ML20101M337
| ML20101M337 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/01/1992 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-92-01, GL-92-1, NSD920629, NUDOCS 9207080182 | |
| Download: ML20101M337 (21) | |
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GENE 4AL OF FtCE r=.r P O DOX 499. COLUMBUS. NEBRASKA f.Me2(A99
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NSD920629 July 1, 1992 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C.
20555 Gentlemen:
Subject:
Response to Generic Letter 92-01, Revision 1 Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46
Reference:
Letter from J.
G.
Partlow (NRC) to all Liccusees dated March 6, 1992, " Reactor Vessel Structural Intcgrity,10 CFR 50.54(f) (Generic Lecter 92-01, Revision 1)"
The Nebraska Public Power District (District) hereby provides its response to Generic Letter 92-01, Revision 1 issued March 6, 1992 (Reference).
Generic Letter 92-01 Revision 1 supersedes Generic Letter 92-01 which was issued February 28, 1992.
Generic Letter 92-01, Revision 1 requests licensees
> provide information to demonstrate compliance with 10 CFR 50.60, " Acceptance C. tteria for Fracture Prevention Measures for Lightwater Reactors for Normal Operau ton," and 10 CFR 50 Appendices G, " Fracture Toughness Requirements," and H, " Reactor Vessel Material Surveillance Program Requirements."
The District has verified that Cooper Nuclear Station (CNS) meets the requirements of these regulations.
The District's responses to the Generic Letter 92-01, Revision 1 questions are provided la the attachment.
As requested in Generic Letter 92-01 Revision 1 and pursuant to 10 CFR 50.54(f),
this response is submitted under oath.
Please contact me if you have any questions or require any additional information.
Sin erel if Cum GIR Horn Nuc car Power Group Manager GRH:MJB/MHM Attachn2nt cc:
NRC Regional Office Region IV Arlington, TX NRC Resident Inspector Cooper Nuclear Station 9207080182 920701 Ag M PDR ADOCK 05000298 (7
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-C.-R.. Horn,--being.first. duly sworn, deposes and says that-he.is an authorized
-representative of the Nebraska Public Power District, a public-corporation and political subdivision of the State of Nebraska; that he_is duly authorized to submit this response on behalf of Nebraska Public Power District; and that the i
statements contained herein are true to the best of his knowledge and belief.
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O G. R. Horn l-Subs ibed in my presence and sworn to beforo n.e this ILJ day of dt -
, 1992.
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Page 1 of 12 RESPONSE.TO.CENERIC LETTER 92-01 REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY I.
INTRODUCTION The information in this attachment provides the Nebraska Public Power District's. (District's) response to. Generic Letter 92-01, Revision 1,
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"Rorotor Vessel-Structural Integrity."l' Generic-Letter 92-01, Revision 1 replaced Generic Letter 92-01, dated February 28, 1992,II Generic Letter 92-01 Revision 1 clarified - some of the information
-i pertaining to the review i:' the Yankee Fuclear Power Station structural integrity, but.did not change the information requirements.
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Generic Letter 92-01, Revision 1 requests licensees to respond to a number L
of questions intended to verify that licensees have adequately addressed the. requirements contained in 10 CFR-50. 60, " Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation,"
" Fracture Toughness Requirements," 10 CFR 50 Appendix H, " Reactor Vessel Material Surveillance Program Requirements,"'and for Pressurized Water Reactors, 10 CFR 50.61,
" Fracture' toughness requirements. for protection against pressurized thermal shock events." - The following discussion provides the District's response to Generic Letter 92-01, Revision 1; II.
DISCUSSION o
The District's' response to the specific information requirements cont 11ned 3
in Generic Letter 92-01, Revision 1 is provided below.
E Reauest 1 Addressees who do not have a surveillance program meeting ASTM E 185 73,
-79, or _ -82 and who do _ not have an integrated surveillance program approved by the NRC-(see Enclosure 2), are requested to describe actions
_taken or to be taken to ensure compliance with Appendix H to 10 CFR Part 50.
Addressies who plan to revise the surveillance program to meet
. Appendix H to 10 CFR Part 50 are requested to indicate when the revised program -will be. submitted to the NRC staff for review.
If the surveillance program is not to be revised to meet Appendix H to 10 CFR l'
Letter from J. G.
