ML20041E970
| ML20041E970 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 03/04/1982 |
| From: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20041E971 | List: |
| References | |
| 16011, NUDOCS 8203160034 | |
| Download: ML20041E970 (55) | |
Text
P e
,e Consumem Power P-James W Cook I
Vice President - Projects, Engineering and Construction oeneral offkes: 1945 West Parnell Road, Jackson, MI 49201 + (517) 7880453 March 4, 1982 M Ir?
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liarold R Denton, Director Office of Nuclear Reactor Regulation C
/8 03s US Nuclear Regulatory Commission Washington, DC 20555 (P
t-MIDLAND PROJECT
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MIDLAND DOCKET NOS 50-329, 50-330 SER OPEN ITEMS CONCERNING MATERIAL TOUGHNESS FILE: 0505.16 SERIAL: 16011 ENCLOSURES:
(1) SAR CHANGE NOTICE NO 3093 (2) RESPONSE TO SER SECTION 5.3 OPEN ITEMS (3) B&W TOPICAL REPORT BAW 1543 Enclosure (1) is being processed for inclusion in the upcoming FSAR revision (No 43) to be issued in April 1982.
This SAR Change Notice is being processed to resolve material toughness issues relating to SER Sections 5.1, 5.2 and 5.3 which were discussed in the January 15, 1982 meeting between Consumers Power Company / Babcock and Wilcox and Barry Elliot of the NRC Staff. Enclosure (2) provides responses to these open items.
SAR Change Notice No 3093 and our February 22, 1982 letter concerning unresolved safety issues both refer to B&W Topical Report BAW 1543. This report has not been previously submitted to the NRC Staff and therefore a copy is enclosed with this letter.
Relative to the draft SER, it is recommended that the referencing of the work being done by the B&W Materials Owners Group (as summarized in BAW 1543) in the enclosed SAR Change Notice No 3093 should be considered as an additional basis for the development of pressure temperature operating limits where the Midland Unit No 1 Weld WF-70 is the limiting material. Specifically, it is recommended that Page 5.3-5 of the draft SER be revised to read:
"During the plant's life, the applicant must calculate the P-T (pressure-temperature) operating limits based upon estimates of gD irradiated fracture toughness for limiting materials as determined by:
1.
RT Predictions based on Regulatory Guide 1.99 or; NDT oc0282-0033a100 8203160034 820304 PDR ADOCK 05000329 E
PDR J
2 2.
Fracture toughness tests of the limiting material or similar materials which have been exposed to appropriate neutron fluences."
As a result of the changes to the fracture toughness and fluence information, B&W has reevaluated the operating curves in the Technical Specifications (Figures 16.3.4-2A through 4B) and determined that these curves must be changed. Therefore, SAR Change Notice No 3093 deletes the existing curves from the Technical Specifications. New curves will be prepared based upon B&W Topical Report BAW 10046A, " Methods of Compliance With Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G."
In addition, it will be necessary to revise certain setpoints and methods utilized for overpressure protection at low temperature as discussed in FSAR Subsection 5.2.2.11.
Modifications to Subsection 5.2.2.11 will be provided by amendment.
It is hoped this additional material will be sufficient to allow for completion of the SER in this area and that the open items can be closed.
JWC/J:!L/dsb CC RJCook, Midland, w/o RHernan, NRC, w/o
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oc0282-0033a100
~
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,..P ENCLOSURE 1 QUAUTY ASSURANCE PROGRAM SAR CHANGE N3TICE
- 1. RENK Materials /CPCo FSAR 7220 JOB NO.
- 2. DISCIPLINE / COMPANY Pro.1 Engr /B&W
- 3. No. 3093
- 4. ORIGINATOR HWSlager, CPCo/WRGray&CJHudson, BgDATE 1/h/82
- 6. REFERENCED SECTIONS OF SAR Table 1.6-2, Appendix 3A-Reg Guide 1. 99, 5. 3.1. h, 5. 3.1.6.1.1, 5. 3.1.6.1. 3, 5. 3. 2, 5.3.3.9, Table 5.3-2, Tech Spec Table 16.h.h-5,16.3/h.h.9, Fig 16.h-1, Q&R 121.17, QLR 121.32, Response to Questions 123.1 & 123.2, Tech Spec Figures 16.3.h-2A,2B,3A,3B, hA, hB, 5.3 1. 5, T5. 3-2A, Q&R 121.11, 121.10, 121.20
- 7. DESCRIPTION OF CHANGE Factor in B&W Owners Group Program on surveillance capsules and prediction of upper shelf energies. Factor in the effect of lower anticipated end of the fluences.
