ML20024A391
| ML20024A391 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/14/1983 |
| From: | Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8306170210 | |
| Download: ML20024A391 (100) | |
Text
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6' PHILADELPHIA ELECTRIC COMPANY 2301 M ARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.19101 12isi 841-4ooo
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EUGENE J. BR ADLEY a ssocsav. ..se..a6 c ou...L DON ALD BLANKEN CUDOLPH A. CHILLEMI E C. MIR K H ALL T. H. M AHER CORNELL
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. . . . . ... ......, June 14, 1983 EDW ARD J. CULLEN. J R.
THOM AS H. MILLER J R.
IREN E A. McMEN N A asse.T A t cou...k Mr. A. Schwencer, Chief Docket Nos. 50-352 Licensing Branch No. 2 50-353 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Limerick Generating Station, Units 1 & 2 Request for Additional Information (RAI) from NRC Materials Engineering Branch, Materials Application Section
Reference:
Letter A. Schwencer to E. G. Bauer, Jr.,
dated March 8, 1983
Dear Mr. Schwencer:
Enclosed, in response to the reference letter, are draft responses to RAI's 252.1 through 252.7 and related draf t FSAR text changes.
The information contained on these draft responses and draft FSAR page changes will be incorporated into the FSAR, exactly as it appears on the attachments, in the revision scheduled for July 1983.
Sincerely,
\
Copy to: See Attached Service List '
J 8306170210 830614 PDR ADOCK 05000352 A PDR
T r
- cc
- Judge Lawrence Brenner (w/o enclosure)
Judge Richard F. Cole "
Judge Peter A. Morris "
Troy B. Conner, Jr., Esq. "
. Ann P. Hodgdon, Esq. "
Mr. Frank R. Romano "
Mr. Robert L. Anthony "
Mr. Marvin I. Lewis "
' Judith A. Dorsey, Esq. "
Jacqueline I. Ruttenberg "
Thomas Y. Au, Esq. "
Mr. Thomas Gerusky "
Director, Pennsylvania Emergency Management Agency Steven P. Hershey Charles W. Elliott, Esq.
Donald S. Bronstein, Esq. "
Mr. Joseph H. White, III David Wersan, Esq.
Robert J. Sugarman, Esq.
Martha W. Bush, Esq.
Atomic Safety and Licensing Appeal Board Atomic Safety and Licensing Board Panel Docket and Service Section
LG5 o s .
QUESTION 000. 252.1 '..
(5.3) -
For each reactor vessel beltline weld:
- a. indicate the post weld heat treatatnt received by each production weld and its associated sample test weld,
- b. indicate the filler material, flux material, and weld process,
- c. indicate whether the sample test welds are prepared using exce'ss base material from the beltline,
- d. provide CVN impact test resu)ts and drop weight test results,
- e. report the copper, nickel and phosphorus chemical composition.
RESPONSE
- a. The typical post weld heat treatment data is referenced in Section 5.3.1.7 and provided in Tables 5.3-9 and 5.3-10. *
- b. Referring to Section 5.3.1.7, the filler material and weld process are identified in Table 5.3-4. The flux material for the submerged arc weld is LINDE 124.
- c. The sample test welds are prepared in accordance with the ASME Code and do not include base material from the beltline.
- d. The test results are discussed in Section 5.3.1.7.1 and shown in Tables 5.3-3 and 5.3-4.
- e. The significant chemical compositions, including copper, nickel, and phosphorus, are listed in Tables 5.3-4 and 5.3-5.
DRAET PCY:hmm:csc/D051012* 6/3/83
, , LGS QUESTION NO. 252.2 ' Y-o.s>
For each reactor vessel beltline plate or forging:
- a. provide CYN impact test results and drop weight test results,
- b. report the copper, nickel and phosphorus chemical composition, material specification and plate identification.
RESPONSE '
- a. The test results are discussed in Section 5.3.1.7.1 and shown in Tables 5.3-3 and 5.3-4. ,
- b. See Tables 5.3-5 and 5.3-6.
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QUESTION NO. 252.3 t.
- o. s >
For each beltline plate or weld that has not been tested to the CVN impact test and drop weight test requirements of Section III, Summer 1972 l Addenda of the ASME Code and the upper shelf requirements of Paragraph IV.B of Appendix G, 10CFR Part 50 submit CVN impact test and drop weight test results from alternative test materials that demonstrates the beltline materials comply with these requirements. The alternative weld material I must be fabricated using the same flux type, filler wire type, weld l
process as the production sample, must be welded by the same manufacturer as the production sample, and heat treated to a equivalent metallurgical
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condition as the production weld. Thi alternative plate material must be fabricated by the same manufacturer as the production plate and must be fabricated to the same material s)ecification and heat treated to the same metallurgical condition as tie production plate.
. RESPONSE l l
The alternative plate and weld comparison data are referenced and discussed in Section 5.3.1.7.3. l i
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QUESTION NO. 252.4
- p. 3)
- f. l l
' For all ferritic RCPB valve materials and piping materials that were not fracture toughness tested to the requirements of the Summer 1972 Addenda and Winter 1972 Addenda of ASME Code, respectively, provide CVN impact and drop weight test data from alternative test materials that demonstrates l the valve and piping material would have met ASME Code requirements, had )
l they been tested. The alternative material must be fabricated by the !
same manufacturer as the production materials and must be fabricated to !
the same material specification and heat treated to the same metallurgical condition as the production plate. -
I
RESPONSE
, l The alternative test data for the mai'n steam isolation valves, which were I are referenced and discussed in Section
. notfracturetoufhnesstested 5.3.1.8.2.4. Al otherferrii.icRCPBvalvematerialsandpipingmaterials were either tested as required by Appendix G or exempted from such .
testing as discussed in Section 5.3.1.8.2. -
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. < LGS QUESTION NO. 252.5 O. 3)
To justify an exemation to the requirements of Paragraph IV.A.2.b of Appendix G, 10CFR ) art 50:
- a. indicate all flange and shell regions near geometric discontinuities that, during vessel operation, will not provide margins of safety in accordance with Appendix G,Section III of the ASME Code, ,
- b. for all locations in Item a, estimate the critical crack size du' ring normal operation which provides a margin ~6f safety equivalent to Appendix G,Section III of the ASME Code. Indicate the method of analysis, ,
- c. indicate which non-destructive test methods can be performed during in service inspection to examine for cracks of the size and location identified in Item a and b.
RESPONSE
- a. The effeet of the main closure flange discontinuity is considered by '
to establish the minimum temperature for adding bolt-up and pressurization The 60Fand90*FtotheRT"Dfespectively,*Ff as shown in Figure l 5.3-4. minimum bolt-up temperature of 80 or Limerick Unit 1, I which is shown in Figure 5.3-4, is based on an initial RT of
+20FfortheshellplatewhichconnectstotheclosurefEge. The minimum bolt up temperature of +70*F for Limerick Unit 2,*which is shown in Figure 5.3-5 is based on an initial RT of +10 F for the NDT closure flange forgings.
Because all toughness testing needed for strict compliance with 10CFR50, Appendix G was not required at the time of vessel procurement, the effect of the reactor vessel discontinuities is considered by adjusting the results of a BWR/6 reactor discontinuity analysis to the Limerick reactors. The BWR/6 analysis performed in accordance with 10CFR50 Appendix G includes the margin of safety implicit in the Appendix G requirements. The adjustment is made by increasing the minimum temperatures required by the difference between the Limerick and BWR/6 feedwater nozzle forgin RT 's ThediscontinuityadjustmentisbasedontheRTNDT of 0FNr.
N Limerick Unit 1.
This information is documented in the revised Section 5.3.1.5.3.
b,c. As shown in Table 5.3-la Item
- for the reactor vessel fianges.IV. A.2.b, 60 F is added to the RTForotherves ,
the results of the BWR/6 analysis are adjusted to Limerick Unit 1 RT conditions. In-service inspection will consist of those ex$InationsrequiredbytheappropriateeditionsofASMESectionXI.
l DRAET l PCY: ham /D051012* 5/19/83 l
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LGS QUESTION NO. 252.6 ',,' p D. 3)
In order for us to complete our review of the applicant's reactor vessel
' surveillance composition, program, report the copper, nickel and phosphorus chem
[ drop weight test results typeandfluxtypeforallsurveillancematerials.the plate material specification, fille
RESPONSE
The requested information is reference.d and discussed in Sections 5.3.1.6 and 5.3.1.9.
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.r. LGS QUESTION NO. 252.7 p.3)
In order for us to complete our review of the applicant's pressure l
temperature limits:
- a. estimate the end-of-life maximum neutron irradiation fluence (E > 1 MeV) at the 1/4 thickness and 3/4 thickness locations in the beltline region, *
- b. indicate the inside diameter and: wall thickness of the beltline I region.
RESPONSE :
- a. The estimated end-of-life maximum neutron irradiation fluences at (1/4)T and (3/4)T are 1.1 E18 n/cm2 and 4.4 E17 n/cm2, respectively.
- b. For conservative flux calculations 251 inches is used as the#inside diameter of the beltline region. thewallthicknessis63/16 inches. :
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/ 1GS FSAR 5.3 REACTOR VESSEL 5.3.1 REACTOR YESSEL MATERIALS 5.3.1.1 Materials Specifications The materials used in the reactor pressure vessel (RPV) and i appurtenances are shown in Table 5.2-4 together with the l applicable specifications.
i j 5.3.1.2 Special Processes Used for Manufacturing and Fabrication The RPV is primarily constructed from low-alloy, high strength l
steel plate and forgings. Plates are ordered to ASME SA 533 Grade B, Class 1, and forgings to ASME SA 508, Class 2. These materials are melted to fine grain practice and are supplied in the quenched and tempered condition. Further~ restrictions include a requirement for vacuum degassing to lower the hydrogen level and improve the cleanliness of the low-alloy steels.
Studs, nuts, and washers for the main closure flange are ordered l to ASME SA 540, Grade B 23, or Grade B 24. Welding electrodes i are low hydrogen type ordered to ASME SFA 5.5.
All plate, forgings, and bolting are 100% ultrasonically tested i
and surface examined by magnetic particle methods or liquid j penetrant methods in accordance with ASME Code,Section III standards. Fracture toughness properties are also measured and controlled in accordance with Section III requirements.
l All fabrication of the RPV is performed in accordance with General Electric (GE) approved drawings, fabrication procedures, j and test procedures. The shells and vessel heads are made from formed plates, and the flanges and nozzles from forgings.
Welding performed to join these vessel components is in
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accordance with procedures qualified in accordance with ASME Code,Section III and II requirements. Weld test samples are required for each procedure for major vessel full penetration welds. Tensile and impact tests are performed to determine the properties of the base metal, heat affected zone, and weld metal.
submerged arc and manual stick electrode welding processes are employed. Electroslag welding is not permitted. Preheat and interpass temperatures employed for velding of low-alloy steel meet or exceed the requirements of ASME Code,Section III.
Post-veld heat treatment at 11000F minimum is applied to all low-alloy steel welds.
i Radiographic examination is performed on all pressure-containing l welds in accordance with requirements of ASME Code,Section III, l Paragraph N-624 including Summer 1975 Addenda. In addition, all l welds are given a supplemental ultrasonic examination.
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The materials, fabrication procedures, and testing methods used in the construction of boiling water reactor (BWR) RPVs meet or exceed requirements of ASME Section III Class I vessels.
5.3.1.3 Special Methods for Nondestructive Examination The materials and welds on the RPV were examined in accordance with methods prescribed, and met the acceptance requirements specified by ASME BSPV Code,Section III. In addition, the pressure-retaining welds were ultrasonically examined using manual techniques. The ultrasonic examination method, including calibration, instrumentation, scanning sensitivity, and coverage is based on the requirements imposed by ASME Code,Section XI in Appendix I. Acceptance standards are equivalent to, or more restrictive than, those required by ASME Code,Section XI.
5.3.1.4 Special Controls For Ferritic agi Austenitic Stainless Steels 5.3.1.4.1 Compliance With Regulatory Guides l
5.3.1.4.1.1 Regulatory Guide 1.31, Control of Ferrite Content in Stainless Steel Weld Hetal
- Controls on stainless steel welding are discussed in Section 5.2.3.4.2.1.
l 5.3.1.4.1.2 Regulatory Guide 1.34, Control of Electroslag Weld Properties Electroslag welding is not employed for the RPY fabrication. j 5.3.1.4.1.3 Regulatory Guide 1.43, control of Stainless Steel Weld Cladding of Low-Alloy Steel Components RPY specifications require that all low-alloy steel be produced to fine grain practice. The requirements of this regulatory guide are not applicable to BWR vessels.
5.3.1.4.1.4 Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel Controls to avoid severe sensitization are discussed in Section 5.2.3.4.1.1.
5.3.1.4.1.5 Regulatory Guide 1.50, Control of Preheat Temperature for Welding Low-Alloy Steel l Preheat controls are discussed in Section 5.2.3.3.2.1.
5.3.1.4.1.6 Regulatory Guide 1.71, Welder Qualification for Areas of Limited Accessibility A r=, pw 5.3-2
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LGS FSAR Welder qualification for areas of limited accessibi'..ity is discussed in Section 5.2.3.4.2.3.
5.3.1.4.1.7 Regulatory Guide 1.99, Effects of Residual Elements on Predicted Radiation Damage to Reactor Pressure Vessel 5aterials Predictions for changes in transition temperature and upper shelf energy are made in accordance with the guidelines of Regulatory Guide 1.9 9.
5.3.1.5 Fracture Touchness, This section is supplemented by Section 5.3.1.7 and 5.3.1.8 in discussing the compliance to the intent of 10CFR50, Appendix G.
5.3.1.5.1 Assessment of 10 CFR Part 50, Appendix G l A major condition necessary for full compliance to Appendix G is satisfaction of the requirements of the Summer 1972 Addenda to Section III of the ASME Code. This is not possible with components that were purchased to earlier code requirements. For the extent of compliance see Tables 5.3-1a and 5.3-2a.
Ferritic materials complying with 10 CFR, Part 50, Appendiz G aust have both drop weight tests and Charpy Y-notch (CVN) tests with the CYN specimens oriented transverse to the maximum material working direction to establish the RT yrg . The CYN tests must be evaluated against both an absorbed energy and a lateral expansion criteria. The maximum acceptable RTurt must be
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determined in accordance with the analytical procedures of AS5E Code Section III, Appendix G. Appendix G of 10 CFR, Part 50 requires a miniana of 75 ft-lb upper shelf CYN energy for beltline material. It also requires at least 45 ft-lb CYN energy and 25 mils lateral expansion for bolting material at the lower of the preload or lowest service temperature.
By comparison, materials for the Limerick Units 1 and 2 reactor vessels are qualified by drop weight tests and/or in most cases l longitudinally oriented CYN tests (both not required) , confirming
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that the material nil-ductility transition temperature (NDTT) :
longitudinal is at least 600F below the lowest service I temperature. When the CYN test was applied, a 30 ft-lb energy level was used in defining the NDTT. There was no upper shelf CYN energy requirement on the Limerick Units 1 and 2 beltline material. The bolting material was qualified to a 30 ft-lb CVN energy requirement at 600F below the minimum preload temperature.
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' I To determine operating limits in accordance with 10 CFR, Part 50, Appendix G, estimates of the beltline material RT uro and the highest RT ,,,of all other material were made, as explained in 5.3-3 3DaRg pg,g
l LGS FSAR Section 5.3.1.5.3. The method for developing these operating limits is also described therein.
5.3.1.5.2 Method of Compliance The method of compliance is based on the last paragraph on p.
19013 of the July 17, 1973 Federal Register. The intent of the proposed special method of compliance with Appendix G for this vessel is to provide operating limitations on pressure and temperature based on fracture toughness. These operating limits ensure that a margin of safety against a nonductile failure of l this vessel is very nearly the same as a vessel built to the Summer 1972 Addenda.
l The specific temperature limits for operation when the core is critical are based on a proposed modification to 10 CFR, Part 50, Appendix G, Paragraph IV, A.2.C. The proposed modification, its justification and the results of an NRC review are given in GE Licensing Topical Report NEDo-21778-A.
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5.3-4 i
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LGS FSAR A minimum boltup and pressurization temperature of 800F is called for, which is at least 600F above the flange region RTuoT for Limerick 1. This exceeds the minimus BTgor temperature required by ASME Code Section.III, Paragraph G-222 (c) , Sunner 1976 and later editions. A flange region flaw size less than 0.24 inch critical flaw depth can be detected at the outside surface of the flange to shell and head junctions where stresses due to boltup are most limiting.
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4 4 dY 5.3-4a
l LGS FSAR 5.3.1.5.3 sethods of obtaining operating Limits Based on l Fracture Toughness operating limits that define minimum metal temperatures versus reactor pressure during normal heatup and cooldown, and during inservice hydrostatic testing, are established using the methods of Appendix G of Section III of the ASME BSPV Code, 1971 Edition including the Summer 1972 Addenda. The results are shown in Figure 5.3-4 (Linerick Unit 1) and 5.3-5 (Limerick Unit 2) .
Estimated RTggy values and temperature limits are given in this section for the limiting locations in the reactor vessel.
- All the vessel shell and head areas remote from discontinuities I were evaluated and the operating limit curves are based on the Limiting location. The boltup limits for the flange and adjacent l
l shell regions are based on a minimum metal temperature of RT woT +
l 60*F. The maximum throughwall temperature gradient from continuous heating and cooling at 1000F per hour was considered.
The safety factors applied were as specified in ASME Code Appendix G and GE BWR Licensing Topical Report NEDo-21778-A.