Partlow (NRC) to all Licensees dated March 6, 1992, " Reactor Vessel Structural Integrity, 10 CFR 50.54(f)
(Generic Letter 92-01, Revision 1)."
l' Letter from J..G. Partlow (NRC) to all Licensees dated February-28, 1992, " Reactor Vessel Structural Integrity, 10 CFR 50.54(f)
(Generie Letter 92-01)."
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At_achment to NSD920629 Page 2 of 12 Part 50, addressees are requested to indicate when they plan to request an exemption from' Appendix H to 10 CFR Part 50 under-10 CFR 50.60(b).
District Responrd r-Although Cooper Nuclear _ Station (CNS) is not listed in Enclosure 2 of Generic Letter 92-01, Revision 1, it is the District's position that its surveillance program for Cooper Nuclear Station is currently in compliance with 10 CFR 50 Appendix H, and has been acknowledged previously by the NRC as being in compliance with Appendix H.
10 CFR 50 Appendix H states:
"That part of the surveillance pogram conducted prior to'the-first withdrawal must meet the requirements of the edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased "
Additionally,' Appendix H states:
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_"For each capsule, vi.thdrawn af ter July 26, 1983, the test procedures and reporting requirements must meet the requirements of ASTM E 185-82 to the extent practical for the configurations of the specimens in the capsule."
1 The CNS vessel was designed to the Winter 1966 Addends of the 1965 ASME Code..
ASTM E 185-66 was the standard in place at tb-time the-g sweveillance program was. designed.
Therefore, the vesu. material L
surveillance specimens were fabricated and located v'. thin the CNS vessel in accordance with that guidance, and accordingly, the fabrication and installation of the CNS vessel surveillance specimens comply with the requirements of 10 CFR 50 Append!x H.
l-The first surveillance capsule was withdrawn during the CNS Reload.9 Cycle l
10- refueling outage in 1985 following 6.8 ' Effective. Full Power Years (EFPY) of operation, and tested, to the extent practical, in accordance with-ASTM E 185-82. 'Following testing and analysis of the-surveillance
~ specimens,-the District submitted a proposed change to the-CNS Technical-o h
~ Specifications to revise the CNS pressure-temperature operating limit j;
curves accordingly. In its safety evaluation approving Amendment No. '120 to the CNS operating licenseF the NRC recommended, to meet ASTM E 182-82 as closely as' possible,. that the District accelerate the withdrawal
. schedule. of _ the second surveillance capsule to 12 EFPY, and consider insertion of a fourth capsule into the. CNS vessel, possibly with
. reconstituted specimens'from an earlier capsule.
F Letter from W. O. Long (NRC) to C. A Trevors (NPPD) dated April' 26, 1983, " Cooper Nuclear Station - Amendment No. 120 To Facility Operation ~ License No. DPR-46 (TAC No. 65793)."
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Attachment to NSD920629 Page 3 of 12
,1 Following various communications, and in support of a license change
-application to extend the~ operating license expiration date to 40 years from receipt of the CNS - operating license, - the District committed to 1)_ remove _the second surveillance capsule _during the Reload 14, Cycle 15
-refueling ' outage during 1991. (following approximately 11 EFPY of operation),-and 2) reconstitute the specimens from this capsule and re-insert the reconstituted specimens during the Reload 15, Cycle 16 refueling outage.
The District also indicated that the withdrawal schedule for the third capsule will be based on the reults of testing the L
.second surveillance capsule.i' In _. Its safety evaluation accompanying Amendment No.
143 to the CNS operating -license which extended the CNS license expiration date to
[.
January 18, 2014,l' the NRC acknowledged the District's commitment to l
reconstitute the surveillance capsule withdrawn during the 1991 refueling outage.
The NRC stated further that the reconstitution of the capsule withdr.swn in 1991 is equivalent to a fourth capsule and thereby makes the District's surveillance program consistent with the requirements of ASTM E-185-82 and 10 CFR Part 50 Appendix H.
The NRC also acknowledged that the withdrawal schedule for the original -third capsule and the reconstituted fourth capsule should be based on the results of the
' analysis of the second capsule.
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The District has withdrawn the second surveillance capsule from the CNS l
reactor vessel; these specimens are currently undergoing testing and analysis, and will be reconstituted and reinserted in the CNS reactor-vessel during the Reload 15 Cycle 16 currently schedulei to commence in March 1993.