Update noncompliances with 10 CFR 50 Appendices G & H.
Rosolve open draft SER questions.
- 8. REFERENCED SPECIFICATIONS OR DRAWINGS Hone
- 9. JUSTIFICATION Respond to NRC Reviewers questions.
- 10. BECHT2L DISCIPLINE INTERFACE REVIEW:
INTERFACING STAFF REVIEW:
O ARCH O PLANT DSN O ARCH O MECH O CIVIL O PQAE OCML O NUCLEAR.
O CONTROL SYS O STRESS O CONTROLSYSTEM O PLANT OSN O ELEC O OTHER _
O ELEC O RELIA 81UTY O MECH / NUCLEAR O GEOTECH O STRESS O MaOS D OTHER
- 11. REVIEWED BY DATE
- 12. REVIEWED BY DATE
- 13. REVIEWED BY DATE (Group Supervisor)
(SAR COORDINATOR)
(NUCLEAR ENGINEER)
- 14. CONCURRENCE BY DATE
- 15. APPROVED BY (CPCo)
DATE
- 16. CONCURRENCE BY DATE (PROJECT ENGINEER)
(NSSS SUPPLIER) mm m.n
" t*
MIDLAND 1&2-FSAR TABLE 1.6-2
,9 B&W TOPICAL REPORTS B&W Report Topical FSAR Approval Report Title Revision Reference Status p>
24 BAW-10000 Correlation of 0
4.4 Accepted by Critical Heat Flux NRC 2-21-73 in a Bundle Cooled by Pressurized Water BAW-10001 Incore Instrumentation 0
7 Accepted by 133 Test Program NRC 2-73 BAW-10003A Qualification Testing 4
7 Accepted by of Protection System NRC 10-10-75 Instrumentation l26 BAW-10008 Reactor Internals Pt 1, 3.9 Accepted by Stress and Deflection Rev 1 NRC 9-28-72 Due to LOCA and 24 Maximum Hypothetical Earthquake BAW-10010 Stabi'ity Margins for Pt 3, 4
Accepted by Xenon Oscillations -
Rev 1 NRC 12-70 Two and Three-Dimen-33 sional Digital Ana-lyses l39 BAW-10021 TEMP - Thermal Energy 0
4.4 Accepted by Mixing Programs NRC-12-29-70 BAW-10027 Research and Develop-0 5.4 Accepted by ment Report for the NRC 3-24-72 24 Once-Through Steam Generator BAW-10029A CRDM Test Program 3
3.9 Accepted by NRC 2-25-76 BAW-10035A Fuel Assembly Stress 1
4 Accepted by 126 and Deflection Due NRC 1-29-75 to LOCA and Seismic 24 Excitation
.c k.s (sheet 1)
Revision 39 11/81
s
'I-k' MIDLAND 1&2-FSAR TABLE 1.6-2 (continued)
B&W Report i
Topical FSAR Approval I
Rcport Title Revision Reference Status BAW-10036 Correlation of CHF in 0
4.4 Accepted by 24 a Bundle Cooled by NRC 2-21-73 i
Pressurized Water BAW-10037 Reactor Vessel Model 2
4.4 Accepted by Flow Tests NRC 12-19-72 BAW-10038 Prototype Vibration Supp. 1 5
Accepted by l33 Measurement Program Rev 1 NRC 4-4-79 139 for Reactor Internals BAW-10039 Prototype Vibration 0
3.9 Accepted by Measurement Program NRC 7-6-74 for Reactor Supp. 1 3.9 Under review Internals Rev 0 Submitted 26 8-79 BAW-10043 Overpressure Pro-0 5.2 Under review
)
tection for B&W's Submitted
~
PWRs 5-72 BAW-10046A Methods of Compliance 1
5.3 Accepted by with the Fracture NRC 7-77 24 Toughness Requirements of Operational Requirements of App G (10 CFR 50) l33 BAW-10051 Design of Reactor 1
3.9 Accepted by Internals and Incore NRC 7-6-73.
Instrument Nozzles Supp. }
3.9 Accepted by 24 for Flow Induced NRC 4-79 Vibration l33 BAW-10052 Multinode analysis 1
15 Accepted by of Small Breaks NRC 2-77 for B&W's 2568 39 l
MW t Nuclear Plants (sheet 2)
Revision 39 11/81
,e-
't.