For the purpose of setting these operating limits the reference temperature, RTggy, is determined from the toughness test data taken in accordance with requirements of the Code to which the vessels are designed and manufactured. These toughness test data, CYN and/or dropweight NDT are analyzed to permit compliance with the intent of 10 CFR, Part 50, Appendix G. Because all toughness testing needed for strict compliance with Appendix G was not required at the time of vessel procurement some toughness results are not available. For example, longitudinal CYN's, instead of transverse, were tested, usually at a single test temperature of +100F or +400F, for absorbed energy. Also, at the time, either CYN or dropweight testing was permitted; therefore, in many cases both tests were not performed as is currently required. To substantiate the design adequacy, toughness property correlations are derived for the vessel materials in order to give a conservative estimate of RTg wr , compliant with '-
the intent of Appendix G criteria.
These toughness correlations vary, depending on the specific material analyzed, and are derived from the results of WRC Bulletin 217, " Properties of Heavy Section Nuclear Reactor
- steels," and from toughness data from the Limerick Unit 1 and 2
, vessels and other reactors. In the case of vessel plate material (SA-533 Grade B, Class 1) , the predicted limiting toughness l property is either NDT or transverse CYN 50 ft-lb temperature minus 60*F. NDT values are available for all beltline and some other Limerick 1 and 2 vessel plates. Where NDT results are missing, NDT is estimated as the longitudinal CYN 35 ft-lb transition temperature. The transverse CYN 50 ft-lb transition temperature is estimated from longitudinal CYN data in the 5.3-5 DRAFT
- ._-. . - - - - =_ _. . - . .
LGS FSAR following manner. The lowest longitudinal CVN ft-lb value is adjusted to derive a longitudinal CYN 50 ft-lb transition temperature by adding 20F per ft-lb to the test temperature. If 1 the actual data equals or exceeds 50 f t-lb, the test temperature ,
is used. Once the longitudinal 50 ft-lb temperature is derived, I an additional 300F is added to account for orientation effects j and to estimate the transverse CYN 50 ft-lb temperature minus 600F, estimated in the preceding manner.
For f orgings (SA-500 Class 2) , the predicted limiting property is the same as for vessel plates. CYN and dropweight values are l available for the vessel flange, closure head flange, and feedwater nozzle materials for Limerick Units 1 and 2. RT ggy is estimated in 'the same way as for vessel plate.
For the vessel weld metal the predicted limiting property is the CYN'50 ft-lb transition temperature minus 600F, as the NDT values are --500F or lower for these materials. This temperature is deri7ed in the same way as for the vessel plate material, except the 300F addition for orientation effects is omitted since there is no principal working direction. When NDT values are l available, they are also considered and the RTarr is taken as the i higher of NDT or the 50 ft-lb temperature minus 600F. When NDT is not available, the RTy g7 shall not be less than -500F, since lower values are not supported by the correlation data.
For vessel wold heat affected zone (HAZ) material the RTwar is assumed the same as for the base material, since ASME Code weld procedure qualification test requirements and post-veld heat 1,
treatment indicates this assumption is valid.
Closure bolting material (SA-540 Grade B24) toughness test requirements.for Linerick Units 1 and 2 are for 30 ft-lb at 600F
- .below the boltup temperature. Current Appendix G requirements are l
, for 45 ft-lb and 25 mils lateral expansion (MLE) at the preload or lowest service tenperature, including boltup. All.Limeriet unit 1 closure stud materials meet current requirements nt +100F.
All but one heat, for which records were not available, of the Limerick Unit 2 closure stud materials meet current requirements at +100F. The'purchuse requirement for Limerick Unit 2 closura l stud material waa for 30 ft-lb at +100F, and no deviation is reported. Thus, 600F is added to the specified test temperature for Limerick Unit 2 to derive the boltup temperature.
Using this general appro.ach, an initial RTuor of 200F is established for the core beltline region for Limerick Unit 1 and later for Limorick Unit 2. l
' The effect of the main closure finnge discontinuity is considered by.. adding 600F and 900F to the RTwdr to establish the minimum
, temperature for boltup and pressurization respectively. The minimum boltup temperature of 800F for Limerick Unit T, which is l
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, LGS FSAR shown in Figure 5.3-4 is based on an initial RTwar of +200F for the shell plate which connects to the closure flange. The minimum boltup temperature of +700F (preliminary) for Limerick l Unit 2, which is shown in Figure 5.3-5, is based on an initial RT ggy ,
of +100F for the closure flange forgings.
Because the toughness testing in strict compliance with 10CFR50 Appendix G was not required at the time of vessel procurement, the effect of the reactor vessel discontinuities is considered by adjusting the results of a BWR/6 reactor discontinuity analysis to the Limerick reactors. The BWR6 analysis performed in accordance with 10CFR50 Appendix G includes the margin of safety implicit in the Appendix G requirement. The adjustment is made by increasing the miniaua temperatures required by the difference between the Limerick and BUR /6 feedwater nozzle forging RTwar's.
The adjustment is based on an RTnpr of 400F for Limerick Unit 1 and an RT ggy of later for Limerick Unit 2.
l t The reactor vessel closure studs have a minimum Charpy impact l energy of 48 ft-lb and a 27 HLE at 100F for Limerick Unit 1. The l
studs for Limerick Unit 2 have a specified Charpy energy of 30 ft-lb at 100F. The lowest service temperature for boltup of Limerick Unit 2 is taken to be 600F above the later value and is later. Charpy test results are discussed in sections 5.3.1.7 and 5.3.1.8.
5.3.1.6 Material surveillance 5.3.1.6.1 Compliance with " Reactor Vessel Baterial Surveillance Program Requirements" The materials surveillance program monitors changes in the
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fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from their exposure to neutron irradiation and thermal environment.
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saterials for the program are selected to represent materials used in the reactor beltline region. The specimens are manufactured from a plate actually used in the beltline region,
- and a weld typical of those in the beltline region, and thus l represent base metal, weld metal, an? the transition zone between base metal and weld. The plate and t Li are heat treated in a manner that simulates the actual 4'at 'reatment performed on the l core region shell plates of tha 1syph ed vessel.
Further details of the vessel surveillance programs are provided in section 5.3.1.9.
The surveillance program includes three capsule holders per reactor vessel.
5.3-7 - j$
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For the specimen, arrangement, see Table 5.3-16 as referenced in Section 5.3.1.9.
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I A set of out-of-reactor baseline CYN specimens is provided with '
h - ,;G the surveillance test specimens.
charpy impact specimens for the reactor vessel surveillance programs are of the longitudinal orientation consistent with the ASME requirements prior to the issue of the 1972 Addenda and ASTH
! E185-73. Based on GE experience, the amount of shift measured by these irradiated longitudinal test specimens is essentially the same as the shift in an equivalent transverse specimen.
'[ For Limerick Units 1 and 2, each set of surveillance specimens is s
loaded in six small capsules rather than one large capsule.
,, Therefore, each capsule holder which contains all six small
>~ capsules can be considered to be the same as one surveillance capsule as defined in 10 CFR, Part 50, Appendix H. Three capsule holders are included in each reactor vessel. Since the predicted adjusted reference temperature of the boltline region is less l s
.than 100*F at end-of-life and the calculated peak neutron fluence is less than 5 x 1088 n/cm2, the'use of three capsule holders neets the requirements of 10 CFR, Part 50, Appendix H, and ASTH E185-73.
The withd'r 'awal schedule of the three sets of specimens in the reactor is. planned as follows:
a ." The first set is withdrawn when its exposure corresponds to the calculated exposure of the reactor vessel wall at 25% of the reactor design life.
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- b. The second set is withdrawn when its exposure corresponds to the calculated exposure of the reactor vessel vall at 75g of'the reactor design life.
- c. The third set is a spare to be withdrawn based on previously developed data.
For the extent of complianceito 10 CFR, part 50, Appendix H, see Tables 5.3-1b and 5.3-25.
5.3.1.6.2 Neutron: Flux and Fluence Calculations A description of the methods of analysis is contained in Sections 4.1.4.5 and 4.3.2.8.
5.3.1.6.3 Predicted Irradiation Effects on Yessel Beltline Haterials Estimated maximum changen in RTway (initial reference temperature) and upper shelf fracture energy as a function of the Og py 5.3-8
LGS FSAR end-of-life (EOL) fluence at the 1/4 depth of the vessel beltline are provided in Section 5.3.1.7. The predicted peak EOL fluence at the 1/4 depth of the vessel beltline is 1.1 x loss n/cm2 after 40 years of service. Transition temperature changes; and variations in upper shelf energy were calculated in accordance with the rules of Regulatory Guide 1.99. Reference temperatures were established in accordance with 10CFR50 Appendix G and NB-2330 of the ASME Code.
5.3.1.6.4 Positioning of Surveillance Capsules and Method of l Attachment Surveillance specimen capsules are located at three azimuths at a common elevation in the core beltline region. The sealed capsules are not attached to the vessel but are in welded capsule holders. The capsule holders are mechanically retained by capsule holder brackets welded to the vessel cladding as shown in Figure 5.3-3. The capsule holder brackets allow the capsule holder to be removed at any desired time in the life of the plant for specimen testing. These brackets are designed, fabricated, and analyzed to the requirements of Section III of the ASME code.
A positive spring-loaded locking device is provided to retain the capsules in position throughout any anticipated event during the lifetime of the vessel.
4 5.3.1.6.5 Time and Number of Dosimetry Measurements GE provides a separate neutron dosimeter so that fluence measurements may be made at the vessel ID during the first fuel cycle to verify the predicted fluence at an early date in plant This measurement is made over this short period to operation.
avoid saturation of the dosimeters now available. Once the fluence-to-thermal power output is verified, no further dosimetry is considered necessary because of the linear relationship between fluence and power output.
[ 5.3.1.7 VESSEL BELTLINE PLATES S WELDS
' This section supplements Section 5.3.1.5 in discussing the compliance to the intent of 10CFR50, Appendir G.
5.3.1.7.1 Test Data Available Charpy V-notch and Drop-Weight impact data are presented in Tables 5.3-3 and 5.3-4. There are two categories of beltline welds identified: " shop" welds and " field" welds. The shop welds represent vessel vertical seams which were made prior to shipment of pre-assembled ring segments to the Limerick 1
plant site. However, exact identification of weld materials used in the beltline girth weld sean is not available. Therefore, a
~'
DLFT
. LGS FSAR conservative assumption is made to consider all electrodes which were released for field-welding the vessel shells.
5.3-s
$4 i
, LGS FSAR Figure 5.3-7 shows the beltline layout. It gives plate heat numbers and locations, as well as weld sean locations and identifications.
5.3.1.7.2 Effects of Irradiation Copper and phosphorus values used to estimate the effects of irradiation on toughness are presented in Table 5.3-5.
Estimated starting (i.e., unirradiated) HTwor values for the beltline plate and weld materials are presented in Table 5.3-5.
These values were calculated using the data in Tables 5.3-3 and 5.3-4 in accordance with ASME Code Section III, NB2300.
l Estimated end-of-life (EOL) RTudt values (for 1/4 thickness
! location from the vessel ID) are also given in Table 5.3-5. The EOL RTwwr are estimated in accordance with Regulatory Guide 1.99, Revision 1.
l 5.3.1.7.3 Upper Shelf Toughness Testing Charpy V-notch upper shelf toughness testing was not required when the Limerick 1 vessel was manufactured. Appendix G of 10CFR50 requires a minimum of 70 f t-lb transverse upper shelf CYN energy for beltline material. Branch Technical Position HTEB 5-2 indicates that 70 ft-lbs is adequate for fluence levels less than 1 x 102* n/cma, All of the Limerick 1 beltline plates were CYN impact tested as longitudinal specimens at only one temperature, +400F. The lowest CVN value obtained for beltline plate was 45 ft-lb with l 50% shear, and the highest was 104 ft-lb with 70s shear. The 50%
~
shear value suggests there is a considerable margik remaining j before the upper shelf (i.e. , 100% shear) level is reached.
Table 5.3-6 summarizes the test certificate for a representative Limerick 1 beltline plate. Similar data are also documented for all other plates. Supporting data from representative plate materials in other BWR plants are provided in Table 5.3-7.
compatibility of the supporting data from other BWRs, Plants A through E, with respect to Limerick is based on criteria such as similarity in material, fabrication, vendor source, welding procedure, etc. All listed plate materials were produced by Luken's Steel Co. These data show that plate with as low as 36 ft-lb (Plant E, heat no. C9570-1) of absorbed energy at +400F can have longitudinal upper shelf energies in excess of 100 ft-lbs.
ETEB 5-2 states that longitudinal values should be reduced to 65%
l of the test value in order to estimate transverse upper shelf.
j To account for irradiation, a further shift in upper shelf i "
toughness can be made using Regulator Guide 1.99, Rev. 1,
- resulting in a maximum reduction of approximately 14% for the l
l t
5.3-9f DRpg
_-, = - - .
. LGS FSAR highest Cu content of 0.12 wtt as shown in Table 5.3-5. Using these conservative assumptions with a goal of achieving at least 50 ft-lb transverse toughness at EOL, the following equation is derived :
50 = .65 (L) - (.14) [ .6 5 (L) ] (where L is unirradiated longitudinal upper shelf value) l This equation predicts a minimum required unirradiated longitudinal upper shelf toughness requirement of 89 ft-lbs.
Table 5.3-7 indicates that toughness in excess of 89 ft-lbs is to be expected for longitudinal upper shelf of this material.
Although upper shelf testing was not required for the beltline velds, Table 5.3-4 shows that the majority of the weld materials, both field and shop, meet the 75 ft-lb sinimus upper shelf requirement. Of those heats where CVN toughness tests were run at only one temperature (usually +100F) and the minimum requirement was not net, there is considerable margin for improved properties at higher test temperatures (e.g. , heat #/ lots 07L8 57/B101A271, 28 ft-lbs and 20% shear & +100F) . Further upper shelf toughness data for similar velds, made by the same vendor as Limerick 1, are given in Table 5.3-8. Tables 5.3-9 and 5.3-10 present the typical weld procedures for this data base; these tables summarize surveillance program weld procedures (including that for Limerick 1) and other vessel material data representative of the Limerick 1 beltline velds. These data are 1 in excess of 75 ft-lbs at the upper shelf. Furthermore, due to i the relatively low quantities of Cu in the Limerick i beltline
, welds, no significant decrease in upper shelf toughness due to irradiation is predicted.
5.3.1.8 NONSELTLINE BfdL211]A FEBRIT.IC PIPING AHR VALVES
(
This section supplements section 5.3.1.5 in discussing the q
compliance to the intent of 10CFR50, Appendix G.
5.3.1.8.1 Nonbeltline Region Table 5.3-11 lists the estimated reference temperature (RTpgr) for various components in the vessel nonbeltline region. These
- values were derived in accordance with the intent of the ASME Code Section III, Paragraph NB-2300.
. 5.3.1.8.2 Ferritic Piping and Yalves 543.1.8.2.1 Piping )
Toughness testing of the main steam piping is in compliance with 10CFR50 Appendix G, since it was tested at +700F in accordance with the ASME Code Section III, 1971 edition with Summer 1972 Addenda.
1 i
5.3-9/
. R, A r ir
5.3.1.8.2.2 Safety Relief Valves The SRVs are exempted by the ASME Code from toughness testing because of their 6-inch size. This is consistent with 10CFR50 Appendix G.
5.3.1.8.2.3 Flued Head Fittings I
Testing of the flued head fittings is in compliance with TOCFH50 Appendix G. These materials were impact tested in accordance with the ASME Code Section III, 1971 Edition with Summer 1972 Addenda. The test temperature was 00F.
5.3.1.8.2.4 Main Steam Isolation Valves The HSIVs were procured to meet the requirements of the 1968 ASME Nuclear Draft for Pumps and Valves Code, which did not require toughness testing for the subject valve material. They were exempted because they are subjected to less than 20% of design pressure at temperatures less than 2500F.
The Limerick 1 MSIY body materials are A216 WCs carbon steel castings. Table 5.3-12 shows the significant chemical composition and heat treatment of these castings. Although impact tests were not run for Limerick 1, these materials are considered to have adequate toughness to meet the code l requirements (i.e., 25 mils lateral expansion) . Evidence of this l
design adequacy is provided in Table 5.3-13 which presents l similar HSIY body material data from other BWR projects identified as Projects & through F. These materials received heat treatments equivalent to those experienced in Limerick 1.
l I
The bonnet (i.e., cover) materials are A105 GR.2 forgings. Table 5.3-14 lists available information for a cann & Saul heat which is used to fabricate the valve covers. Reference 1 shows Charpy Y-notch in excess of 25 mils lateral expansion at +400F, and RT values no greater than -100F for SA105 material normalized at 15650F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and air cooled after forging.
1 l Additional toughness data for SA105 forging materials obtained l from fittings in another BER plant is presented in Table 5.3-15.
l These materials were normalized at 16500F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and air cooled. The toughness data given is for longitudinally oriented specimens whereas the code requirements are for transverse specimens. However, prior GE impact test experience with carbon steel material indicates it is appropriate to approximate transverse properties at about 40% of the corresponding longitudinal properties. On this basis, the data given in Table 5.3-15 demonstrates that the transverse properties meet the 25 mils lateral expansion code requirements.
i 1
LGS FSAR 5.3.1.9 RPY SURVEILLANCE PROGRAM This section supplements Section 5.3.1.6 in discussing the complaince to the intent of 10CFR50, Appendix H.