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Therefore,- due-to changes in ASTM E 185 surveillance specimen fabrication requirements -- between-1966, and 1973, the District's vessel material surveillance program for CNS does not strictly comply with ASTM E 185-73 or later revisions.
However, as discussed - above and as acknowledged previously by the NRC, the CNS surveillance program does meet, as closely as possible, the requirements of ASTM E 185-82, and therefore is in compliance with 10 CFR 50: Appendix H.
Further details on the CNS surveillance program are provided in the responses to Requests 2.b.4 and 2.b.5.
l' In addition,-the District' continues to participate in industry-sponsored efforts to gather additional data on radiation embrittlement.
The II Letter from G. R. Horn (NPPD) to NRC dated June 7, 1991, " Response l
to Questions on License Extension to 40 Years from Operating
-License Issuance."
l' Letter from P. V. O'Connor (NRC) to C. R. Horn dated July 5,1991,
" Cooper Nuclear Station - Amendment No. 143 to Facility Operating License No. DPR-46 (TAC No. 74843)."
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Attachment to NSD920629 Page 4 of 12 District is a member of the Boiling Water Reactor Owners' Group (BWROG)-
Supplemental Surv+111ance Program (SSP)
Committee, and CNS is participating as a host reactor. The objective of the SSP is to develop supplemental surveillance data which will allow the District to better understand the extent of beltline embrittlement with increasing fluence for
- a. variety of plate and weld materials.
The first su,3plemental surveillnnce capsule was installed in CNS during ira 1991 Refueling Outage.
The testing being undertaken by the SSP Committee will greatly increase the BWR surveillance data base.
Descriptions of the SSP test program and hardware are presented in the Committee's Phase 1 report and Phase 2 Progress report.
These reports are being prepared for submittal to the NRC in the near future.
Reauest 2.a Addressees of plants for which the Charpy upper shelf energy is predicted to be.less than 50 foot-pounds at the end of their licenses using the guidance in Paragraphs C.1.2 or C.2.2 in Regulatory Guide 1.99, Revision 2,
are requested to provide to the NRC the Charpy upper shelf energy predicted for December 16, 1991, and for the end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken pursuaat to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.
District Response The Upper Shelf Energies (USE) of the beltline materials at CNS are not expected to be less than 50 ft-lb by the end of its licensed operating period. The following discussion provides a brief description of the USE evaluation supporting this conclusion.
USE data were taken during fabrication of the plates used in the CNS beltline. Evaluation of these plates according to Regulatory Guide 1.99, Revision 2, shows the USE at 32 Effective Full Power Years (EFPY) to be above 50 f t-lb.
For the beltline submerged arc welds, fabrication USE data were not taken; rather, fabrication Charpy tests were performed at 10'F with a 30 f t-lb requirement.
However, testing performed c 3 the vessel surveillance weld, which is representative of the beltline welds, demonstrated an unirradiated USE of 112 ft lb.
Evaluation according to s
Regulatory Guide 1.99, Revision 2 of the beltline welds assuming initial USE of 112 ft-lb results in 32 EFPY USE predictions well above 50 ft-lb.
.-In addition,.the District is a member of the BWROG RPV Fracture Toughness Committee.
The objective of that committee is to ' develop methods to estimate USE in cases where plant beltline materials were tested at only transition temperatures; the methods being developed will greatly increase the BWR fracture toughness data base.
A description of the method is
Attachment to NSD920629 Page 5 of 12 presented in the BWROC Fracture Toughness Committee Report which was transmitted to the NRC in June, 1992.i' Recuest 2.b b.
Addressees whose reactor vessels were constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the following material properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.A of 10 CFR Part 50, Appendix G:
Suboart 2.b.(1)
(1) the results from all Charpy and drop weight tests for all unirradiated beltline materials, the unirradiated reference temperature for each beltline,aterial, and the method of determining the unirradiated refe
.e temperature from the Charpy and drop weight test; District Response As indicated above, the CNS vessel was fabricated to the Winter 1966 Addenda of the 1965 ASME Code.
For the CNS vessel beltline plate materials, Charpy and dropweight tests were performed.
The Charpy specimen orientation was longitudinal and the tosc requirement was to meet 30 ft-lb at the specified temperature.