f n 6tt h BAW-1511P 1rradiation-Induced Reduction in 0
3.A; Under review Charpy Upper-Shelf Energy of 5.3 Submitted 10/80 Reactor Vessel Welds BAW-1543 Integrated Reactor Vessel 0
5.3 Under review Material Surveillance Program Submitted 3/82 i
miO282-0034a100
g 1
mr I M % dv $
0 except for the predictions of upper shelf energies for the Unit #1 beltline weld material identified as WF-70 and for the surveillance capsule material identified as WF-209-1.
In lieu of Regulatory Guide 1.99, the B&W Materials Owners Group, of which Consumers Power is a member, has developed a more realistic prediction l
technique for estimating the drop in upper shelf energy at low fluence levels I
of up to 5 x 10 N/cm.
This prediction technique is contained in BAW-1511P,
" Irradiation Induced Reduction in Charpy Upper-Shelf Energy of Reactor Vessel Welds".
The chemistries and upper shelf energy level data, which were evaluated in the development of the BAW-1511P prediction technique, are shown in Tables 1 and 2 of Appendix C to BAW-1511P.
Included in this data are materials with similar unirradiated upper shelf energies and chemical compositions of the WF-70 and WF-209-1 materials.
Using the chemistry for the WF-70 material found in Table 3 on Page 4-4 of BAW-1511P and the unirradiated upper shelf energy of 66 ft-lbs, (see Table s r% d 5 3-2 4 5.3-2), the fluence required to lower the upper shelf energy to 50 ft-lbs was 6
calculated using the empirical model on Page 2-2 of BAW-1511P and found to be l
10 greater than the 5 x 10 N/cm applicability for the model.
Similarly, the chemical compositions and unirradiated upper shelf energy of s
Aesd #;3-Es4 TablY5.3-2 were used to demonstrate equivalent results for the WF-209-1 e
4 material.
The predictions of the reference temperature (RTNDT) adjustment and Charpy V-notch upper shelf energy (C USE) reduction due to neutron irradiation will be y
l miO282-0034a100
.s.
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8 Cos4h u r c/
_ _s confirmed by withdrawal and evaluation of B&W Owners Group surveillance capsules which are being irradiated in reactors of similar design and construction, as described in BAW-1543. These capsules, containing specimens of WF-70 weld metal (in addition to others), are scheduled for withdrawal at fluence levels corresponding to the end of life fluence in Midland I at the l
l 1/4 thickness location.
Withdrawal and testing will occur several years before the WF-70 weld in Midland 1 is predicted to exhibit a C USE of y
50 ft-lbs or less.
As a result of Consumers Power Company's participation in the B&W's Owners Group, the requirements of 10CFR50, Appendix G, Parts V.C, V.D, and V.E will be satisfied if applicabl7 These requirements involve the detenmination of actual fracture toughness of the RV materials, the application of alternative analysis techniques and the performance of a thermal annealing process.
miO282-0034a100
- s.
MIDLAND 152-FSAR P/ '*
5.3.1.3 Special Methods for Nondestructive Examination
(-
The required nondestructive examinations carried out during fabrication are presented in Table 5.4-5.
These inspections are performed in accordance with procedures meeting the requirements 33 of the edition and addenda of the ASME Code,Section III listed l
in Table 5.2-1.
Nondestructive examination techniques used are l
selected to provide adequate sensitivity, reliability, and reproducibility to inspect surfaces and detect internal l
discontinuities.
Acceptance standards are in accordance with the requirements of the ASME Code,Section III for the given product l33 and/or fabrication process.