The base plate and weld materials used to fabricate the surveillance test plate are identified in Table 5.3-5. The base metal from a core beltline plate heat No. C7689-1 was used for surveillance test material. With respect to initial RTggy and percent of copper by weight, this material is considered equivalent to other beltline plates and its utilization for test plate fabrication is in compliance with current recommendations for selection of surveillance materials. The test plate weld, like the core beltline vertical weld seams, was made using both Shielded Metal Arc (SHAW) and Submerged Arc (SAW) welding processes. The test plate weld procedure is presented in Table 5.3-10. For each of the two welding processes, only one heat of weld material was used. The SAW material heat / flux No.
IP4218/3929-989, which was also used for beltline seams BE, BA, j and BB (Figure 5.3-7) , is considered suitable for surveillance l nonitoring because it represents the most limiting SAW material l in terms of shif t and predicted end-of-lifS RTwgr . The SHAW ,
material heat / lot 421A6811/F022A27A which was used for !
surveillance material was not used for production beltline welds however, the weight percentages of copper and phosphorus which it 1 contains (.09 Cu and .018 P) are generally graater than those for actual beltline material. Horeover, the unirradiated RTggr Of this material is equivalent to the initial RT,mr of the beltline weld materials. The CBSI weld procedure for test plate ;
fabrication involves utilizing stick electrode to fuse back-up bars and completing the major volume of the weld with SAW. This includes backgouging of the back-up bar to complete the back side of the weld. Therefore, the test plate weld metal is essentially submerged-arc-welded material. Table 5.3-5 indicates that all beltline materials, both plate and weld, are highly resistant to irradiated degradation of notch toughness.
Table 5.3-16 lists the actual number of specimens and their l orientations in each surveillance capsule (including tensile l specimens). The number and orientation of the Charby impact specimens are consistent with the ASME requirements prior to the issuance of the Summer 1972 Addenda and ASSE 185-73.
Prior surveillance experience indicates the amount of radiation-induced shift in properties measured by longitudinally oriented specimens is applicable to equivalent transverse-oriented specimens. Therefore, the shift when determined can be used for the transverse RTwer values for the beltline materials.
Referring to Table 5.3-16, the longitudinal orientation of the base metal HAZ specimens are such that they simulate beltline vertical seams in this manner.
5.3-9[
e d-,@y
l .
5.3.1.10 Reactor Yessel Fasteners l
The reactor vessel closure head (flange) is fastened to the reactor vessel shell flange by multiple sets of threaded studs
(
and nuts. The lower end of each stud is installed in a threaded hole in its vessel shell flange. A nut and washer are installed
.on the upper end of each stud. The proper amount of preload can be applied to the studs by a sequential tensioning using hydraulic tensioners. The design and analysis of this area of the vessel is in full compliance with all Section III Class I Code requirements. The material for studs, nuts, and washers is SA-540 Grade B23 or B24. The maximum reported ultimate tensile stress for the bolting material is 164,000 psi which is less than l the 170,000 psi limitation in Regulatory Guide 1.65. Also, the Charpy impact test recommendations of Paragraph IV.A.4 of Appendix G to 10 CFR, Part 50 were not specified in the vessel order since the order was placed prior to issuance of Appendix G to 10 CFR 50. However, impact data from the certified materials report shows that all bolting materials meet the Appendix G impact properties.
A phosphate coating is applied to threaded areas of studs and nuts and bearing areas of nuts and washers to act as a rust inhibitor and to assist in retaining lubricant on these surfaces.
5.3.2 PRESSURE-TEMPERATURE LIMITS 5.3.2.1 Limit Curves l
The basis for setting operational limits on pressure and temperature for normal, upset, and test conditions for the reactor pressure vessel is described in Section 5.3.1.5.
5.3.2.1.1 Temperature Limits for Boltup A miniana temperature of 100F is required on Limerick 1 and later on Limerick 2 for the closure studs. A sufficient number of studs can be tensioned at a temperature between 100F and 800F to seal the closure flange 0-rings for the purpose of raising reactor water level above the closure flanges in order to assist in warning them. The flanges and adjacent shell are required to
, be warned to minimum temperatures of 800F (Linerick Unit 1) and later (Limerick Unit 2) before they are stressed by the full l intended bolt preload (all bolts tensioned) . The fully preloaded boltup limits are shown on Figures 5.3-4 and 5.3-5.
l D
e 5.3-pt 10
[
--_-.-_-...._-,,._.m~-.,_,.---,____,m_ -
_ . - , - - _ _ - . _ _ _ , , - . _ , _ , . . _ . . . _ . _ . , . _ - - - . _ - .,, , . ~ - . . - - - - - _
+
. LGS FSAR 5.3.2.1.2 Teacerature Limit 12I Preoperational Tests and Inservice Inservice Inspection
-Based on the NRC general revision to 10CFR50, Appendix G Document No. [ 7590-1] Paragraph IV. A.4, the preoperational system hydrostatic test at 1563 psig prior to fuel loading may be
- performed at a miniana temperature of 1000F for Limerick Unit 1 l without fuel in the reactor, and later for Limerick Unit 2.
l These limits are established by the 400F maximum HTgor of the l
reactor vessel materials.
i The fracture toughness analysis for system pressure tests with fuel in the reactor yields the curves labeled A shown in Figures 5.3-4 and 5.3-5. The curves labeled " core beltline" are based on an initial HTudr of 200F for Limerick Unit 1 and later for Limerick Unit 2. The predicted shift in the RTwov from Figure 5.3-6, based on the neutron fluence at 1/4 of the vessel wall I thickness, must be added to the beltline curve to account for the effect of fast neutrons.
5.3.2.1.3 Operating Limits During Heatup, Cooldown, and Core operation
! The fracture toughness analysis is done for the normal heatup or cooldown rate of 1000F/ hour. The temperature gradients and thermal stress effects corresponding to this rate are included.
The results of the analyses are a set of operating limits for
, nonnuclear heatup or cooldown shown as curves labeled B in l Figures 5.3-4 and 5.3-5. Curves labeled C in these figures apply whenever the core is critical. The basis for the C curves is described in GE BWR Licensing Topical Report NEDO-21778-A.
5.3.2.1.4 Reactor Vessel Annealing Inplace annealing of the reactor vessel because of radiation embrittlement is not anticipated to be necessary because the predicted value of adjusted reference temperature does not exceed 2000F (see 10 CFR, Part 50, Appendix G, Paragraph IV.C) .
5.3.2.2 Operatino Procedures By comparison of the pressure versus temperature limits in Section 5.3.2.1 with intended normal operating procedures for the most severe upset transient, it is shown that the limits are not exceeded during any foreseeable upset condition. Reactor operating procedures are established so that actual transients are not more severe than those for which the vessel design adequacy has been demonstrated. Of the design transients, the upset condition producing the most adverse temperature and pressure condition anywhere in the vessel head and/or shell areas yields a minimum fluid temperature of 2500F and a mariana pressure peak of 1180 psig. Scran automatically occurs with
~ ~
DRAFT
- - + e - - - - +-+'--w -----~ - ~ '
- _ - - , . . --m-,.- - -_ _ _ , _ _ . , . ,-.-.---,m,- -
<m- - - - - - , , . , < , -----w-
[ LGS FSAR initiation of this event, prior to the reduction in fluid temperature, so the applicable operating limits are given by curve A Figures 5.3-4 and 5.3-5. For a temperature of 2500F, the marimua allowable pressure exceeds 1180 psig for the intended margin against nonductile failure. The maximum transient pressure of 1180 psig is therefore within the specified allowable limits. ,
l 5.3.3 REACTOR VESSEL INTEGRITY l The reactor vessels are fabricated for GE's Nuclear Energy Division by Chicago Bridge and Iron Company; and are subject to
- the requirements of GE's Quality Assurance program.
l Measures are established to ensure that purchased material, equipment, and services associated with the reactor vessels and appurtenances conform to the requirements of the purchase documents. These measures include provisions, as appropriate, for source evaluation and selection, objective evidence of ,
quality furnished, inspection at the vendor source, and i examination of the completed reactor vessels.
GE provides inspection surveillance of the reactor vessel
,- fabricator *s in-process manufacturing, fabrication, and testing operations in accordance with GE's Quality Assurance program and approved inspection procedures. The reactor vessel fabricator is responsible for the first level inspection of his manufacturing, fabrication, and testing activities and GE is responsible for the first level of audit and surveillance inspection.
t Adequate documentary evidence that the reactor vessel material, manufacture, testing, and inspection conform to the specified quality assurance requirements contained in the procurement specification is available at the fabricator's plant site.
5.3.3.1 Desion 5.3.3.1.1 Description 5.3.3.1.1.1 Reactor Yessel
,- The reactor vessel shown in Figure 5.3-1 is a vertical, cylindrical pressure vessel of welded construction. The vessels
! for Limerick are designed, fabricated, tested, inspected, and stamped in accordance with the ASME Code Section III, Class A l including the Summer Addenda 1969. Design of the reactor vessel
- and its support systen meets seismic Category I requirements.
The materials used in the reactor pressure vessel are shown in Table 5.2-4.
The cylindrical shell and bottom head sections of the reactor vessel are fabricated of low-alloy steel, the interior of which 5.3-11
. LGS FSAR is clad with stainless steel weld overlay. Nozzle and nozzle veld zones are unclad except for those mating to stainless steel piping systems.
Inplace annealing of the reactor vessel is unnecessary because shifts in transition temperature caused by irradiation during the 40-year life can be accommodated by raising the minimum pressurization temperature. Radiation embrittlement is not a problem outside of the vessel beltline region because the irradiation in those areas is less than 1 x 10:e nyt with neutron energies in excess of 1 hey.
l Quality control methods used during the fabrication and assembly of the reactor vessel and appurtenances ensure that design specifications are met.
! The vessel top head is secured to the reactor vessel by studs and nuts. These nuts are tightened with a stud tensioner. The vessel flanges are sealed with two concentric metal seal-rings designed to permit no detectable leakage through the inner or outer seal at any operating condition, including heating to operating pressure and temperature at a maximum rate of 100*F/hr
- in any one-hour period. To detect seal failure, a vent tap is i located batveen the two seal-rings. A monitor line is attached to the tap to provide an indication of leakage from the inner j seal-ring seal.
l l 5 . 3 . 3 .1.1. 2 Shroud Support l
The shroud support is a circular plate welded to the vessel vall.
This support is designed to carry the weight of the shroud, shroud head, peripheral fuel elements, neutron sources, core plate, top guide, the steam separators, the jet pump diffusers, and to laterally support the fuel assemblies. Design of the shroud support also accounts for pressure differentials across the shroud support plate, for the restraining effect of components attached to the support, and for earthquake loadings.
The shroud support design is specified to meet appropriate ASHE code stress limits.
1 5.3.3.1.1.3 Protection of Closure Studs The BWR does not use borated water for reactivity control. This section is therefore not applicable.
5.3.3.1.2 Safety Design Basis The design of the reactor vessel and appurtenances meets the following safety design bases:
- a. The reactor vessel and appurtenance will withstand adverse combinations of loading and forces 5.3-12 DRhPT
l
. LGS FSAR resulting from operation under abnormal and accident conditions
- b. To minimize the possibility of brittle fracture of the nuclear system process barrier, the following are required:
- 1. Impact properties at temperatures related to vessel operation are specified for materials used in the reactor vessel
- 2. Expected shifts in transition temperature during design life as a result of environmental conditions, such as neutron flux, are considered in the design. Operational limitations ensure that RTyer temperature shifts are accounted for in reactor operation.
- 3. Operational margins to be observed with regard to the transition temperature are specified for each mode of operation.
5 . 3 . 3 .1.3 Power Generation Design Basis l' The design of the reactor vessel and appurtenances meets the following power generation design bases:
- a. The reactor vessel has been designed for a useful life of 40 years.
- b. External and internal supports that are integral parts
, of the reactor vessel are located and designed so that
! stresses in the vessel and supports that result from i
reactions at these supports are within ASME Code limits.
- c. Design of the reactor vessel and appurtenances allows for a suitable program of inspection and surveillance.
5.3.3.1.4 Beactor Yessel Design Data The reactor vessel design pressure is 1250 psig and the design temperature is 5750F. The maximum installed test pressure is 1563 psig.
5 . 3. 3 .1. 4 .1 Yessel Support The reactor vessel support assembly consists of a ring girder and the various bolts and shims necessary to position and secure the assembly between the reactor vessel support skirt and the support pedestal. The concrete and steel support pedestal is constructed as an integral part of the structure foundation. Steel anchor bolts are set in the concrete with their threads extending above 5.3-13 DRpFT
. LGS FSAR the surface. The anchor bolts extend through the ring girder botton flange. High strength bolts are used to secure the flange of the reactor vessel support skirt to the top flange of the ring girder. The ring girder is fabricated of ASTM A-36 structural steel according to AISC specifications.
l 5.3.3.1.4.2 Control Rod Drive (CRD) Housings The CRD housings are inserted through the CRD penetrations in the reactor vessel bottom head and are velded to the reactor vessel.
Each housing transaits loads to the bottom head of the reactor.
These loads include the weights of a control rod, a CRD, a control rod guide tube, a four-lobed fuel support piece, and the four fuel assemblies that rest on the fuel support piece. The housings are fabricated of Type 304 austenitic stainless steel.
5.3.3.1.4.3 Incore Neutron Flux Monitor Housings Each incore neutron fluz monitor housing is inserted through the incore penetrations in the bottom head and is velded to the inner surface of the bottom head.
An incore flux monitor guide tube is velded to the top of each housing and either a source range monitor / intermediate range monitor (SR5/IRH) drive unit or a local power range monitor (LPRB) is bolted to the seal / ring flange at the bottom of the housing (see Sections 7.6 and 7.7) .
5.3.3.1.4.4 Reactor Vessel Insulation The reactor vessel top head insulation is designed to permit complete submersion in water during shutdown without loss of insulating material, contamination of the water, or adverse effect on the insulation efficiency after draining. All reactor vessel insulation is of the stainless steel, reflective type.
The top head insulation framework is designed to seismic Category I requirements and is used as an anchor point for reactor vessel head spray and vent piping.
The insulation above the reactor vessel stabilizer brackets is close-fitting, freestanding insulation designed to be 100g removable for inservice inspection of the reactor vessel.
i l The insulation below the stabilizer brackets is suspended fron l the brackets to allow a minimum of 8 inches annular clearance between the reactor vessel and the insulation for remote inservice inspection of the reactor vessel. The suspended insulation is also equipped with removable access ports.
Reactor vessel bottom head insulation includes horizontal flat panels connected to a cylindrical shell covering the inside of the reactor support skirt. The top row of the cylindrical shell
'~'~"
DRAFT
- LGS FSAR panels are removable to expose the bottom head for inservice inspection.
Quick removable insulation is provided around all reactor vessel nozzles to allow manual or remote automatic examination of nozzle-to-vessel and nozzle-to-piping welds.
5 . 3. 3 .1.4 . 5 Reactor Vessel Nozzles All piping connected to the reactor vessel nozzles is designed to not exceed the allowable loads on any nozzle.
The vessel top head nozzles are provided with a flange with large groove facings. The drain nozzle is of the full oenetration weld design. The recirculation inlet nozzles (located as shown in Figure 5.3-1) , feedwater inlet nozzles, the RHR low pressure coolant injection inlet nozzles, and the core spray inlet nozzles all have tharmal sleeves.
Nozzles connecting to stainless steel piping have safe ends made of stainless steel or Inconel/ASME Code Section III, SB-166 . l These safe ends are welded to the nozzles after the pressure :
vessel has been heat treated to avoid furnace sensitization of I
, the stainless steel safe ends. The material used is compatible l with the material of the nating pipe.
The nozzle for the standby liquid control injection pipe is l designed to minimize thermal shock effects on the reactor vessel, 4 if the use of the standby liquid control system is required.
5.3.3.1.4.6 Materials and Inspections The reactor vessel is designed and fabricated in accordance with the appropriate ASME Code as defined in Section 5.2.1. Table 5.2-4 defines the materials and specifications. Section 5.3.1.6 defines the compliance with reactor vessel material surveillance progran requirements.
5.3.3.1.4.7 Reactor Yessel Schematic (BWR)
The reactor vessel schematic is contained in Figure 5.3-1. Trip
(
system water levels are indicated as shown in Figure 5.3-2.
5.3.3.2 Materials of Construction All materials used in the construction of the RPY conform to the requirements of ASME Code Section II materials. The vessel heads, shells, flanges, and nozzles are fabricated from low-alloy steel plate and forgings purchased in accordance with ASME specifications SA533 Grade B Class I and SA508 Class 2. Special requirements for the low-alloy steel plate and forgings are discussed in Section 5.3.1.2. Cladding employed on the interior
~
DRMT
. LGS FSAR surfaces of the vessel consists of austenitic stainless steel weld overlay.
These materials are selected because they provide adequate strength, fracture toughness, fabricability, and compatibility with the BWR environment. Their suitability is demonstrated by long-tern successful operating experience in reactor service.
5.3.3.3 Fabrication Methods The RPV is a vertical, cylindrical pressure vessel of welded construction fabricated in accordance with ASME Section III, Class I requirements. All fabrication of the RPV is performed in accordance with GE approved drawings, fabrication procedures, and test procedures. The shell and vessel head are made from formed low-alloy steel plates, and the flanges and nozzles from low-l alloy steel forgings. Welding performed to join these vessel components is 17 accordance with procedures qualified in ASME Section III and II requirements. Weld test samples are required for each procedure for major vessel full penetration welds.
, submerged are and manual stick electrode welding processes are employed. Electroslag velding is not permitted. Preheat and
, interpass temperatures employed for welding of low-alloy steel meet or exceed the requirements of ASME Code Section III, subsection NA. Postweld heat treatment of 11000F minimum is appliad to all low-alloy steel welds.