In order to d;monstrate fracture toughness equivalent to Appendix G requirements, a General Electric procedure, described in the Cooper Nuclear Station Reactor Pressure Vessel l#
Surv 111ance Materials Testing and Fracture Toughness Analfsis report was _,ed to adjust the 30 f t-lb longitudinal Charpy data to determine the temperature T a7 at which an equivalent 50 ft-lb transverse Charpy energy 3
ceuld be expected. The unirradiated FT,37 was then selected as the higher of (Tsot-60*F) or the dropweight Nil-Ductility Temperature (NDT).
For the beltline weld materials, only Charpy tests were performed.
The specimens were cut transverse to the weld length and the test requirement was 30 f t-lb at 10*F.
As with the plate, the GE procedure was used to adj us t the 30 ft-lb Charpy data to determine T and to account for the 307 absence of dropweight testing data. The unirradiated RT was determined y37 from the procedure as the higher of either (T307-60*F) or -50*F.
i' Letter from C. L. Tully (BWROG) to NRC dated June 12, 1992, "BWR Owners' Group Submittal on Upper Shelf Enerby Estimation."
Il Report MDE-103-0986, dated May 1987, " Cooper Nuclear Station Renctor Pressure Vessel Surveillance Materials Testing and Fracture Toughness Analysis," submitted to the NRC by letter from G. A. Trevors (NPPD) to NRC dated July 6,1987, " Reactor Vessel Material Surveillance Program, NRC Docket No. 50-298/DPR-46."
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Attachment to NSD920629 Page 6 of 12 Charpy data, dropweight test results and -estimated RTm values for the beltline' materials are provided in Appendix A of this attachment.
l Subpa:!t 2.b.(21 (2).
the. heat treatment received by all beltline and surveillance materials; District Response Heat treatment was not explicitly considered in the Appendix G analysis, as - there are no requirements or methods provided which relate to heat treatment. Ilowever, implicit in the Appendix C analysis is the assumption that-the Charpy data used to develop the RTm values-is representative of
.the beltline materials,-so the heat treatment of Charpy specimens should represent or bound that of the ' beltline materials.
After the beltline plates were quenched and tempered, specimen samples and plate _ used in the surveillance program were trimmed from the plates. The specimen samples and surveillance materials received a simulated Post-Weld Heat Treatment (PWHT) at 1150*Fi25'F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. The beltline material PWHT-temperature was
.the-same,. but the beltline PWHT time was significantly less.
The additional PWHT time - for the specimens was
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intended to. cover the. possibility of future vessel repairs requiring subsequent PWHT.
Since the surveillance specimen-PVHT was as long or--
longer than the beltline PWHT time periods, the surveillance specimens and the corresponding surveillance data are representative of the vessel beltline mat rials.
Subcart 2;b.(3)
(3) the-heat number for inch beltlico
' ste or forging and the heat
. number of wire and ff.x loc number 3 to fabricate each beltline weld; District Response The beltline consists of portions of the lower shell and lower-intermediate shell, Each shell is formed from three plates, so the
- beltline includes portions of six plates, six vertical welds and one girth
. weld.
All beltline plate and weld materials were considered in the Appendix G Evaluation. The requested information is provided in Appendix A to this letter.
The. District is continuing its search.through vessel fabrication records to determine the material chemistries for the beltline l
weld' materials identified in Appendix A as "not available."
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Page 7 of 12 Subnart 2.b.(4)
(4)
- thel heat number for each surveillance plate or forging and the heat number-of wire-and flux lot number used to fabricate the surveillance veld; District Response d
Appendix C includes, by reference'in para raph III.A, Appendix H which includes by reference "the requirements of che edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased. " For CNS, this 'is ASTM E 185-66, which required that " test specimens shall be taken from materials used in the irradiated region."
ASTM E 185-66 further states that " Samples shall reprasent one heat of the base metal and one butt weld if a weld occurs in the irradiated region."
The surveillance plate material was trimmed'from the beltline plate with
-heat number C2307-2. The District is continuing its efforts, through the review'of the original CNS. vessel fabrication documentation to determine the -specific wire heat number and flux. lot data for the surveillance weld.
Notwithstanding this effort, the specification for the surveillance weld
. required. that it be made with - the same procedure as the longitudinal beltline weld 1-233, i.e.,
MIL-E-18193 type B-4 modified wire with flux type Linde 1092, same wire feed rate, heat input, etc.
