not exceed the PTS screening criterion before the expected expiration date of the operating license. Hence, the actions required by 10 CFR 50.61(b)(3) and (b)(4) are not applicable. The results indicate that reactor vessel integrity will be maintained for the San Onofre Unit 1 reactor vessel throughout its service life.
References
- 1.
Letter SCE to NRC, Docket No. 50-206, Pressurized Thermal Shock to Reactor Pressure Vessels, San Onofre Unit 1, dated January 25, 1982
- 2.
Yanichko, S. E.; Anderson, S. L.; and Kaiser, W. T.; "Analysis of Capsule F from the Southern California Edison Company San Onofre Reactor Vessel Radiation Surveillance Program," Westinghouse Electric Corporation, WCAP-9520, May 1979
- 3.
Summary Report on Reactor Vessel Integrity for Westinghouse Operating Plants, WCAP-10019, dated December 1981 WGF:5808F
TABLE 1 CURRENT AND PROJECTED VALUES OF RTPTS OF REACTOR VESSEL BELTLINE MATERIALS SAN ONOFRE UNIT 1 Initial 10.12 EFPY 24.05 EFPY Plate or Cu Ni(a) RTN T (b) M(c)
Fluence/EFPY ARTPTS RTPTS ARTPTS RTPTS Component Seam No.
(%/
(%D
_oOD (1018 n/cm2j
.1of.L Of) oF o)
Inter. Shell W7601-1
.17
.20 60 48 3.69 117 225 147 255 Inter. Shell W7601-8
.18
.20 40 48 3.69 124 212 157 245 Inter. Shell W7601-9
.18
.20 55 48 3.69 124 227 157 260 Upper Shell W7601-3
.15
.20 8
48 2.21 88 144 111 167 Upper Shell W7601-6
.16
.20 24 48 2.21 95 167 120 192 Upper Shell W7601-7
.15
.20 12 48 2.21 88 148 111 171 Lower Shell W7601-2
.17
.20 34 48 1.11 84 166 107 189 Lower Shell W7601-4
.14
.20 51 48 1.11 68 167 86 185 Lower Shell W7601-5
.14
.20 82 48 1.11 68 198 86 216 Long. Weld 7-860A
.27(a)
.20
-56 59 3.00 183 186 232 235 Long. Weld 7-8608
. 27(a)
.20
-56 59 1.20 143 146 181 184 Long. Weld 7-860C
.27(a)
.20
-56 59
.695 123 126 156 159 Long. Weld 6-860A
.27(a)
.20
-56 59
.720 125 128 158 161 Long. Weld 6-8608
.27(a)
.20
-56 59
.417 108 111 136 139 Long. Weld 6-860C
.27(a)
.20
-56 59 1.80 160 163 202 205 Long. Weld n-860A.27(a)
.20
-56 59
.360 103 106 131 134
-2 Initial 10.12 EFPY 24.05 EFPY Plate or Cu Ni(a) RTNDT (b) M(c)
Fluence/EFPY ARTPTS RTPTS ARTPTS RTPTS Component Seam No.
NOL jLFf
_oFL
((0/8 n/cm2 i
_oL_
_Cf__o ofj Long.
Weld 8-860B
. 2 7(a)
.20
-56 59
.208 89 92 113 116 Long. Weld 8-860C
. 2 7(a)
.20
-56 59
.900 132 135 167 170 Circum. Weld 2-860
.27 (a)
.20
-56 59 2.21 169 172 213 216 Circum. Weld 1-860
.27(a)
.20
-56 59 1.11 140 143 177 180 (a) Conservative estimate -
no chemical analysis available (b) Initial reference temperature-measured value or generic mean value per 10 CFR 50.61(b)(2)(i)
(c) Margin to be added to cover uncertainties per 10 CFR 50.61(b)(2)(ii)
WGF:5808F
FIGURE 2-1 IDENTIFICATION AND LOCATION OF BELTLINE REGIOI0 MATERIAL OF THE SAN ONOFRE REACTOR VESSEL S
170 U.*.
- 0A 6
CAGO Go G
s-860 W7601-8 CORC o-8000 A-7a-SOo ISI U.-7*860C W7&
I-'
617601I-'q 1 4 6 0j-SSg A
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