All previous BWR pressure vessels employed similar fabrication methods. These vessels have operated for periods of up to 16 years and their service history is excellent.
The vessel f abricator, Chicago Bridge and Iron Co., has had extensive experience with GE reactor vessels dating back to 1966.
CBI Nuclear Co. was formed in 1972 from a merger agreement between Chicago Bridge and Iron Co. and GE and has continued as the primary supplier for GE domestic reactor vessels.
5.3.3.4 Inspection Requirements l
All plate, forgings, and bolting were 1003 ultrasonically tested and surface examined by magnetic particle methods or liquid penetrant methods in accordance with ASME Code Section III requirements. welds on the reactor pressure vessel were examined in accordance with methods prescribed and meet the acceptance requirements specified by ASME Code,Section III. In addition, the pressure retaining welds were ultrasonically examined in accordance with ASME Code,Section XI requirements prior to shipping.
~
5.3-16 DRh*bk["
- LGS FSAR 5.3.3.5 Shipment agi Installation The Limerick reactor vessels were assembled at the site. Methods and procedures are discussed in the PSAR, Appendix G. Suitable measures were taken during installation to ensure that vessel integrity was maintained; for example, access controls were applied to personnel entering the vessel, weather protection was l provided, and periodic cleanings were performed.
5.3.3.6 Operatina conditions Procedural controls on plant operation are implemented to hold thermal stresses within acceptable ranges. These restrictions on coolant temperature are:
- a. The average rate of change of reactor coolant temperature during normal heatup and cooldown shall not exceed 1000F during any one-hour period.
- b. If the coolant temperature difference between the done (inferred from Paar ) and the bottom head drain exceeds 1450F, the reactor recirculation pumps shall not be started, and neither reactor power nor recirculation pump flow shall be increased. <
I
- c. The pump in an idle reactor recirculation loop shall not
, he started unless the coolant temperature in that loop l is within 500F of average reactor coolant temperature.
1 The limit regarding the normal rate of heatup and cooldown (item 1 a .) ensures that the ressel closure, closure studs, vessel support skirt, and CRD housing and stub tube stresses and usage remain within acceptable limits. The limit regarding a vessel i temperature limit on recirculation pump operation and power level increase restriction (item b.) augments the iten a. limit in further detail by ensuring that the vessel botton head region is not varmed at an excessive rate caused by rapid sweep out of cold coolant in the vessel lower head region by recirculation pump
~
operation or natural circulation (cold coolant can accumulate as l a result of control drive inleakage and/or low recirculation flow rate during startup or hot standby) . The item c limit further restricts operation of the recirculation pumps to avoid high thermal stress effects in the pumps and piping, while also minimizing thermal stresses on the vessel nozzles.
The above operational limits are maintained to ensure that the stress limits within the reactor vessel and its components are within the thermal limits to which the vessel is designed for normal operating conditions. To maintain the integrity of the vessel if these operational limits are exceeded, the reactor vessel is also designed to withstand a limited number of transients caused by operator error. Also, for abnormal 5.3-17 l
DRAFT r---- -,
,,--,m----- - -. . , , . ,,-.,.----,,,,-we , -, . , - - . . - .--.,,---.,,.,.,--,----,---,.,.,-.,.y . - , , - - - ,,-,,
. LGS FSAR operating conditions where safety systems or controls provide an automatic temperature and pressure response in the reactor vessel, the reactor vessel integrity is maintained since the severest anticipated transients are included in the design conditions. Therefore, it is concluded that vessel integrity is maintained during the most severe postulated transients, since all such transients are evaluated in the design of the reactor vessel. The postulated transient for which the vessel has been designed is shown in Figure 5.2-5 and discussed in Section 5.2.2.
5.3.3.7 Inservice Surveillance Inservice inspection of the RPV is in accordance with the requirements of the 1974 Edition of the ASME BSPY Code,Section II, including the Summer 1975 Addenda. The vessel is examined once prior to startup to satisfy the preoperational requirements of the ASME Code,Section II. Subsequent inservice inspection is scheduled and performed in accordance with the requirements of 10 l CFR Part 50.55a, Subparagraph (g).
The materials surveillance program monitors changes in the fracture toughness properties of ferritic materials in the
, reactor vessel beltline region resulting from their exposure to l
neutron irradiation and the thermal environment. Specimens of actual reactor beltline material are exposed in the reactor l vessel and periodically withdrawn for impact testing. Operating l r procedures are modified in accordance with test results to ensure adequate brittle fracture control.
Material surveillance programs and inservice inspection programs are in accordance with applicable ASHE code requirements, and provide ensurance that brittle fracture control and pressure vessel integrity are maintained throughout the service life of the RPY.
Inservice inspection and testing of the reactor coolant pressure boundary is discussed in detail in Section 5.2.4.
543.4 REFERENCE
- 1. Metal Progress, July 1978, pp. 35-39.
i 5.3-18
e
- l
' LGS FSAR TABLE 5.*3-la (Page 1 of 3)
APPENDIE G MATRIE FOR LIMERICE UNIT 1 COT LY APPENDIE G YES/ )
PAR A. NO. 10PIC ALTERNATE ACTIONS OR N/A OR CO'tMENTS I,II Introductions definitions --
III.A Compliance With AsME Code, Section Yes NB- 2 300 See Section 5. 3.1. 5.1. 2 for discussion t
III.B.3 Location and orientation of impact Yes See III.A. above.
test specimens III.B.2 materials used to prepare test No specimens Compliance escept for CVN orientation and CVN upper shelf III.B.3 Calibration of temperature, No Paragraph NB-2360 of the ASME BCPV Code Section III instrumentation, and Charpy test was not in esistence at the time of purchase of the
, machines Limerick 13 nit i RPV. However, the requirements of t he
, 1975 edition of the ASME BSPV Code,Section III summer 1971 addenda, are met. For the discussions of the GE interpretations of compliance and HSC ,
acceptance see Refs 5 and 2. The temperature instru-ments and Charpy test machines calibration data are retained until the nest calibration. This is in accordance with Reg Guide 1.08 Dev 2 GE Alternative Position 1.88 (see section 1.0) and ANSI N45.2.9, 1974.
Therefore, the instrument calibration data for Limerick Unit I are not currently available.
III.B.4 Qaalification of testing personnel No No written procedures were in esistence as now required by the regulationt howeves, the individuals were .
qualified by on-the-job training and past esperience.
Por a discussion of the GE interpretation of cor-pliance and NRC acceptance see Defs I and 2.
III.B.5 Test results recording and certification Wy Yes See pets 1 and 2.
III.C.I Test conditions too see III.A. III.B.2, above.
g III.C.2 Materials used to prepare test specimens for repctor vessel beltline Yes Compliance m base metal and weld metal tests.
Test welds are not necessarily made on the same heat as that of the base plate.
~
IV.A.I Acceptance standard of materials -- --
~~ ,
W IV.A.2.a Calculated stress intensity factor Yes --
IJ25 FSAR O
TABLE 5.3-la (Cont'd) (Page 2 of 3)
I CDeeLY 1
APPEN011 3 YES/MO PARA. NO. 10PIC ALTERNATE ACTIONS OR N/A OR CoetDrrS i
IV.A.2.b Requirements for nossles, flanges and No Plus 60*r added to the RTNDT for the reactor vessel shell region near geometric flanges. For feedwater nossles, the results of the discontinuittee a
SWR /6 analysis are adjusted to Limerick unit 1 RTN or
! conditions.
IV.A.2.c RPV metal temperature requirement No Regulation change in process (see LTR teFDO-21778Al when core is critical IV.A.2.d Minimum permissable temperature during Yee hydro test IV.A.3 Materials for piping, pumpe, and valwee No see Section 5.2.3.3.1.
IV. A. 4 Materials for botting and other fasteners Yes Neet requirements for closure studs at 10*F pl IV. B Minimum upper shelf energy for RPV belt- No j
~ No upper shelf teste run. Bow ver, recommend line 4
acceptance based on lowest CVN for plate of 45 f t-lb (L) at +40*r with 505 shear and Cu of t 0.18 to 0.125 which by Regulatory Guide 1.99 indicates shelf decrease of only 145. lowest j CVNe for welds are. 35 f t-lb and 505 shear of el0*F, and 31 f t-Ib and 305 shear at +10*F with Cu of 0.03 to 0.095 which by Regulatory Guide 1.99 ladicates. shelf decrease of only 117.
j End-of-life upper shelf values (1005 shear) are predicted to be in escess of 50 f t-lb based on the preceding data. Upper shelf tests will te run g-p
- on surveillance specimens to verify adequacy.
L. IV.C Requirement for annealing when NA I% RTMDy >200*F V.4 Requirements for material --
See Table 5.3-lb surveillance program g V. B Conditions for continued operation Yes
-i Meet requirements of IV.A.2 V.C Alternattwe if V.P. cannot be NA satisfied V. D Relutrement for RPV thermal MA h annealing if V.C. cannot be met ,
I e
___ - - ' ~ ~
. e i
TABLE 5.3-ta (Cont'd) (Page 1 of !)
4 CDW LY APPEND 1E G YES/MO ALTE8 MATE ACTIONS PARA. 100. TOPIC OR N/A OR COMMENTS
- v. E aeporting requirement for v.C and v.o un neferences su Letter nrm-ele-17, o.G. sherwoos (GE) to Edson o. case (unes dated october 17, 1971.
- Letter, Robert a. Minogue (unct to s.o. sherwood (GE) dated February 14, 1918.
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J LGe FSAM TABLE 5.3-lb (Page 1 of 2)
APPENDIE N NATRIE Pon LINERICE I COMPLY APPENDIE N VES/MO ALTERNATE ACTIONS PARA. NO. 1011C 09 N/A Os CONNE3rts
! Introduction MA 1 II.A Fluence (1987 n/ cme - surveillance NA program not required I
i
!!.E Standards requirements IASTNp for No sooncompliance with ASTN E185-73 in that the surveillance surveillance specimens are not necessarily from
' the limiting beltline material. Specimens are from representative beltline material, however, and can be used to predict behavior of the limiting i material. Heat and heat / lot numbers for surveil-lance specimens are to.be supplied.
II.C.I Surveillance specimen is No Noncompliance in that sp cimens are not neces-I taken from locations alongside sarily taken from alongside specimens required the f racture test specimens by Section III of Appendia G and transverse CVNs (Section IIT.B of Appendia G) are employed. However, representative 1 materials are used, and RTNDT shift appears to 14 independent of specimen orientation.
II.C.2 Locations of surveillance capsules Yes Code basis is used for the attachment of brackets to in RPV vessel cladding. See Section 5. 3.1.6.4.
i II.C.3.a Withdrawal schedule of capsules, Yes Thrre capsules planned. Starting RTNDT of limiting RTN or 5 100*F material is based on alternative action (see para-graph III. A of Appendia Gl .
II.C.3.b Withdrawal schedule of capsules, MA 100< RTNOT 5 200*F
.jf II.C.3.c Withdrawal schedule of capsules, MA MTMDT > 200*F III.A Fracture toughness testing No CVN tests only l requirements of specimens t
III 8 Nethod of determining adjusted p2 II.a and II.C.1 above reference teaterature for base metal, HAZ and weld metal y IV.A Reporting requirements of test Yes results IV.B Requirement for dosimetry Yes measurement
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O 1 TABLE 5.3-2a (Page 1 of 33 APPENDIE G MATRIE FOR LINERICK UNIT 2 COMPLY APPENOIE G YES/NO ALTERNATE ACTIONS PARA. NO. TOPIC OR WA 08 CONNENTS I,II Introductions definitions --
III.A Compliance with AsME Code, section Vee NS-2300 see section 5.3.1.5.1.2 for discussion III.B.3 Location and orientation of impact Yes See III.A, above.
test specimens III.B.2 Materials used to prepare test specimens No Compliance escept for CVN orientation and CVN upper shelf III.B.3 Cattl ration of temperature, No Paragraph NO-460 of the AsME BSPV Code instrumentation, and Charpy test Section III was not in existence at tte machines
/ time of purchase of the Limerick ten it 2 DPV. Rowever, the requirements of the 1971 edition of the ASNE B&PV Section III Code, Summer 1971 Addenda, were met. For
. the discussions of the GE interpretations of compliance and NRC acceptance see Refs I and 2. The temperature instruments and Charpy test machines calibration data are retained until the next recalibration.
This is in accordance with Peg Guide 1.88 Rev 2, GE Alternative Position 1.08 (see section 1.33 and ANSI M45.2.9, 1974.
Therefore, the instrument calibration data for Limerick Unit 2 are not currently available.
. s 111.B.4 Qualification of testing personnel No i No written procedures were in entstence as pse . now required by the regulation; however, l the individuals were qualified liv i on-the-job training and past evoerience.
- .1 For the discussion of the GE g
interpretation of compliance and NRC
' acceptance see Refs I and 2.
T III.B.5
[ Test results recording and certifica-tion Yes See Refs I and 2.
w gs III.C.I Test conditions No see III.A, III.B.2, atH)ve.
I III.C.2 Materials used to prepare test Yes Compliance on base metal and weld metal a specimens for reactor vessel beltline tests. Test welds are not necessarily e
1 1
I I
i I
i 148 rsan 1 g TasLE 5.3-2a (Cont 'd) (Page 2 of 3)
COMPLY APPENDIE G YES/NO PARA. NO. ALTERNATE ACTIONS M OR N/& OR COMN'g S _
made on the same heat as the base plate.
IV.A.l acceptance standard of materials --
IV.A.2.a Calculated stress intenalty factor Yes IV.4.2.b Requirements for nossles, flanges, and No shell region near geometric Plue f 0*r was added to the RTNDT for the discontinuities reactor vessel flanges. For feedwater nousles the results of the BWR/6 analysis were adjusted to Limerick Unit 2 PTNDT conditions.
j IV.A.2.= prv metal temperature requirement when core is critical No Regulation change in process (see ITR NEDO- 217704) .
IV.A.2.4 Minimum permise1ble tesperature during j Yes hydro test IV.A.3 Materials for pip!ng, pumps, and valves No See Section 5.2.3.3.1.
IV.A.4 Materials for botting and other Yes see Section 5.2.3.3.1.1 for diacussion fasteners IV. B Minimum upper shelf energy for RPV No No upper shelf tests run. However, t.elt line
- recommend acce.tance based on lowest tvN l for plate of 61,35,37 ft-Ib (50,30,301 ehear), longitudinal, 0.145 Cu for heat B3416-1. Regulatory Guide 1.99 indicates i
t g3 only 145 shelf decreases therefor *.
end-of-life upper shelf is estipiated to te at least 50 f t-Ib based on preceding rVN i
i f W' Il results at #40*F. Imeest CVN's for welds are 20 f t-Ib (no 5 shear records), for O.035 Cu, at el0*r 35 ft-lb (505 sheart.
for 0.035 Cu, at e t0*rg and Il ft-lb (305 shear), f or 0.045 Cu, at
- IO*f.
4 End-of-life upper shelf values (100f
- .. shear) are predicted to be in excess of 50 f t-Ib, based upon preceding data and
!
- pegulatory Guide 1.99.
I V. C pequirement for annealing uhen NA
} RTMDT > 200*r V. A seguirements for material l See Table 5.3-2b surveillance program V. a conditions for continued operation
. Yes Meet requirements of IV.A.2
e e 4
W 1As FSAR TABLE 5.3-2a (Cont'op (Page 3 of 3)
COMPLY APPENDIE G PARA. NO. TES/MO ALTERNATE ACTIONS TQF{C OR WA 09 CO@ TENTS V.c Alternative if V.s cannot te NA satisfied i
V. 0 Requirement for RPV thermal annealing NA l $f V.C cannot be met l V. E poporting requirement for V.C and V.D NA i
References 1.
- Letter NFN-444-77, G.G. Sherwood (GE) to Edson G. Case (NBC) dated October 17, 1977.
- 2. Letter, Bobert S. Minogue (NBC) to G.G. Sherwood (GE) dated February 14, 1978.
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TABLE 5. 3-2b (Page 1 of 2) l APPENDIE N NATRIE FOR LIMERICK UNIT 2 I
COseLY APPENDIE N VES/NO PARb NO. 10PIC ALTERNATE ACTIONS OR N/A OR CCeutENTS 1 Introduction MA i
II. A Fluence <tStr N/cm8 - surveillance NA program not required II. B standarde requirements (ASTN) for No Noncomp!!ance with ASTN E105-13 in that the surveillance surveillance specimens are not necessarily f rom the !!aiting telt-line material. Specimene are f rom representative beltline material, however, and can be used to predict behavior of the limiting material. Heat and heat / lot numbers for surveillance specimens are to be supplied.
II.C.t Surveillance specimene are taken no Moncompliance in that specimens are not necessarily from locations alongside the fracture taken from alongelde specimens required by section test specimens (Section III.B of Appendia III of Appendia G and transverse CVils are not G) employed. However, representative materials are used, and RTNDT shift appears to be independent of specimen orientation.
II.C.2 locations of surveillance capsules Yes
- in RPV Code basis is used for the attachment of brackets to vessel cladding. See Section 5.3.1.6.4.
4 II.C.3.a Withdrawl schedule of capsules, Yes Three capsules planned. Starting RTNDr of limiting RTNDT 5100*F material is based on alternative action (see paragraph r
III.A of Appendia G).
i CJ E j II.C.3.b Nithdrawl schedule of capsules, NA 100C RTNor 5200*F G II. C. 3.C Withdrawl schedule of capsules, NA RTMDT >200*F
,r
' III.A Fracture toughness testing require . No CVN tests only ments of specimens LE.