Therefore, the data available on the surveillance weld indicates that it is representative. of the. beltline welds, as required by ASTM E 185-66. More importantly, the available surveillance weld data provides the District all of the-information needed to meet the objective of Appendix H to monitor toughness changes due to irradiation.
l The usefulness of the-results from surveillance weld testing is augmented by the following:
Archive surveillance weld material has been tested, providing complete baseline Charpy curve and chemical composition data; this information was provided to the NRC in Report MDE 103-0986, which was transmitted to the.NRC by letter-dated July 6,1987,l'
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. Irradiated surveillance weld Charpy specimen test curves can be
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compared credibly with the baseline Charpy curve.
The copper and nickel content are known and the fluence is established from dosimetry in each. surveillance capsule.
Therefore, all necessary information is available to compare
.the surveillance weld
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Letter from C. A. Trevors (NPPD) to NRC dated July 6,1987,
" Reactor Vessel Surveillance Program, NRC Docket No. 50-298, DPR-46."
E Attachment..
to NSD920629 Page 8 of 12 irradiation embrittlement with Regulatory Guide 1.99, Revision-2 predictions.
Suboart ' 2'.b. ( 5)
-(5)-
the chemical' composition, in particular.the weight in percent of copper, nickel,. phosphorous, and sulfur for each beltline and surveillance. material; and District Response Chemical composition weight percent data for beltline materials are shown
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in Appendix A to this letter.
In some cases for weld materials, upper bound assumptions of 0.35% copper and 1.0% nickel were used in the absence of actt.al_ chemistry data.
Beltline material chemistries, or upper bound assumptions, were used with Regulatory Guide 1.99, Revision 2 to determine the limiting beltline material, the adjusted reference temperature versus
.EFFY for the material, and the predicted plate USE at 32 EFPY.
Chemical composition weight percent data for the surveillance plate and weld are shown in Appendix B to this letter.
The chemical composition data were used to comnare measured Charpy curve shif ts with Regulatory
~ Guide l'.99, Revision 2 predictions. The surveillance weld chemistry was used to estimate weld USE at 32 EFPY.
Suboart 2.b.(6)
(6)-
the heat number of the wire used for determining the weld metal chemical composition if different than Item (3) above.
District Response This does not apply to CNS; see the response to Item 2.b.(3).above.
Recuest 3-
-Addressees are requested to provide the following information regarding commitments made to respond to GL 88-11:
Suboart 3.a-a.
_How the embrittlement effects of operating at an irradiation temperature (cold. leg or recirculation suction temperature) below 525*F.were considered.
In particular licensees are requested to describe _ consideration given to determining the effect of lower l~-
-irradiation temperature -en the reference _ temperature and on the Charpy-upper shelf energy.
Attachment to NSD920629 Page 9 of 12 1
District Response i
l Operation with the CNS beltline. region below 525'F was not considered'in the Appendix G analysis because the -steady state operating. temperature of the coolant in the beltline region is slightly greater.
Based on the temperature in _the recirculation suction piping (which draws water directly from. the beltline region) the steady state temperature in the beltline is greater than 527'F.
'Only during start-up and operation without feedwater heating, which occurs when feedwater heaters are out of service or when the turbine is off line and the reactor steam is routed through che turbine bypass, does the beltline experience coolant less then 525'F when the core is critical.
The equivalent full power ~ time cf operation in these transient conditions
- has been estimated to be less than 1%, and the associated temperatures for' 1
-most of that time are 515'F or higher. The CNS 32 EFPY fluence is 1.5x10 s n/ca', so the fluence accumulated below 525'F would be approximately 1.5x10" n/cm*; This combination cf low fluence and small deviation from the 525'F level is not expected to significantly affect beltline RTm or USE predictions.
6 Since-surveillance specimens are exposed to the same temperature conditions as the beltline materials, temperature effects, if any, will be reflected in the surveillance results. When the surveillance results are factored into the Appendix G analysis per Regulatory Guide 1.99, Revision 2, temperature effects, if any, will be acco inted for inherently.
In addition, the BWROG Supplemental Surveillance Program will collect data periodically over the next 8 years.
The capsules will include eutectic temperature monitors with which to determine the appropriate maximum irradiation. temperature.