III.B Nethod of determining adjusted No II.B and II.C.I above reference temperature for base metal, HAI, and weld metal IV.A Reporting requirements of test w Yes results
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ThaLE 5.3-2b gcontedt grage 2 of 2)
CEMeLY '
s APPEM01E N VES/NO PASA. NO. TOPIC ' ALTERNATE ACTIONS Da w/A on COMMaurs IV.3 pequiremente for doelmstry measure- Yee ment IV.C moporting requiremente of pressure / Vee temperature limite 9
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LGS FSAR TABLE 5.3-3 LIMERICK 1 BELTLINE PLATE TOUGHNESS DATA CHARPY
- SHELL HEAT #/ TEST TEMP.
COURSE SLAB # NOT (*F) ORIENTATION (1) (*F)
ENERGY (FT-LB.) LAT. EXPANSION (MILS.) % SHEAR l
NO. 1 1.0. 14-1 C7688-1 TOP -10 L +40 84 78 58 62 48 64 40 50 50 BOTTOM -10 L +40 78 58 85 78 58 85 40 50 50 I.D. 14-3 C7688-1 TOP -10 L +40 69 84 79 75 67 58 50 50 50 BOTTOM -10 L +40 104 90 86 66 72 78 70 70 70 I.D. 14-2 C7698-2 TOP -10 L +40 77 88 73 75 66 52 50 50 70 BOTTOM -10 L +40 100 98 87 79 72 64 50 60 60 NO. 2 I.D. 17-3 C7698-1 TOP -10 L +40 82 84 84 61 63 61 50 50 50 BOTTOM -10 L +40 85 96 80 69 63 66 50 50 50 I.D. 17-1 C7689-1 TOP -10 L +40 87 93 77 73 69 62 50 60 60 a
i BOTTOM -10 L +40 75 86 81 61 71 78 50 60 60-I.D. 17-2 C7677-1 TOP -10 L +40 71 71 61 52' 48 56 4G 40. 40 i
BOTTOM -10 L +40 71 45 65 54 58 55 40 50 50 i
i (1) Longitudinal or transverse DRAET Page 1 of 1 PCY: rm: csc/A05122*-1
) 6/3/83
LGS FSAR TA8LE 5.3-4 BELTLINE WELD TOUGHNESS DATA
. BELTLINE SHOP WELD TOUGHNESS DATAIII t
IDENTITY PROCESS HEAT NO.
CV . ABSOR8ED ENERGY LATERAL EXP.
FLUX LOT TEMP.(*F) (FT-LBS) (NILS) X SHEAR N17-Nozzle SMAW 07L669 K004A27A +10 50 50 54 44 44 46 50 60 40 i Weld 8-E SMAW 411A3531 H004A27A +10 60 60 68 51 52 54 N-17 Nozzle 60 50 60
- Welds 8-A SMAW 06L165 F017A27A +10 B-0, B-E, 8-F 60 61 62 _40 52 46 70 .60 70 N-17 Nozzle SMAW 401Z9711 A022A27A +10 98 104 70 69 73 99 60 70 80 Welds 8-A SMAW 662A746 H013A27A +10 35 38 47 8-D, 8-E, B-F,, 35 31 43 50 50 50 N-17 Nozzle Welds 8-A, SAW 3P4000 3932-989 +10 97 95 88 85 82 64 B-8, B-C, 88 88 70 N-17 Nozzle Weld 8-F SAW S3986 Run #934 +10 46 51 49 38 44 43 N-17 Nozzle 40 40 40 Weld 8-A, SAW IP4218 3929-989 8-D, 8-E +10 98 100 102 72 65 83
+10 82 65 83 94 91 90 58 66 77 98 95 95 Surveillance SAW 421A6811 F022A27A +10 80 Test Plate 85 91 64 73 72 70. 75 73 Weld (1) This table is complemented by Table 5.3-5.
Page 1 of 16 PCY:rm/A05122*-2 5/17/83
LGS FSAR TABLES 5.3-4 BELTLINE WELD TOUGHNESS DATA BELTLINE FIELO WELO TOUGHNESS DATA MECHANICAL TEST RESULTS Test No.: 983 Test Specimen PW ht 9 1100*F to 1150*F for 621/2 hrs.
Trade Name: Atom Arc 8018 NM Oiameter Size: 1/8" 1,400 LBS Lot Number: 8101127A TENSILE PROPERTIES Heat Number: 07L857 Specimen Type: .505" UTS: 89,000 psi YKP: 76,000 psi CHEMICAL TEST RESULTS Elongation in 2 inches: 30%
Carbon: .060 Red of Area: 71.7%
Manganese: 1.20 Nickel: .97 Silicon: .42 Molybdenum: .55 Copper: .03 IMPACT PRODERTIES Phosphorus: .012 Specimen Type: Charpy V-Notch Sulfur: .017 Test Temp: +10*F Energy (Ft.-lbs.): 28, 36, 39 Lateral Expansion (mils): 27, 41, 45
% Shear: 20, 40, 50 OTHER TESTS Concentricity: 4%
Moisture 9 1800*F: 0.18%
PCY: ham:csc/005171*
5/18/83 Page 2 of 16 DRAFT
LGS FSAR TABLE 5.3-4 8ELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
MECHANICAL TESTS Test No.: 38 Test Specimen PW 9 1100*F Trade Name: Atom Arc 8018NM to 1150*F for 621/2 hrs Ofameter Size: 5/32" 6,750 lbs.
Lot Number: C115A27A Heat Number: 402C4371 TENSILE PROPERTIES Specimen Type: .505" UTS: 94,000 psi YLP: 87,000 psi CHEMICAL TEST RESULTS Elongation in 2 inches: 26%
Carbon: .033 Red of Area: 71.3%
Manganese: 1.22 Nickel: .92 Silicon: .49 Molybdenum: .57 Copper: .02 IMPACT PROPERTIES Phosphorus: .009 Specimen Type Charpy V-Notch Sulfur: .014 Test Temp: +10*F Energy (ft-1bs.): 82, 81, 92 Lateral Expansion (mils): 62, 61, 66
% Shear Area: 80, 70, 70
.r OTHER TESTS Concentricity: 5%
Moisture 9 1800*F: 0.18%
r, c DR. AP::;_. ia PCY: ham:csc/D05171* Page 3 of 16 5/18/83
B
~
LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
MECHANICAL TESTS Test No.: WO #11-D Heat Treatment 1100-1150*F for 62 1/2 hrs.
Type Electrode: E8018NN Trade Name: Atom Arc 8018NM Electrode Diameter: 3/16" Lot Number: H004A27A TENSILE PROPERTIES Heat Number: 411A3531 Specimen Type: .505" UTS: 84,500 psi YLP: 71,500 psi Elongation in 2 inches: 29%
Red of Area: 72.5%
CHEMICAL TEST RESULTS Carbon: .066 Manganese: 1.13 Nickel: .96 IMPACT PROPERTIES Silicon: .51 Specimen Type: Charpy V-Notch t Molybdenum: .47 Test Temperature: -20*F Copper: .02 Energy (Ft.-Lbs.): 41, 68, 48 Phosphorus: .018 Lateral Expansion (mils): 39, 53, 41 i Sulfur: .017 % Shear: 35, 35, 25
,, - .::s.a: rm d
PCY: hem:csc/D05171* Page 4 of 16 5/18/83
BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
MECHANICAL TESTS Test No.: 27 Test Specimen PW ht 9 1100*F to 1150*F for 621/2 hrs Trade Name: Atom Arc 8018NM Ofameter Size: 7/32" 13,800 lbs Lot Number: C109A27A Heat Number: 09M057 TENSILE PROPERTIES Specimen Type: .505" CHEMICAL TEST RESULTI UTS: 94,500 psi Carbon: .063 YLP: 85,000 psi Manganese: 1.18 Elongation in 2 inches: 27%
Nickel: .89 Red of Area: 69.8%
Silicon: .47 Molybdenum: .53 IMPACT PROPERTIES Copper: .03 Specimen Type: Charpy V-Notch Phosphorus: .009 Test Temp: +10'F Sulfur: .021 Energy (Ft.-lbs): 43, 43, 44 Lateral Expans, ion (mils): 40, 41, 41 l % Shear: 50, 60, 50 OTHER TESTS Concentricity: 4%
i Moisture @ 1800*F: 0.18%
l DRAFT PCY: hem:csc/D05171* Page 5 of 16 5/18/83
. BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
!!ECHANICAL TESTS Test No.: 346 Trade Name: Stress relieved 50 hrs 9 1150*F Atos Arc 8018NM Diameter Size: 3/16" TENSILE PROPERTIES 7,950 lbs.
UTS: 87,500 psi Lot Number: J417B27AF YLP: 75,000 psi Heat Number: 412P3611 Elongation in 2 inches: 28%
Red of Area: 71.2%
CHEMICAL TEST RESULTS Carbon: .07 Manganese: 1.10 Chromium: .03 IMPACT PROPERTIES Nickel: .93 See next page for impact values Silicon: .36 Molybdenum: .47 Copper: .03
- OTHER TESTS l Phosphorus: .016 Concentricity: 3%
Sulfur: .019 Vanadium Moisture @ 1800*F: 0.2%
.02 Aluminum (.01 i
DRAFT PCY: hem:csc/005171* Page 6 of 16 5/18/83 l
t
. LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
TEST TEMPERATURE DROP WEIGHT TESTS SPECIMEN (*F) RESULTS MATERIAL: 8018 NM 1 -90 LOT: J417B27AF 2 Break
-80 No Break 3 -70 4 No Break
-70 No Break NOT TEMPERATURE = -80*F CVN IMPACT TESTS TEST TEMPERATURE ENERGY SPECIMEN LATERAL %
.(*F) (FT.LBS) EXPANSION (NILS) SHEAR 1 -100 8 6 3 2 -100 3
12 10 5
-80 15 13 10 4 -80 5
16 14 10
-80 19 15 6 -20 10 52 41 30 7 -20 65 54 50 8 -20 69 53 45 9 +40 100 80 90 10 -
+40 l
11 103 68 80
+72 133 91 12 +72 90 138 92 90 13 +130 136 89 100 14 +130 137 95 100 15 +130 146 97 100 Tcy = -20*F REFERENCE TEMPERATURE Material T T NOT CV RT NOT
-(Drop Weicht) (Charpy V-Notch) (References)
Wald Metal -80*F -20*F -80*F l
1 PCY: ham:csc/D05171*
5/18/83 ~ Page 7 of 16 D 8*I A "f' j' 1
l
LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
Test No.: 46 MECHANICAL TEST RESULTS Trade Name: Atom Arc 8018NM Test Specimen PW ht 9 1100*F Diameter Size: 3/16" to 1150*F for 62 1/2 hrs 7,900 lbs.
Lot Number: C118A27A Heat Number: 03M014 TENSILE PROPERTIES CHEMICAL TEST RESULTS Specimen Type: .505" UTS: 92,500 psi Carbon: .041 YLP: 82,500 psi Manganese: 1.23 Elongation in 2 inches: 26%
Mickel: .94 Red of Area: 69.5%
Silicon: .53 Molybdenum: .58 IMPACT PROPERTIES Copper: .01 Specimen Type: Charpy V-Notch Phosphorus: 312 Test Temp: +10*F Sulfur: .015 Energy (Ft.-lbs.): 42, 44, 47 Lateral Expansion (mils): 37, 37, 51
% Shear: 40, 40, 40 .
OTHER TESTS Concentricity: 5%
Moisutre 9 1800*F: 0.16%
DRAFT
~
PCY: ham:csc/005171* Page 8 of 16 5/18/83
.' LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
MECHANICAL TEST RESULTS Test No.: 242 Stress relieved 50 hrs @ 1150*F Trade Name: Atom Arc 8018NM Diametcr Size: 1/8" 2,100 lbs. TENSILE PROPERTIES Lot Number: 5411B27AD UTS: 87,600 psi Heat Number: L83355 YLP: 77,900 psi Elongation in 2 inches: 25%
CHEMICAL TEST RESULTS Red of Area: 71.4%
Carbon: .07 Manganese: 1.25 Chromium: .03 IMPACT PROPERTIES Nickel: 1.08 See next page for impact values Silicon: .38 Holybdenum: .53 OTHER TESTS Copper: .03 Concentricity: 5%
Phosphorus: .017 Moisture 91800'F: 0.2%
Sulfur: .018 Vanadium .02 Alucinum <0.01 l
DRhPT PCY: ham:csc/D05171* Page 9 of 16 5/18/83 l
.- LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
TEST TEMPERATURE DROP WEIGHT TESTS SPECIMEN (*F) RESULTS MATERIAL: 8018 NM 1 -90 LOT: 5411827AD Break 2 -80 No Break 3 -80 No Break NOT TEMPERATURE = -90*F ,
CVN IMPACT TESTS TEST TEMPERATURE ENERGY SPECIMEN LATERAL %
(*F) (FT.LBS) EXP. (MILS) SHEAR 1 -105 7 6 5 2 -105 8 7 5 3 -90 19 11 8 4 -90 21 11 10 5 -90 21 13 10 6 -30 27 25 25 7 -30 30 24 25 8 -30 34 29 25 9 -20 31 26 30 10 -20 -
36 29 30 11 -20 45 37 30 12 -10 51 39 40 13 -10 52 37 40 14 -10 63 52 50 15 +40 112 ,
83 80 cv = -10*F T
REFERENCE TEMPERATURE T T NOT CV RT NOT Material (Drop Weicht) (Charpy V-Notch) (References)
Weld Metal -90*F -10*F -70*F
."' /3 p, 7
PCY: ham: csc/D05171*
5/18/83 Page 10 of 16 I '
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[
LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
MECHANICAL TEST RESULTS Test No.: 374 Stress relieved 50 hrs @ 1150*F
'irade Name: Atem Arc 8018NM Diameter Size: 5/32" TENSILE PROPERTIES 2,000 lbs UTS: 90,000 psi Lot Number: J424827AE YLP: 76,500 psi Heat Number: 640892 Elongation in 2 inches: 27%
Red of Area: 71%
CHEMICAL TEST RESULTS Carbon: .08 Manganese: 1.20 Chromium: .014 IMPACT PROPERTIES Nickel: 1.00 See next page for impact values Silicon: .44 i Molybdenum: .55 Copper: .09 Phosphorus: .015 0THER TESTS Sulfur: .018 Concentricity: 3%
Vanadium: .02 Moisture @ 1800*F: 0.2%
Aluminum: .02 l
A lM %
a as?sig= y PCY: ham:csc/005171* Page 11 of 16 5/18/83
LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
TEST TEMPERATURE DROP WEIGHT TESTS SPECIMEN
( (*F) RESULTS MATERIAL: 8018 NM 1 LOT: J424827AE -70 Break 2 -60 3 No Break
-60 No Break NOT TEMPERATURE = -70*F CVN IMPACT TESTS TEST TEMPERATURE ENERGY SPECIMEN LATERAL %
(*F) (FT.L85) EXP. (MILS) SHEAR 1
-108 14 2 3 3
-108 16 3 3 3
-70 15 8 5 i 4 -70 5 20 9
-70 10 6 27 15
-10 10 38 26 30 7 -10 8 42 31 30
-10 45 l 9 0 31 30 '
55 38 35
! 10 0 62 11 44 40 0 62 12 48 40
+40 56 42 50 l 13 +40 75 l 14 55 60
+130 118 15 87 100 i
' +130 122 16 89 100
+130 130 ,
82 100 j cv = -10*F T
j l
i I
REFERENCE TEMPERATURE T T NOT CV RT NOT Material (Drop Weicht) 1 (Charpy V-Notch) (References)
-Weld Metal -70*F O'F -60*F PCY: ham: esc /005171*
5/18/83 Page 12 of 16 DRAFT
. LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
MECHANICAL TEST RESULTS Test No.: 261 Stress relieved 50 hrs @ 1150*F Trade Name: Atom Arc 8018NM Diameter Size: 7/32" 2,400 lbs Lot Number: 5419B27AG Heat Number: 401P6741 TENSILE PROPERTIES UTS: 85,000 psi YLP: 78,000 psi Elongation in 2 inches: 30%
Red of Area: 73%
j CHEMICAL TEST RESULTS
) Carbon: .06 l Manganese: 1.16 Chromium: .03 IMPACT PROPERTIES Mickel: .92 See next page for impact values Silicon: .34 Molybdenum: .47 Copper: .03 OTHER TESTS Phosphorus: .013 Concentricity: 3%
l Sulfur: .014 Moisture @ 1800*F: 0.2%
Vanadium: .02 i Aluminum: <0.01 g
PCY: ham:csc/D05171* Page 13 of 16 5/18/83 I
s--* --
LGS FSAR TABLE 5.3-4 BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
TEST TEMPERATURE DROP WEIGHT TESTS SPECIMEN ('F) RESULTS MATERIAL: 8018 NM 1 -70 !!reak LOT: 5419827AG 2 -60 No Break 3 -60 No Break NOT TEMPERATURE = -70*F CVN IMPACT TESTS TEST TEMPERATURE ENERGY LATERAL %
SPECIMEN (*F) (FT.LBS) EXP. (MILS) SHEAR 1 -90 13 8 5 2 -90 14 8 5 3 -70 11 12 10 4 -70 13 14 8 5 -70 16 16 15 6 -10 31 24 25 7 -10 44 30 30 8 -10 76 57 40 9 0 51 37 45 10 0 57 44 40 11 0 68 50 40 12 +40 83 61 50 13 +40 100 80 70 14 +130 136 93 100 15 +130 139 94 100 16 t130 146 94 100 T
cv
= 0*F REFERENCE TEMPERATURE T T RT NOT CV NDT Material (Drop Weicht) (Charpy V-Notch) (References)
Weld Metal -70*F O'F -60*F DRApr PCY: ham:csc/005171* Page 14 of 16 '
5/18/83
LGS FSAR TABLE 5.3-4 BELTLINEFIELD$ELDTOUGHNESSDATA(Cont'd)
MECHANICAL TEST RESULTS Test No.: CN-165 .
Heat Treatment: 1150*F Electrode Specification: WMS-444F, Rev. 1 for 50 hrs Electrode Type: CBI1H Trade Name: Raco 1 H TENSILE PROPERTIES Electrode Diameter: 3/32" Specimen Type: .505" Heat Number: SP6756 UTS: 90,500 psi YLP: 84,000 psi Elongation in 2 inches: 25%
j Red of Area: 64.1%
CHEMICAL TEST RESULTS Carbon: .13 Manganese: 1.89 IMPACT PROPERTIES Chromium .08 Snecimen Type: Charpy V-Notch Nickel: .96 Orientation: Perpendicular to Silicon: .07 weld direction Molybdenum: .48 See next page for impact values Copper: .08 Phosphorus: .008 Sulfur: .012 Vanadium: .006 Aluminum: .02 DRAFT PCY: ham:csc/005171* Page 15 of 16 5/18/83
~
TABLE 5.3-4
. BELTLINE FIELD WELD TOUGHNESS DATA (Cont'd)
TEST TEMPERATURE DROP WEIGHT TESTS SPECIMEN (*F) RESULTS MATERIAL: CBI INim 1 -40 HEAT: SP6756 2 No Break
-60 Break 3 -50 4 No Break
-50 No Break NOT TEMPERATURE = -60*F CVN IMPACT TESTS (@ 1/2T LOCATION)
TEST TEMPERATURE ENERGY LATERAL %
SPECIMEN (*F) (FT.LBS) EXP (MILS) SHEAR 1 -20 97 60 60 2 -20 115 75 75 3 -20 105 45 70 4 -20 107 74 65 5 -20 94 65 65 6 0 134 55 100 7 0 121 78 100 8 0 124 75 100 T
cv = 0*F REFERENCE TEMPERATURE T T RT Material (DropNicht) (CharpfV-Notch) (Refekkces)
Weld Metal -60*F 0*F -60*F DIRAFT PCY: ham:csc/D05171* Page 16 of 16 5/18/83
. LGS FSAR TABLE 5.3-5 LIMERICK 1 BELTLINE PLATE AND WELOS EOL RTun, (PEAK EOL FLUENCE = 1.1 x 1018 n/cm2gg{
R.G. 1.99 A. Plates ASME NB-2300 Extrap. Estimated I.D. Heat Wt % Cu Wt % P Wt % Ni Start RTu.n, (*F) A RTo,m, (*F) EOL RTu.n, (*F)
.W. ,W. .W .
14-1 C7688-1 .12 .011 .51 +10 32 42 14-2 C7698-2 .11 .010 .48 +10 27 37 14-3 C7688-2 .12 .011 .51 +10 32 42 17-1(1) C7689-1 .11 .007 .48 +10 23 33 17-2 C7677-1 .11 .016 .50 +20 36 56(3) 17-3 C7698-1 .11 .010 .48 +10 26 36 B. Welds
- 1. Shop Welds (i.e., Vertical Seams) R.G. 1.99 Seams ASME NB-2300 Extrap. Estimated Heat / Lot Used In Wt % Cu Wt % P Start RTun, (*F) A RTun, (*F) EOL RTon, (*F 4 WW 5 ITU E 31W I 411A3531/ BE .02 .018 -50 30 -20 H004A27A 06L165/ BA, BD .03 .021 -50 35 -15(3)
F017A27A BE, BF 662A746/ BA, BD .03 .021 -20 35 +15 H013A27A BE, BF 3P4000/(2) BC, BB .02 .015 -50 25 -25 3932-989 BA S3986/(2) BF .05 .019 -42 32 -10 Run #934 1P 4218/(1)(2) BE, BA .06 .010 -50 17 -33 3929-989 BB 421A6811/(1) Weld .09 .018 -50 33 -17 F022A27A Test Plate (1) - Surveillance Program Material (2) - Submerged Arc Welding (3) - The most limiting value Page 1 of 2 PCY: rm/A05121*-10 DRAET 6/3/83
LGS FSAR TABLE 5.3-5 (Continued)
R.G. 1.99 Heat / lot ASME N8-2300 Extrap. Estimated Wt % Cu Wt % P Start RTm,. C'F) A RTm ,. (*F) EOL RT ('F)
- 2. Field Welds (i.e.. Girth) 07L857/8101A27A .03 .012 -6 20 +14 402C4371/C115A27A .02 .009 -50 15 -35 411A3531/H004A27A .02 .018 -50 30 -20 -
09M057/C109A27A .03 .009 -32 15 -17 412P3611/J417827AF .03 .016 -80 27 -53 03M014/C118A27A .01 .012 -34 20 -14 L83355/5411827A0 .03 .017 -70 28 -42 640892/J424827AE .09 .015 -60 28 -32 401P6741/S419827AG .03 .013 -60 23 -37 SP6756/ .08 .008 -60 13 -47
07L669/K004A27A .03 .014 -50 23 -27 401Z9711/A022A27A .02 .021 -50 35 -15 411A3531/H004A27A
- 662A746/H013A27A Data Previously Provided Under " Shop" Welds 3P4000/3932-989 I
53986/Run #934 (3) The shell plate and weld are subjected to fluence level in excess of 1017 n/cm2; see .
Note (1) of Table 5.3-11.
Page 2 of 2 DPen" v~y l
PCY:rm/A05121*-11 5/19/83
LGS FSAR TABLE 5.3-6 LIMERICK 1 TYPICAL BELTLINE PLATE (5URVt1LLANLt PLAlt)
Mill Order No: 27265-1 . MECHANICAL TEST RESULTS Tansile Properties Requiremente: SA-533 GR. B. Class 1 UTS: 84,600 psi-85,100 psi Melt No: 7689 YLP: 63,900 psi 64,400 psi TEST RESULTS CHEMICAL % Elongation in 2": 26 (Wt%) 28 Carbon 0.20 Impact Properties Manganese 1.33 Nickel 0.48 Specimen T Charpy V-Notch Silicon 0.23 Test Temp:ype:+40*F Molybdenum 0.48 - Energy (Ft-lb): 87, 93, 77 Phosphorous 0.007 75 Sulfur 0.014 Lateral Exp. (mils):, 86,73,81 69, 62 61, 71, 78
% Shear: 50, 60, 60 50, 60, 60 Dropweight Test l TEST TEMP T0P/ BOTTOM
(*F) RESULTS
+30 1 No Break
+20 1 No Break
+10 1 No Break 0 2 No Break
-10 1 Break
-20 1 Break
-30 1 Break l
NDT=-10*F l
Page 1 of 2 i
. PCY: c/LO5183 l 5/18f83 I
. LGS FSAR TABLE 5.3-6 (Cont.) ~
MtLocation .
Drop Weights - Top and Bottom - Longitudinal Bend - Top Middle - Tranverse Tensions - Top and Botton - Transverse Impacts - Top and Botton - Longitudinal Tests 1/4T from rolled surface No closer than "T" from quenched and tempered edge Specification A. S. M. E SA-533 Gr. B CL-1 Pressure Vessel Quality Ultrasonic Testina per Procedure LS-U.T.-4 4
Heat Treatment Procedure LS-102 Rev. 5 Plates Austenitized at 1650' held 1/2 hr./ inch. min., and water quenched Tempered at 1260* held 1/2 hr./ inch min. , and air cooled.
Stress relieved at 1075* held I hr. min. and air cooled.
Test coupons then cut from plate.
Tests only Stress600*F.,
below relievedthen 9 1150' held 50 hrs. and furnace cooled to air cooled.
Maximum Heatino Rate Maximum Coolina Rate 100*F/hr. 100*F/hr.
Mechanical Property Requirements Tensile: 80/100,000 psi Yield: 50 min. 2% offset Elong: 18% in 2" sin.
Impacts:
30 ft.-lbs 9 + 40*F. Lateral Expansion and Drop Weights: -% Shear Fracture for information only.
! Type P-3 specimens with a NOT temperature no higher than +40*F.
Grain Size:
Final Plate Grain Size #5 or finer determined l
on a fully heat-treated test coupon.
l Page 2 of 2 4 PCY:csc/IO5182-1
LGS FSAR Table 5.3-7 SA-533 GRADE B. CLASS 1 PLATE TOUGHNESS DATA BASE INCLUDING UPPER SHELF (VENDOR- LUKEN5 STEEL CO. )
f PLANT A
-- ** Charpy Lateral -
EDT Temp Orient. 8xpansion Beat No. (*F) (*F) (L er 7) Ener8y (ft-Ibs) (mils) ! Shear C5978-1 +10 -40 L 7.0. 7.0. 11.0 5.3.7 0. O. 0
+10 25.0. 33.0, 30.0
+40
- 23. 25. 23 10, 10. 10 33.0. 48.0. 48.0 40. 35, 36 20. 20, 20
+110 118.0. 116.0. 109.0 79. 76. 74 80. 80, 80 l +160 123.0. 136.0. 136.0 82. 84. 84 90. 95, 95 C5978-2 -10 -40 L 22.0, 24.0. 24.0 17, 18. 19 0. O 0
+10 49.0, 46.0. 42.0
+40
- 38. 36. 33 25, 25. 25 62.0. 60.0, 41.0 46, 44. 34 35, 35, 25
+110 98.0. 90.0. 100 l +160 73. 67. 75 80. 70. 80 119.0. 120.0. 118.0 88. 86. 82 99. 100. 100 t
C5979-1 -10 -40 L 9.0. 11.0. 19.0 l +10
- 5. 7. 13 0, O. 0
( 61.0. 57.0. 43.0 45, 41. 32 30. 30, 20
+40 73.0. 92.0, 65.0 51. 63. 43 35. 45, 35
+110
+116 117.0. 116.0. 100.0 78. 76. 68 80. 80. 70 134.0. 136.0. 134.0 87. 86, 85 99. 99. 100 C6345-1 40 -80 L 8. 6
-40
- 4. 4 0. 0
- 29. 15. 23 21. 13, 16 5. O.1
+10 109, 88. 77 76. 58, 56 50. 35, 35
+40 103. 96. 122 68. 65. 77 45, 40, 60
+110 147, 147 84. 82 100, 100
+160 151. 165 87. 94 100, 100 C6318-1 -20 -40 L 25, 17. 14
+10
- 18. 11. 9 1. O. 0 l 80. 66. 72 57. 47. 50 35. 30, 30 440 85. 95, 112 64. 68. 75 40, 40. 50
+110 126. 145, 117 81. 89. 76 90. 100, 90
+160 140, 140, 139 86. 89. 88 100, 100. 100 C6345-2 -40 -80 L 10. 12 7. 9
. -40 0. 0
- 32. 16, 49 20. 11, 33 10. O. 20
+10 93. 94. 67 69. 66. 48 40. 40, 25
+40 109. 125. 128 70, 72. 82 50. 60, 70
+110 127, 165 81, 82 85. 100
+160 153. 161 86, 88 200. 100 Page 1 of 8 I*
LGS FSAR Ta ble 5.3-7 SA-533 GRADE B. CLASS 1 PLATE TOUGHNESS DATA BASE. INCLUDING UPPER SHELF (VENDOR- LUKENS STEEL CO.)
PLANT A
.' Charpy Lateral Best EDT Temp Drient Ener8y Espansion I "glo . ' (*F) ('T) por T) (ft-lbs) (mils) C. ear C5979-2 -10 -80 L 4.0, 11.0 5, 9 0, 0
-40 28.0, 17.0, 18.0 20, 16, 14 15, 10, 10
+10 64.0, 63.0, 49.0 44, 44, 34 30, 30, 20
+40 72.0, 76.0, 79.0 47, 49, 50 35, 35, 35
+110 107.0, 102.0 77, 74 85, 80
+160 134.0, 141.0 79, 83 100, 100 C5996-1 -10 -40 L 12.0, 53.0, 12.0 10, 20, 11 0, 15, 0 l +10 65.0, 60.0, 77.0 46, 42, 54 30, 30, 40
+40 88.0, 113.0, 78.0 56, 70, 54 40, 60, 40
+110 111.0, 126.0, 74, 85, 83 40, 90, 90 134.0 l
I
+160 146.0, 148.0, 86, 49, 66 100, 100, 143.0 100 C6318-1 -40 7 7.5 5.0 1
-10 32.5, 31.0 27.0, 28.5 5, 5
+10 41.5, 42.0 36.0, 37.0 25, 25
+40 31.5, 60, 26.0, 44.0 35, 30, 49, 63 34.0, 49.0 40, 40
+61 70, 71.0 57.2, 57.5 50, 50
+120 95 70.5 99
+200 300.0, 91.0, 75.0, 63.5, 99, 100, 90.0 69.5 100
~ - ._. . _.
.' TARI F S.3-7 SA-533 GRADE B. CLASS 1 PLATE TOUGHNESS DATA BASE. INCLUDING UPPER SHELF (VENDOR- LUKENS STEEL CO.)
PLAtlT A
.. Charpy Lateral ,
EDT Temp Orient. Expansion -
Best Wo. (*F) (*F) (L or T) Energy (ft-lbs) (mils) 8 Shear A5333-1 -10 -40 L 21, 13. 11 17, 11, 9 5. O. 0
+10 56. 67. 53 41, 47, 40 20, 30, 20 440 82. 100. 84 56, 70. 60 40. 50. 40
+110 126, 120. 133 87. 81, 84 80. 80. 80
+160 155. 155, 145 92. 50,89 100, 100. 100 8-0078-1 -10 -40 L 10. 14, 25 10. 13, 21 0. O. 5
+10 73. 49, 70 54, 39. 53 40, 30, 40 440 94, 100, 100 65, 68. 70 60. 60. 60
+110 118, 128, 140 82. 86, 89 90. 90, 100
+160 151. 136, 143 90. 84. 88 100, 100, 100 C6123-2 -10 -80 L 11, 8 9. 7 0, 0
-40 28. 38. 10 22, 28. 9 10, 10. 0
+10 77. 60. 73 56. 45, 53 40. 35, 40
- 440 113. 108. 122 73. 71. 75 65. 60. 75 l
+110 120. 149 89. 91 90. 100
+160 149. 151 91. 93 100, 100 C5987-1 -10 -40 L 19. 13, 14 14. 10, 13 0. O. 0
+10 63. 55, 35 47. 41, 26 35, 35, 30 440 30. 99. 87 59, 68. 61 50. 70. 55
+110 122, 134, 122 84. 86. 84 200. 100, 100
+160 122, 134, 127 86. 86. 84 100, 100. 100 C5987-2 -10 -40 L 15.0. 8.0. 10.0 7. 7. 7 0. O 0
+10 76.0. 79.0, 51.0 54, 59. 57 35, 35, 30 440 57.0. 76.0, 75.0 42. 57. 54 30. 35, 35 l +110 106.0. 102.0. 113.0 72. 68. 76 80. 80, 80
+160 140.0. 133.0, 138.0 87. 81. 84 100. 100, 100 i
I C6003-2 -10 -40 L 10. 7.0. 8.0 7. 4. 3. . O. O. 0
+10 37.0. 31.0, 51.0 28. 22. 37 20, 20, 30
+40 65.0. 49.0, 50.0 44, 34, 36 35. 30, 30
+110 81.0. 95.0. 82.0 60. 67. 61 60. 70. 60
+160 121.0. 107.0. 120.0 82. 78. 87 100, 99. 100 C5996-2 -10 -80 L 7.0. 10.0 5. 8 0. 0
-40 18.0. 25.0. 25.0 14, 20. 19 10. 10. 10
- 45. 50. 47 20. 30, 25 l +10 62.0, 71.0. 66.0 1
440 81.0. 100.0. 91.0 52, 71. 64 35. 50, 40
+110 124.0. 130.0 83. 88 90. 90
+160 149.0. 151.0 89. 91 100. 100
-S DRs77
. LGS FSAR TAELE 5.3-7 SA-533 GRADE B. CLASS 1 PLATE TOUGHNESS DATA BASE. INCLUDING UPPER SHELF (VENDOR- LUKENS STEEL CO. )
PLANT E
- Charpy Eateral Emergy Espansion 2 West WI Tamp orient Sp . ' $ (*F) (L er T) (ft-lbs) (sils) Shear C4882-1 -60 -40 L 16.0, 14.0 15, 11 0, 0
-40 41.0, 32.0, 54.0 28, 22, 38 20, 20, 25
+10 75.0, 68.0, 48.0 52, 49, 35 30, 30. 25
+40 83.0, 95.0, 100.0 59, 65, 70 45, 50, 60
+110 104.0, 116.0 75, 82 85, 90
+160 130.0, 131.0 86, 84 100, 100
-80 L 10, 8 0, 0 C4882-2 -40 10.0, 7.0
-40 46.0, 43.0, 30.0 36, 35, 26 10, 10, 10 l
+10 75.0, 50.0, 66.0 58, 44, 52 30, 20, 20
+40 90.0, 80.0, 88.0 71, 62, 63 40, 35, 40 120.0, 107.0 84, 82 85, 80
+110 j
137.0, 129.0 94, 90 100, 100
+160 C4882-2 -80 T 19.0 15.5 1 l 20, 10, 20
-50 41.0, 25.0, 37.5 29.0, 16.5, 26.0 41.0, 40.0 30.0, 29.5 20, 20
-30 61.0, 68.0 45.0, 49.5 30, 30
+10 77.0, 71.0 54.0, 54.5 50, 30
+39 91.0, 71.0 62.0, 53.0 75, 40
+70
+121 113.0 79.0 95 115.5, 113.5 82.5, 74.0 95, 95
+200
' '1% &
i D*
4 of 8 ~#
, LGS FSAR TABLE 5.3-7 SA-533 GRADE B, CLASS 1 PLATE TOUGHNESS DATA BASE. INCLUDING UPPER SHELF (VENDOR- LUKENS STEEL CO.)