The data will include results of some PWR
- materials and 11SST-02 standard-material.
These data should provide additional insight into temperature effect differences between the BWR and P
PWR environments.
Suboart 3.b b.
11ow. their surveillance results on the predicted amount of embrittlement were considered.
- District Response Surveillance results were factored i
- 3eltline embrittlement predictions following receipt of results of tect.. r.he first surveillance capsule.
At this; time, Regulatory Guide 1.99, :.evision 1 was the current guidance.
As.a result, the District submitted Proposed Change No. 48 to the CNS 1.
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Attachment to NSD920629 Page 10 of 12
- Technical Specifications 2',
which provided re'rised pressure temperature limitation curves based on the methods provided in Regulatory Guide 1.99, Revision 1,
and adj us ted conservatively for the results of the surveillance testing.
In response to a reque6e for additional information, the District submitted to the NRC additional details to support the proposed change.MI The NRC evaluated the proposed change against Regulatory Cuide 1.99, Revision 2, which was approved at that time, although awaiting publication as a final guide.
As a result, the NRr issued Amendment No. 120 to the CNS Operating License,ll' concluding in its safety evaluation:
"Since the ART [ Adjusted Reference Temperature) values used by the licensee to calculate the proposed Pressure-Temperature limits are greater than the values predicted using the formula in R.G.
1.99, Rev. 2 the proposed Pressure-Temperature limits will meet R.G.
1.99, Rev.
2."
Additionally, the safety evaluation concluded:
"To confirm that the Pressure Temperature limits proposed by the licensee will meet the safety marg ns of i
Appendix G, 10 CFR Part 50 for the proposed operating
-periods, the staff has used the method of calculating Pressure Temperature limits in USNR0 Standard Review Plan 5.3.2, NUREG 0800, Rev. 1, July 1981 to evaluate the proposco Pressure Temperature limits. The staff's calculation includes the licensee's ART values.
Our calculations confirm that the proposed Pressure.
Temperature limits meet the safety margins of Appendix i
G, 1C CFR Part 50 for the operating periods identified on the curves "
Therefere, using Regulatory Guide 1.99 Revision 1 guidance, adjusted conservatively for measured reference t-speratura shift, the Dietrict 21 Let'er from L, G. Runc1 (NPPD) to NRC dated Octc a r 28, 1987,
" Proposed Change No. 48 to the Cooper Nuclear Station Technical Specifications, NRC Docket O 50-298, DFu-46" M'
Letter from G. A. Trevors to NRC dated February 22, 1988, "Suppi mental Submittal; Proposed Change No. 48 to the Cooper l_
Nuc1(nr Station Technical Specifications NRC Docket No.56-298, DPR c M'
Letter from W. O. Lcog (NRC) to G. A. Trevors (NPPD) dated April 26, 1988, " Cooper Nuclear Station + Amendment No. 120 to Facility Operation Lic:ense No. DPR 46 (TAC No. 65793)."
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.j Attachment i
to NSD920629 Page 11 of 12 l
-demonstrated that - the CNS operating limits enveloped the constraints l
proviued by 10 CFR Appendix G.
I Regulatory Guide 1.99, Revision 2, paragraph C.2, requires credible data from two surveillance capsules before adjustments to the predictions methods are made.
Only one capsule has been tested: the second surveillance capsule is currently undergoing testing and analysis.
Therefore, although the current beltline predictions are based on Regulatory Guide 1.99, Revision 1 methods adjusted cont.orvatively for
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measured reference temperature shif t, these predictions have been shown to be conservative with respect to Regulatory Guide 1,99, Revision 2 methods.
The District will reevaluate these predictions when the results of testing the second surveillance capsule become available.
.t Subpart 3.c c.
If a measured increase in reference temperature exceeds the mean-plus two standard deviations predicted by Regulatory Guide 1.99, 1
Revision 2; or if a measured decrease in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.1.2 in.
4 Regulatory Guide 1.99, Revision 2i the licensee is requested-to report the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as pr$dicted for December 16, 1991, and for the end of its current licer.se.
District' Response i
Measured increases'in reference temperature and measured decreases in USE based on the first surveillance capsule are provided in Appendix C to this letter.
Measured increases in reference temperature from the first surveillance capsule were within the mean plus-20 Regulatory Guide 1.99, Revision 2 prediction-for the weld, but-not for'the plate. As discussed
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in ' the response to 3.b above. -the current CNS pressure temperature operating limit curves were developed based on the Regulatory Guide l.99, Revision 1 methods, adjusted for the increase in reference temperature measured from the surveillance test specimens, resulting in predicted reference temperature shif ts greater than that predicted using Regu' story Guide l'.99, Revision 2 methods.
Measured decreases in USE from the first surveillance capsule were within the. prediction for the. plate, but not for - the veld.
Since only one surveillance result is available, the effect of the measured decreases on the beltline predictions has not been considered, per paragraph C.2 of
)
Regulatory Guide 1.99, Revision 2.
I.
GONCLUSION lt is - the District's position that it is in full compliance with 10 CFR 50.60 and 10 CFR-50 Appendices G and H.
However, as the licensee
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Attachment Lo NSD920629 Page 1? of 12 of Cooper Nuclear Station, the District is committed to assuring that the plant is always operated safely within the bounds of its design.
Accordingly, the District will continue its multi-pronged approach to obtain additional radiation embrittlement data.
The District is continuing its search of vessel fabrication records to locate additional CNS vessel beltline weld information.
In addition, the District is working with the BWROC on two separate programs relating to embrittlement issues.
The District is actively involved with the BWROG Fracture Toughness Committee which has developed a topical report on USE estimation methods for plants where only transition data is available. The District is also a member of the BWROG Supplemental Surveillance Committee of which CNS is a host reactor.
This effort will provide additional CNS and BWR-specific embrittlement data through the testing of adoitional surveillance materials.
The District will proceed with these efforts to assure that CNS cont inues to operate safely.
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4 APPENDIX A 9
i BELTLINE MATERIAL CHEMISTRY AND RTNDT INFORMATION i
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'T T--N'+g
- -?it-32+=-'r e9-r-MT--tr'
- "WDFW*T--g-=1*a#
1?
TNs*
F'UTY WT"--'tt T Y-
CIIEMICAL COMPOSITION OF RPV BELTLINE HATI'RI ALS '
Composition by Wefght Pereent identificatinn-llent/ tot No.
C Hn P-S St Hi-Ma Cu
?.
. Louer Plates:
i l
i
=
51 0.20 1.31,
0.010 0.018 0.21 0.68
.O.45 0.20 C-2803-1.
'C2274-1 C-2803-2 C2307-1 0.23 1.30-0.009 0.015 0.21 0.73 0.45 0.21 C-2803-3 C2274-2 0.23 1.32 0.009 0.017 0.22 0.6H 0.46 0.20 1.ower-Intermediate Plates:
l 0-2801-7
.C2407-1 0.22 1.35-0.010
'O.016 0.24 0.65 0.46 0.13 i
G-2802-1 C2331-2 0.21 1.35 0.010 0.017 0.22 0.58 0.48 0.17 G-2802-2 C2307-2 0.23 1.25 0.010-0.014' O.20 0.73 0.47 0.21 l
1.ower Lont,9tudAnal Welds:
- s 1-2-233 A,B.C lleat 12420. Flux not avatlable 1092 Lot 3724
- ! i llent.12420, Flux not available l
1092, tot 3708
}
lover-intermediate 1.ongitudinal Wei.ls:
l-233 A.B.C lleat 27204 with
- 1.16 0.013 0.007 0.21 0.97 0.46 0.19 Tandem Weld lleat 12008 Flux 1092, Lot 3724 l
l.ower to Lower-Int.
Clrth Weld:
1-240 lle a t 21935. Flux 0.17 0.016 0.010 0.20 1092, Lot 3869 4
" 11t hama, Unit 1 Survei1 lance Test Report, Kolic Technical Institute, Se pt. 1973.
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7--____.
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RESULTS OF FAllRICATION TEST PROGRAM.FOR SELECTED RPV LOCATIONS
- g. :
t 4
Tensile 1
Total Area Test Cha r'py :
T 40 i
Ident.
llea t '
Yield
.UTS Elong Reduc.
Temp.