PLANT C' l
Charpy Lateral i Beat.. BT Temp Ortent Ener87 Earpansson 1 Go. M (*F) (L or T) (ft-Ibs) (mils) Shear C9481-1 -30 +40 L 74, 74, 81 61, 58, 60 So, 50, 50
+40 103, 61, 85 48, 66, 72 40, 50, 30 C9481-1 -40 T 17.0 15.0 5
+10 23.5, 22.0 21.0, 20.5 10, 10
+25 36.0 31.0 20-25
+40 45.0, 35.d. 42.0, 34.2 30-35, 30, 42.0 38.0 30-35
+51 40.5 35.0 30
+70 51.0, 50.0 44.5, 42.5 40, 40
+93 71.0 58.5 70
+120 93.0 69.5 90-95
+200 93.5, 100.0, 74.0, 72.0 95, 65, 93.0 69.0 95 5 of 8 DRDry
LGS FSAR TABLE 5.3-7 SA-533 GRADE 8. CLASS 1 PLATE TOUGHNESS DATA BASE. INCLUDING UPPER SHELF (VENDOR- LUKENS STEEL CO.)
PLNIT D Charpy Lateral -
Esat* EDT Temp Orient Energy Espansion 2 go .* . ,{*JT), (*F) (L or T) (ft-1bs) (mils) Shear C4574-2 -30 -80 L 8.0, 16.0 6, 13 0, 0
-40 34.0, 32.0, 27.0 25, 24, 20 10, 10, 5
+10 48.0, 49.0, 60.0 36, 37, 43 15, 15, 20
+40 76.0, 63.0, 69.0 56, 47, 31 30, 20, 25
+110 98.0, 103.0 72, 76 95, 95
+160 121.0, 119.0 85, 82 100, 100 C4574-2 -20 7 22.0 17.5 1
+10 32.0, 35.0 22.5, 27.5 5, 5
+40 50.0, 52.5 35.5, 41.5 10, 10
+65 64.0, 55.0 47.0, 42.5 30, 30
+102 75.0 60.5 50
+119 108.0, 88 75.0, 66.0 100, 85
+201 112.5 83.5 100
+202 108.5 79.0 100 I
i 6 of 8 '
dDAPy
LGS FSAR TtELE 5.3-7 SA-533 GRADE 8, CLASS 1 PLATE Tot'GHNESS DATA 8ASE. INCLUDING UPPER SHELF (VINDOR - LUKENS STEEL CO.)
PLANT E CHARPY LATERAL HEAT NOT TEMP ORIENT ENERGY EXPANSION NO. 1*F), (*n (L OR T) (FT-LBS) __(MILS)
SHEAR C9533-2 0/10 -50 L 7-12 7-4 1-1
-30 9-7 7-4 1-1
-20 5-14*/19-40* 11-7*/5-30* 10-10/20-20
+10 34-26/45-47 28-23/36-35 20-20/30-30
/ +70 48-54/70-76 48-40/57-60 40-40/40-40 3
+100 82-62/82-71 53-65/63-61 60-60/60-60
+130 72-68/92-86 60-57/72-69 60-60/80-80
+212 88-81/110-99 75-73/87-81 99-99/99-99 C9570-2 -40 -70 L 6-4 5-4 (top) 1-1
-50 10-13 20-20 10-10
-20 14-27 23-14 20-20
+10 30-42 29-39 30-30
+70 70-64 56-54 40-40
+100 76-79 64-63 60-60
+212 106-108 66-81 90-90 CTop/ Bottom D. iglJa. - ny*
PCY: ham:csc/005172* Page 7 of 8 5/18/83
LGS FSAR TABLE 5.3-7 SA-533 GRADE 8. CLASS 1 PLATE TOUGHNESS DATA BASE, INCLUDING UPPER SHELF (VEN00R - LUKENS STEEL CO.)
PLANT E CHARPY HEAT NOT LATERAL TEMP ORIENT ENERGY EXPANSION NO.
G Q (L OR T) (FT-LBS) (MILS)
SHEAR C9570-2 -50 -70 L 5-7 3-5 1-1 (Botton)
-40 12-12 r 16-18 10-10
-20 36-45 30-36
, 20-20
+10 35-46 29-34 30-30
- +40 48-56 47-49 50-50
+70 79-81 74-75 -
70-70
+100 90-110 ~70-82 80-80
! +130 114-112 86-85 99-99 C9570-1 -20/-10 -20 L 11-8*/13-12* 6-10*/12-12*' 1-1/10-10
-10 20-21/15-14 20-20/15-14 10-10/10-10
+10 19-19/43-43
, 19,-18/34-33 10-10/30-30
+40 36-44 37-41 30-30
+72 60-62/69-70 52-50/56-56 j '40-40/40-40
+100 72'-75/79-80 65-62/64-55 -- '50-50/60-60 l +130 84-83/94-88 69-71/74-71 70-70/80-80
+212 96-110/105-103 82-78/81-80 l
90-90/90-90
- Top / Bottom DRAY, PCY: hem:csc/D05172* Page 8 of 8 5/18/83 l
_ , , , _ . , . . -e-e>e = 7' -
LG5 FSAR TABLE 5.3-8 UPPER SHELF TOUGHNESS FOR BELTLINE WELDS PLANT A .
INMiELECTRODE(TRADENAME-TECHALLOY)
LINDE 124 FLUX SUBMERGED AnC POST WELD 1150'F FOR 50 HRS TYPICAL
' Charpy HeatNo/ NDT Wire Ener ansion Flux Lot ('F) ('
Teb (Sort){1) (ft-lk) Exp(mils) %
Shear KN203/0171 -80 -130 5 7 6 7 7 5 5
-80 34 18 22 32 16 21
-20 40 35 40 68 70 62 61 57 56 80 70 75 l+10 75 72 64 64
+40 90 90 94 82 81 71 100 95
+212 94 92 86 76 80 80 100 100 100
-80 -130 T 7 5 6 5 5 5
-100' 25 16 24 19 10 10
-80 24 22 25 21 19 25
-20 25 20 30 48 49 54 44 42 46 45 45 60
-10 59 54 54 48 49 46
+10 60 45 60 78 67 65 56 95 80
+43 80 79 68 68 95 95
+212 36 89 87 87 86 85 100 100 100 u ) M ngle or landem bb l -
b.
PCY: rm/A051816* Page 1 of 7
'c5/18/83 l
,c -
~
TABLE 5.3-8 UPPERSHELFTOUGHNESSFORBELTLINEWELDS(Continued)
PLANT A '
E8018-G WELD ELECTRODE, SHIELDED METAL ARC (TRADE NAME - ATOM ARC 8018 NM)
POST WELD 1150'F FOR 50 HRS TYPICAL Charpy lateral Heat No/ NOT T Energy ansion %
Flux Lot ('F) ("F (ft-lbs) Exp(mils) Shear 640967/ -80 -105 13 14 4 4 5 5 D502B27AF -80 16 22 28 11 13 18 10 12 15
-20 58 76 86 18 42 56 15 50 50
+40 102 106 119 69 74 86 90 80 90
+72 119 127 86 82 90 90
+130 130 140 150 90 92 80 100 100 100 DRpy7 PCY:rn/A051816* Page 2 of 7 5/18/83
LGS FSAR TA8LE 5.3-8 UPPER SHELF TOUGHNESS FOR 8ELTUNE WELDS (Continued)
PLANT 8 E8018-G WELD ELECTRODE (TRADENAME-E8018NM}SHIELDEDMETALARC POST WELD 1150*F FOR 50 HRS TYPICAL Charpy Lateral Heat No/ NOT Temp Energy Flux Lot ansion %
(*F) (*F) (ft-lbs) Exp(mils) Shear 401P2871/ -50 -90 7 10 H430827AF .
7 11 3 5
-70 15 16 16 14 15 16
-20 8 8 10 66 76 64 58 61 58 15 15 15
-20 63 81 50 61 15 15
-10 27 39 54 25 35 46 35 35 35 0- 27 50 56 25 42 46 40 45 45
+10 75 76 107 60 62 74 60 50 80
+40 90 100 71 76
+130 70 80 130 140 142 91 94 93 100 100 100 402P3162/ -70 -70 11 7 14 9 6 8 5 5 5 H426827AE -40 33 52 32 27 42 22 10 15 10
-20 65 62 37 52 48 30 20 10 20
-20 52 55 36 38 15 15
-10 60 54 53 44 37 --
40 30 30
+40 96 99 57 68 60 60
+212 119 122 124 93 90 68 100 100 100 DRAyy PCY:rm/A051816* Page 3 of 7 5/18/83
LGS FSAR TABLE 5.3-8 UPPERSHELFTOUGHNESSFORBELTLINEWELDS(Continued)
PLANT B INIH ELECTRODE (TRADE NAME - RACO)
LINDE 124 FLUX 3SU8MERGEL ARC POST WELD 1150 F FOR 50 HRS TYPICAL Charpy -
Heat No/ NDT T Wire Flux Lot Energy ansion %
(*F) ( (Sort)(2) (ft-lbs) Exp(mils) Shear SP7397/(1) -70 -70 0342 -10 T 22 16 36 22 18 28 5 5 5
+10 58 68 61 54 50 47 25 20 20
+10 76 73 75 60 65 60 30 45 50 75 69 58 56 35 35
+40 91 84
+70 75 63 80 85 79 75 77 73 63 74 90 95 95
+212 84 81 87 69 67 75 100 100 100
-70 -70 20 34 27
-10 S
16 32 22 5 5 5 54 50 59 47 47 53 25 20 20
+10 65 59 69
+10 60 56 65 50 25 75 70 75 56 61 45 55
+40 71 78
+70 65 68 75 90 92 101 94 82 65 69 95 95 100
+212 100 95 96 88 58 82 100 100 100 u) inis material is in Plant B's vessel surveillance program.
(2)SingleorTandem l -
l DRAFT PCY:rm/A051816" Page 4 of 7 5/18/83
LGS FSAR TABLE 5.3-8 UPPER SHELF TOUGHNESS FOR BELTLINE WELDS (Continued)
PLANT C E8018NMWELDELECTR00E(TRADENAME-ATOMARCE8018NM)
SHIELDED METAL ARC POST WELD 1150*F FOR 50 HRS TYPICAL Charpy lateral i Heat Nc/ NOT Temp Energy Expansion %
Flux Lot (*F) (*F) (ft-lbs) (mils) Shear
! 492L4871/ -60 -108 10 11 5 4 l 4 4 A421827AE -90 25 30 32 6 6 6 8 10 10
-30 19 28 31 19 23 25 20 25 25 l
-20 22 26 30 23 21 27 25 25 30
-10 38 41 43 28 32 30 30 30 30 0 50 51 57 36 38 40 30 40 45
+40 135 137 84 80 90 80
+130 151 160 161 80 82 81 100 100 100 422K8511/ -80 -90 14 17 15 16 5 5 G313A27AD -80 14 16 20 15 16 20 10 10 10 26 26 40
-40 26 24 33 30 30 30
-20 65 74 127 44 48 76 40 50 60
+25 107 108 74 80 80 70
+40 125 125 140 84 89 82 100 100 90
+50 153 143 156 95 81 91 90 80 90
+68 153 143 165 85 96 91 100 100 100 640892/ -60 -108 14 16 3 3' 3 3 J424827AE -70 15 20 27 8 9 15 5 10 10
-10 38 42 45 26 31 31 30 30 30 0 55 62 62 38 44 48 35 40 40
+40 56 75 42 55 50 60
+130 118 122 130 87 89 82 100 100 100 40150371/ -60 -60 42 45 23 35 36 20 5 5 5 B504B27AE -20 61 84 77 48 66 62 30 25 25
-20 68 67 51 52 25 25 0 80 85 82 63 62 60 35 50 35
+40 95 97 71 76 40 75
+70 87 85 77 111 107 109 80 90 80'
+212 122 114 130 92 92 69 100 100 100 402P3162/ -70 -70 11 7 9 6 5 5 H426827AE -40 33 52 32 27 42 22 10 15 10
-20 65 62 37 52 48 30 20 10 20
-20 52 55 36 38 15 15
-10 60 54 68 44 37 53 40 30 30
+40 96 99 57 68 60 60
+212 119 122 124 93 90 68 100 100 100 PCY:rs/A051816* Page 5 of 7 5/18/83
, LGS FSAR TABLE 5.3-8 UPPER SHELF TOUGHNESS FOR BELTLINE WELDS (Continued)
PLANT C E8018NM ELECTRODE (TRADE NAME - ATOM ARC E8018NM)
SHIELDED METAL ARC POST WELD 1150*F FOR 50 HRS TYPICAL Charpy lateral Heat No/ NOT T Energy ansion Flux Lot %
(*F) (* (ft-lbs) Exp(mils) Shear 401P2871/ -50 -90 7 10 7 11 3 5 H430827AE -70 15 16 16 14 15 16 8 8 10
-10 27 39 54 25 35 46 35 35 35 0 27 50 56 25 42 46 40 45 45
+10 75 76 107 60 62 74 60 50 80
+40 90 100 71 76 70 80
+130 130 140 142 91 94 93 100 100 100 07R458/ -60 -70 9 9 7 7 5 5 S403827AG -60 10 11 13 9 9 11 15 10 10 0 59 61 70 51 52 58 50 50 60
+40 99 101 77 78 80 75
+72 106 110 85 87 80 80
+130 129 131 132 81 78 81 100 100 100 03LO48/ -60 -105 8 9 2 3 3 3 8525827AF -80 10 16 19 7 10 11 10 10 10
-20 31 50 65 22 37 50 30 30 30
-10 36 53 58 34 43 45 40 40 40 0 61 75 79 44 58 59 50 60 60
+40 104 108 . 75 77 80 80
, +130 122 123 126 89 83 91 100 100 100 02R486/ -70 -100 12 13 3 5 3 5 J404827AG -90 16 17 19 6 8 7 8 8 10
-30 17 30 31 15 24 23 15 20 20
-20 41 42. 44 33 34 35 30 30 30
-10 52 64 66 39 45 46 40 40 40
+40 84 87 63 68 60 60
+130 121 124 129 91 96 95 100 100 100 L83978/ -80 -100 10 12 6 7 4 5 J414827AD -80 14 15 24 10 12 18 10 10 12
-20 51 52 81 37 40 63 35 50 40
-20 64 63 69 51 47 55 15 15 15
-20 67 56 53 45 15 10
+40 120 123 72 73 80 80
+72 128 140 78 81 90 90
+130 148 156 168. 90 81 87 100 100 100 l
A051816* . Page 6 of 7 t
, LGS FSAR TABLE 5.3-8 UPPERSHELFTOUGHNESSFORBELTLINEWELDS(Continued)
PLANT C INm ELECTRODE (TRADE NAME - RACO)
LINDE 124 FLUX SU8 MERGED ARC 3
POST WELD 1150 F FOR 50 HRS TYPICAL Charpy lateral Heat No/ NOT Temp Energy ansion Flux Lot %
(*F) (*F) (ft-lbs) Exp(mils) Shear SP7397/ -50 -70 25 21 18 15 5 5 0156 -50 42 27 19 33 25 20 10 15 10
+10 64 67 55 53 53 52 30 35 40
+10 64 70 53 54 40 45
+40 91 84 85 78 68 79 85 90 95
+212 103 92 94 59 66 59 100 100 100 I
3P4966/ -80 -80 51 27 9 45 25 12 5 5 5 0342 -20 71 66 54 57 57 45 30 25 20 l . +10 85 84 71 68 72 61 70 80 65
+10 83 76 67 64 65 55
+40 87 91 71 60 75 80 l +70 100 101 97 82 89 71 90 95 90 66 84 86
+212 108 111 108 100 100 100 4P7465/ -60 -80 27 14 21 12 5 0 0751 -70 48 43 26 42 36 22 t
! 15 15 5 0 63 57 68 54 45 63 30 25 35
+10 56 58 90 62 62 86 30 25 45
+10 87 55 83 42 40 30
+40 67 97 71 90 45 50
+212 118 102 112 88 71 72 100 100 100 1P6484/ -20 -80 5 8 6 11 5 5 0156 -60 22 16 12 23 13 10 10 10 10 0 17 36 30 20 27 28 25 20 25
+10 30 38 17 25 38 12 15 15 15 i 34 38 28 30 15 20
+30 34 46 42 29 37 45 25 50 35
+40 72 60 72 54 47 49 50 45 50
+212 93 81 83 65 66 69 100 100 100 5P5657/ -60 -80 39 39 27 37 5 5 0931 -60 19 20 32 18 22 28 10 10 10 0 51 55 58 50 50 63 30 30 55
+10 69 69 66 61 65 59 50 50 40
+10 62 57 60 63 60 40
+40 77 66 73 72 70 80
+212 88 91 85 86 75 83 100 100 100 A051816* Page 7 of 7
LGS FSAR Table 5.3-9 WELD PROCEDURE SPECIFICATION FOR VESSEL MATERIAL REPRESENTATIVE OF LIMERICK 1 BELTLINE WELOS (Note: This specification is extracted from the surveillance program of another BWR plant with similar beltline weld material.)