Energy HDT' SOT
- x* iDT I.ocarton Number Number (ksi)
(ksi)
(%)
(%)
(*F)
(ft-lb)
(
- F)
(*F) j?i Beltline:
1.ouer Shell Plates C-2803-1 C2274-1
'63.6 86.5 29.5
'69.4 10 50,40,33
-10 14 14 C-2803-2 C2307-1 69.6 91.3 27.5 68.4 10 40,70,50
-10 0
0 C-2803-3 C2274-2.
68.8 90.6 26.5 68.3 10 58,44,49
-10
-8
-8 Lower-Intermediate C-2801-7 C2407-1 71.2 91.2-25.5 70.9 10 50,6H,63_
-10
-20
-10 Shell Plates C-2802-1 C2331-2 72.1 92.9 28.0 68.1 40 58,70,59
-40 10 10 C-2802-2 C2307-2 71.4 91.9-28.3'
.67.4 10 89.73,72
-40
-20
--20 I.ongitudinal Weld 2-233 Ilt. 12420 65.3 81.9 31.0 69.3 10 64,69,56 n/a
-50
-50 Lot 3724 Cirth Weld 1-240 li t. 21935 70.5 86.8 28.0 69.8 to 62,59,60 n/a
-50
-50 Lot 3869 l
Non-BeltIine:
F
-t Ifpper Shell Plate C-2801-4 C2327-j 70.6 92.6 26.3 68.6 10 45,40,33
-20 14 14
+
f Vessel Flange C-2809 AWC-67 73.7 95.8 24.0 69.0 10 62.76,65 10
-20 10 llead Flange C-2810 AXO 70.0 91.8 26.5 71.5 10 96,89,74 20
-20 20 Top flead Torus G-2812 C2660-2 68.2 87.9 29.8 69.8 in 52,67,79
-10
-20
.-10 i
110t tom lica.1 Torus
'C-2806-1 C2137-3 65.7 85.7 27.3 69.0 10 39,26,36
-10 28 28 l
Closure Holta C-2863 37385 153.4 167.5 14.3 46.1 10 38,36,15 n/a I.ST - 7p*F t
I
,i.
i....,... - -
APPC: DIX B SURVEILLWCE F.ATERIAL CHEMISTRY 1:4 FORMATIO!1 r
i
4 1
i CilEMICAL COMPOSITION OF SURVEILIANCE MATERIALS d
Composition by Weight Percent Identification C
_ _ tin..
P S
SJ Ni Mo Cu i
Plate:
lleat C2307-2 0.23 1.25 0.010:
0.014 0.20 0.73 0.47 0.21
- Irrad. Specimen J64 1 23 0.007 0.77 0.51 0.22 Irrad. Specimen J6L 1.30 0.006 0.78 0.50 0.22 i
j.
Weld:
4 Unirradiated Weld 0.15 l'.37 0.012 0.010 0.24 0.73
.0.51 0.22.
Irrad. Specimen J 74 1.26 0.012 0.75 0.55 0.23 Irrad. Specimen J 7D 1.37 0.012 0.74 0.52 0.22 4
i 1
-l
,i' 4
- A dash (-) mark denotes an element that was not evaluated.
s
0 i
.L.
l i
1 r
APPENDIX C l.
f t
i COMPARISON OF SURVEILLANCE RESULTS VITH REG. CUIDE 1.99. REVISION 2 PREDICTIONS
+
T 5
?
e E
-+
l l
l L
o e.-
m.,4,..,_-.,
_,....,.,~,......,_..... _.,._,...-_,, _....
_ _ - -, + -..
+
I COMPARISON OF SURVEII.INJCE RESULTS VITH REC. CUIDE 1.99, REVISION 2 PREDICTIONS (Fluence - 2.3x1017 n/cm2)
REFERENCE TEMPERATURE INCREASE Plate:
L
%Cu - 0.22 Predicted ARTNDT - 32.7'F
%Ni - 0.78 Predicted ARTNDT+2a - 65.4*F Measured Shift - 74*F Veld:
%Cu - 0.-23 Predicted ARTNDT - 36.7'F
%Ni - 0.75-Predicted ARTNDT+2a - 73.4*F Measured Shift - 55'F USE DECREASE Place:
Unitradiated USE - 129 ft lb Irradiated USE - 111 ft lb Measured Decrease in USE - 17 ft lb (13%)
b Predicted Decrease in USE - 13%
Vald:
Unitradiated USE - 112 ft lb Irradiated USE
_11 ft lb J
l _.
Measured Decrease in USE - 27 ft-lb (24%)
j Predicted Decrease in USE - 16%
.