Reference S9ecifications General WPS 800 Latest Revision General WPS 820 Latest Revision Procedure Qualification No. Position Thickness Range 1890 (SMA) V 3/16" to 8" 1891(SMA) H 3/16" to 8" 1892 (SMA) OH, F 3/16" to 8" 1893 (SA-1) F 3/16" to 8" 2200 (SA-2) F 3/16" to 8" Post Heat Treatment Precedure qualified with 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> at 1150*F + 25'/-50*F.
Postweld heat treatment of the weldment shall be in accordance with a C8&I approved procedure.
Base Metal ASME SA-533 Gr 8 Class 1 or SA-508 Class 2 ASME Group No. P128 Subgroup 1 Shielding Gas: None Backup Gas: None Flux: Linde 124 i
DRAFT PCY:im/90-1 Page 1 of 2 5/19/83
. TABLE 5.3-9 (Cont'd)
Preheat Requirements:
Minimum preheat of 300*F shall be applied uniformly to the full thickness of the weld joint and adjacent base material for a minimum distance of "T" or 6", whichever is least, where "T" is the material thickness.
Maintain 300*F min. preheat temperature until start of postweld heat treatment except for longitudinal and circumferential shell and head seams, preheat may be dropped to 250*F min. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after completion of welding. All turnoff tabs and flux dams must be removed prior to dropping preheat below 300*F.
Interpass Temperature Requirements -
The interpass temperature shall not exceed 500 F maximum.
Filler Metal Submerged Arc Specification - N.A.
Classification - N.A.
Analysis -A3 (except Ni 0.50 to 1.25)
Usability - F6 Trade Name - CBI INMM (1% Nickel) or equal ,
Shielded Metal Arc Specification - SFA-5.5 Classification - E8018-G Analysis - A3 (except Ni 0.50 to 1.25)
Usability - F4 Trade Name - Alloy Rods E8018NM l
Electrical Characteristics SMA - DCRP ~
Submerger Arc ',
Tandem Wire 8 Lead Wire - DCRP Trail Wire - AC Single Wire - DCRP PCY:im/9D-2 Page 2 of 2 5/19/83
LGS FSAR TABLE 5.3-10 TYPICAL SURVEILLANCE PROGRAM WELD PROCEDURE FOR LIMERICK 1 & OTHER BW Reference Specifications General WPS 800 Latest Revision General WPS 820 Latest Revision Procedure Qualification NA Position Thickness Rance 963 (TW) F (Sub Arc) 4 1/2" to 9.9" F,V,H (SMA) 1261 (SW) F (Sub Arc) 2 3/4" to 8" F, V (SMA)
Post Heat Treatment Procedure qualified with 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> at 1150*F + 25'/-50*F.
Post approved C8&I weld heatprocedure.
treatment of the weldment shall be in accordance with Base Metal ASME SA-533 Gr B Class 1 or SA-508 Class 2 ASME Group No. P128 Subgroup 1 Shieldina Gas: None Backup Gas: None Flux: Linde 124 4
DRAPir Page 1 of 4 PCY:csc/IO5182-2
LGS FSAR TABLE 5.3-10 (Cont'd)
Preheat Requirements Minimum preheat of 300*F shall be applied uniformly to the full thickness of the weld joint and adjacent base material for a minimum distance of "T" or 6", whichever is least, where "T" is the material thickness.
Maintain preheat temperature until start of post weld heat treatment.
Interpass Temperature Requirements The interpass temperature shall not exceed 500*F maximum.
t Filler Metal Submerced Arc Specification - N.A.
Classification - N.A.
Analysis - A3 (except Ni 0.50 to 1.25)
Usability - F6 Trade Name - Adcom INP91 (1% Nickel) or equal Shielded Metal Arc Specification - SA-316 Classification - E8018-G Analysis - A3 (except Ni 0.50 to 1.25)
Usability - F4 l Trade Name - Alloy Rods E8018NM Electrical Characteristics SMA - DCRP l
Submerged Arc l
Tandem Wire Lead Wire - OCRP Trail Wire - AC Single Wire - DCRP e
l Page 2 of 4 r
l PCY:csc/IO5182-3 l
t.ua rsan TABLE 5.3-10 (Cont'd)
GENERAL WELDING TECHNIQUE Operation Beads Weld Electrode Description Layer Current Voltage Proc. Size Type ( Amo s_)__ (volts) Travel SMA As SMA 1/8" E8018NM 90-135 23-25 Tieq'd 5/32" 110-160 24-26
-3/16" 150-220 24-26 7/32" 250-350 25-27 1/4" 300-400 25-27 Submerged Arc Single Wire As
- Adcom
! 550-650 28-32 10-18 (DCRP) Req'd 1NMM i or Equal i
Tandem Wire As Lead
- 650-750 32-36 24 min.
Req'd Trail
- 550-650 34-37 l
- 5/32" or 3/16" l
GROOVES Submerged Arc - Flat j
SMA - Flat, Vertical & Horizontal l
The spacer bar shall be of i
- the same material as.the base material.
I )
~
)
} WELD BUILDUP
(
\
I , ',~,~ m m h b h )
/////////
Page 3 of 4 PCY:csc/105182-4
LGS FSAR TA8LE 5.3-10 (Cont'd)
GENERAL WELDING TECHNIQUE Operation Beads Weld Description Electrode Current Voltage Layer Proc. Size g (Amps) (volts) Travel SMA As SMA 1/8" E8018NM 90-135 23-25 Req'd 5/32" 110-160 24-26 3/16" 150-220 24-26 7/32" 250-350 25-27 1/4" 300-400 25-27 Submerged Arc Single Wire As 5/32" Adcom 450-700 28-35 S-18 (DCRP) Req'd or INM 3/16" or Equal GROOVES Submerged Arc - Flat SMA - Flat & Vertical
[ {
The spacer bar shall be of
\
- 1 the same material as the base material.
> \ ?
w W
) WELD BUILDUP
\
~
l i /
r , ,
f T T T
/////////
{f h J Page 4 of 4 PCY:csc/I05182-5
LGS FSAR TABLE 5.3-11 ESTIMATED RT NOT FOR COMPONENT > IN VESSEL NONBELTLINE REGION Component Material R L m ,. (*F)
- 1. Vessel Flange SA508 C1.2 -30
- 2. Top Head Flange SA508 C1.2 0
- 3. Top Head Torus SA533 Gr.8, Cl.1 +10
- 6. LPCI Nozzle (1) --
-6
- 7. Vessef2pinClosure --
Stud i
(1) The Limerick 1 RPV design results in this component experiencing a predicted end-of-life (EOL) fluence of 1.6 x 1017 n/cm2 at 1/4 of the thickness. This fluence, based on a conservatively assumed Cu content of 0.18% and a measured phosphorous content of 0.011%,
yields an estimated EOL RT NOT of +14*F. The EOL estimate is in accordance with Regulatory Guide 1.99, Revision 1.
(2) This component meets the CVN test requirements of 45 ft-lbs absorbed energy and 25 mils lateral expansion at +10*F.
DRAF~r Page 1 of 1 PCY:rm/A05121*-12 -
5/19/83 l
l
- LGS FSAR TABLE 5.3-12 LIMERICK 1 MSIV BODY DATA Appifcable Code: ASME Section III, W68, Draft Pump & Valve Code Valve Vendor: Atwood & Morrill Company Material Vendor: Quaker Alloy Casting Company Material Spec: ASTM A216 GR.WC8 Heat Number: F8304-1 C Mn Si P S Chemical Composition (wt.%) .26 .90 .30 .019 .012 Heat Treatment: Normalize 1700*F (7 hr. 10 min.) air cool
+ Temperature 1340*F (7 hrs.) air cool
' +Postweld heat treatment / stress relieve 1140*F/
1170*F (5 hrs. 10 min.) air cool t
l DRAPT PCY: ham:csc/005173* Page 1 of 1 5/18/83
- LGS FSAR TABLE 5.3-13 MSIV BODY DATA FROM OTHER BWRs Project: A Ap,,licable Code: ASME Section III, 1974 Valve Vendor: Atwood & Morrill Company Material Vendor: Quaker Alloy Casting Company Material Spec: ASME SA216 Grade WCB Heat Number: F6406 C Mn Si P S Chemical Composition (wt.%) 0.23 0.89 0.53 0.019 0.012 Heat Treatment: 1680/1710*F (5 hrs., 30 min.) air cool
+ Temperature 1350*F (5 hrs., 30 min.) air cool
+ Post weld 1200*F (6 hrs.) air cool Charpy V - Notch Impact Toughness Test Temperature: +60*F l Energy (Ft-1b): 32, 31, 34 Exp. (Mils): 33, 32, 31 l ' % Shear: 40, 40, 40 1
DRAyy PCY: ham:csc/005174* Page 1 of 6 l 5/18/83
LGS FSAR TABLE 5.3-13 MSIV BODY DATA FROM OTHER BWRs Project: B Applicable Code: ASME Section III, 1974 Valve Vendor: Atwood & Morrill Company Material Vendor: Atwood & Morrill Company Material Spec: ASME SA216 Grade WCB Heat Number: 35 C Mn Si P S Chemical Composition (wt.%) 0.24 0.82 0.46 0.022 0.013 Heat Treatment: 1650/1800*F (8 hrs.), air cool to 400*F
+ Temperature 1150*/1250*F (8 hrs. ), air cool
+ Post weld 1095*/1195*F (18 hrs.) furnace cool to 800*F (100*F/hr) air cool Charpy V-Notch Impact Toughness Test Temperature: +60*F Energy (Ft-lb.) 31.5, 37.5, 39.5 Exp. (Mils): ~33, 41, 40
% Shear: 10, 10, 10 DRAPT PCY:hmm:csc/005174* Page 2 of 6 5/18/83
LGS FSAR TABLE 5.3-13 MSIV BODY DATA FROM OTHER BWRs
- l. Project: C 1
Applicable Code: ASME Section III, 1974 with Summer 1975 Addenda Valve Vendor: Atwood & Morrill Company Material Vendor: Quaker Alloy Casting Company Material Spec: ASME SA216 Grade WC8 Heat Number: F3547 C Mn Si P S i
Chemical Composition (wt.%) 0.23 0.88 0.38 0.016 0.015 Heat Treatment: 1700/1725*F (6 hrs., 20 min.) air cool
+ Temperature 1345*F (6 hrs., 45 min.) air cool
+ Post weld 1200*/1225'F (6 hrs., 20 min.) air cool Charpy V-Notch Impact Toughness Test Temperature: +60*F Energy (Ft-lb): 66, 56, 54 l Exp. (Mils): 53, 50, 53
% Shear: 40, 40, 40 i
l PCY: ham:csc/005174* Page 3 of 6 5/18/83
, LGS FSAR TABLE 5.3-13 MSIV BODY DATA FROM OTHER BWRs Project: D Applicable Code:
ASME Section III, 1971 with Summer 1973 Addenda Valve Vendor: Rockwell International I
Material Vendor: Rockwell International l
Material Spec: ASME SA216 Grade WCC Heat Number: 1750262 i
C Mn Si P S Al i
Chemical Composition (wt.%) 0.21 1.19 0.43 0.011 0.009 0.043 Heat Treatment: 1700*F (10 hrs.) normalize
+1225*F (7.5 hrs.) temperature
+1100*F (6 hrs.) post weld Charpy V-Notch Impact Toughness Test Temperature: +40*F 1
Energy (Ft-lb): 29.0, 33.0, 35.0 Exp. (Mils): 25.0, 26.0, 30.0
% Shear: 15, 15, 15 i
DRAPT PCY: ham:csc/005174* Page 4 of 6 5/18/83
o LGS FSAR TABLE 5.3-13 MSIV BODY DATA FROM OTHER BWRs l
l Project: E Applicable Code:
ASME Section III,1971 with Summer 1973 Addenda Valve Vendor: Rockwell International i
Material Vendor: Rockwell International Material Spec: ASME SA216 Grade WCC Heat Number: 3760171 C Mn Si P S Al Chemical Composition (wt.%)
0.17 1.09 0.50 0.008 0.011 0.060 Heat Treatment: 1700*F (8 hrs.) normalize
+1275*F (8 hrs.) temperature i i
+1100*F (6 hrs.) post weld Charpy V-Notch Impact Toughtiess Test Temperature: +40*F Energy (Ft-lb): 35, 38, 29 Exp. (Mils): 32, 36, 29
% Shear: 20, 20, 20 l
DRAFT PCY: hem:csc/005174* Page 5 of 6 5/18/83
i s
LGS FSAR TABLE 5.3-13 MSIV BODY DATA FROM OTHER BWRs Project: F Applicable Code: ASME Section III, 1974 Valve Vendor: Atwood & Morrill Company Material Vendor: Quaker Alloy Casting Company Material Spec: ASME SA216 Grade WCB Heat Number: F7516 C Mn Si P S Chemical Composition (wt.%) 0.25 0.78 0.53 0.018 0.013 Heat Treatment: 1690/1710*F (6 hrs., 5 min.) air cool
+ Temperature 1350/1360*F (6 hrs.) air cool
+ Post weld 1200*F (6 hrs., 5 min.) air cool Charpy V-Notch Impact Toughness Test Temperature: +60*F Energy (Ft-lb.): 30, 24, 34 Exp. (Mils.): 37, 27, 33
% Shear: 40, 40, 40 DRAPy PCY: hmm:csc/005174* Page 6 of 6 5/18/83
LGS FSAR TABLE 5.3-14 LIMERICK 1 - MSIV BONNET COVER MATERIAL Applicable Code:
1968 ASME Nuclear Pump and Valve Code Valve Vendor: Atwood & Morrill Company Material Vendor: Cann & Saul Steel Company Material Specification: ASTM A105, Gr. 2 Heat No. 219222 C Mn Si P S Chemical Composition (Wt. %) .30 .68 .19 .009 .014 Heat Treatment Normalize 1650'F (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) - Air Cooled I
l i
Daggy .
Page 1 of 1 l
PCY:csc/105182-6
t LGS FSAR TABLE 5.3-15 MAIN STEAM SWEEP 0LET MATERIAL DATA FROM OTHER BWRs Project: A Applicable Code:
ASME Section III, 1974 Edition, 574 Addendum Vendor:
Bonney Forge Division, Gulf Wester _n Manufacturing Material Vendor: Sharch Steel Material Specification: SA 105N Heat No.:
631218 (Sharon Steel)
C Mn Si P S Chemical Composition (Wt. %) .28 .87 .22 .014 .015 Heat Treatment Normalize 1650*F (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) - Air Cooled Charpy V-Notch Impact Toughness (Longitudinal)
Test Temperature: +70*F Energy (Ft-lb.) 68.2, 83.5, 76.0 Lat. Exp. (Mils) 64, 71, 69
% Shear: 80, 80, 80 DRAyy Page 1 of 2 PCY:csc/IO5182-7
Project: A Applicable Code: ASME Section III, 1974 Edition, S74 Addendum i
Vendor: Bonney Forge Dii f sion, Gulf Western Manufacturing Material Vendor: Sharon Steel Material Specification: SA 105N Heat No.: 630614 (Sharon Steel)
C Mn Si P S Chemical Composition (Wt. %) .26 .86 .16 .022 .017 Heat Treatment Normalize 1650*F (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) - Air Cooled Charpy V-Notch Impact Toughness (Longitudinal)
Test Temperature: +70*F Energy (Ft-lb.): 76.6, 74.9, 62.0 107.7, 108.5, 109.3 Lat. Exp. (Mils) 68, 69, 63 75, 84, 85
% Shear: 80, 90, 80 100, 100, 100 h
Page 2 of 2 PCY:csc/105182-8
TABLE 5.3-16 SURVt1LLANCE CAPSULE Capsule Tensile Charpy V-Notch No. 1 3 Base Metal (BM) 8 BM, Long.
(Azimuth 300*) 4 Weld Metal (WM) 8 WM 3 Heat Affected 8 HAZ Zone (HAZ)
No. 2 3 BM 8 BM, Long.
(Azimuth 120*) 4 WM 8 WM 3 HAZ 8 HAZ No. 3 3 BM 20 BM, Long.
(Azimuth 30") 4 WM 16 WM 3 HAZ 12 HAZ NOTE: Each capsule also includes a Fe, Ni, and Cu flux wire. A separate neutron dosimeter is attached at Azimuth 30 and contains 3 Cu and 3 Fe flux wires.
l l
1 j Page 1 of 1 l
DRAPT PCY:rm/A05121*-9 5/17/83 l
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I LIMERICif GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REP'054%
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s DRAry LIMERICit GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT PREDICTED ADJUSTMENT OF REFERENCE TEMPERATURE, "A", AS A FUNCTION OF FLUENCE AND COPPER CONTENT F IGU R E '5.3'6
c LGS FSAR 15* 135* 255' 400.Y2"
-_____ .--____ _____ + 366.31" HEAT NO. HEAT NO. HEAT NO.
C7677-1 C7689-1 C7698-1 g BELTLINE
> ACTIVE FUEL ZCNE 77.3* 197.3* 317.3*
- _ _ - - .. ___ + 216.31" HEAT NO. HEAT NO. HEAT NO.
C7688-2 C7698-2 C7688-1 E h 5 126 1/2" i
SHOP ELDS: SEAMS BA, BB, BC, l 80, BE, BF FIELD WELDS: SEAM AB Q
LOCATIONS FOR LIMERICK 1 Page 1 of 1 l