5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels The controls on welding, composition, heat treatments, and similar processes covered by Regulatory Guides 1.34, 1.43, 1.50, and 1.71 for ferritic materials in the reactor vessel are addressed in Subsection 5.2.3.3.
Subsection 5.2.3.4 addresses similar controls covered by Regulatory Guides 1.31, 1.34, 1.37, 1.44, and 1.71 for austenitic stainless steels in the reactor vessel.
See Appendix 3A for a discussion of conformance with 33 regulatory guides.
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l 24 svuucuu o.m 5.3.1.5 Fracture Toughness The pressure boundary'ferritic materials of the reactor vessel were ordered and tested in accordance with the requirements of the ASME B&PV Code,Section III, and addenda listed in Table 5.2-1.
%.P* All the pressure boundary ferritic materials of the reactor vessel met the required Charpy V-notch impact energy at a
temperature of 40F or lower as discussed in Subsection 5.2.3.3.1.
39 In addition, the reference temperature (RTuor) and Charpy V-notch upper shelf energy (CyOSE) of the Midland reactor vessels hee.kavs been obtained directly from most of the beltline materials in accordance with the 1972 Summer Addenda to the ASME B&PV Code,Section III, and Appendix G to 10 CFR 50.
As discussed in Subsection 5.2.3.3.1,' BAW-10046A provided (WFfD@e basis for estimating RT or (and CyUSE) for weld O) for which I41 N
material ~s were not available.
The pertinent fracture toughness data for the beltline region of the Midland Plant Units 1 and 2 reactor vessels are presented in Table 5.3-2.
This table also 39 includes predicted changes in RTuor due to neutron irradiation based on Regulatory Guide 1.99 (refer to Appendix 3A).
Also listed are the properties of the atypical weld metal based on g
Revision 41 5.3-3 2/82 l
l MIDLAND 1&2-FSAR i
BAW-10144A, although the existence of this weld metal in the 39 Midland it i reactor vessel in the upper to lower shell girth weld (W
- 0) is considered highly unlikely.
141 L
F The RT c ata presented in Table 5.3-24 and the estimated RT or 141 uoy N
values discussed in Subsection 5.2.3.3.1 will be used to establish operating limitations on pressure and temperature for 39 the Midland plants, in accordance with the procedures defined in BAW-10046A, as discussed in Subsection 5.3.2.
5.3c2.,6 Material Surveillance Throughout the lifetime of a reactor vessel the fracture toughness, impact, and tensile properties of the ferritic beltline region materials will change because of neutron irradiation.
These changes require periodic adjustments of pressure-temperature relationships for heatup and cooldown during normal, upset, and testing conditions.
The effects of neutron irradiation typically include a reduction in impact energy (as measured by Charpy V-notch tests),
a reduction in fracture toughness, an increase in the 8
brittle-to-ductile transition temperature, and an increase in yield and tensile strengths.
The purpose of the surveillance program is to monitor changes in the fracture toughness impact and tensile properties of the reactor vessel beltline region materials.
Because the NSS-12 and NSS-13 vessels contain different materials, a separate surveillance program has been developed for each.
(NSS-12 and NSS-13 reactor vessels will be utilized in Midland plant Units 1 and 2 respectively.)
Previously, changes in fracture toughness due to irradiation were assessed only by means of Charpy energy shifts and reference 28 temperature changes (ARTway).
This required the use of indirect empirical relationships between Charpy energy data and fracture toughness.
Precracked compact fracture specimens afford a more direct means of evaluating fracture toughness.
These have been included in the surveillance program to supplement the Charpy data.
8 Test data obtained from the surveillance programs provide a determination of the pressure-temperature conditions under which 1
/
Revision 41 5.3-4 2/82
1 f*.
MIDLAND 1&2-FSAR the reactor vessels may be operated.
The programs meet or exceed the requirements of Appendix H to 10 CFR 50.
ks 5.3.1.6.1 Basis for Material Selection 5.3.1.6.1.1 Selection Criteria Only the ferritic materials in the shc11 courses and welds 8
surrounding the core of.the reactor vessel receive neutron fluences great enough to significantly change their mechanical properties.
This section of the reactor vessel is referred to as the " beltline region. " Appendix G to 10 CFR 50 defines this beltline region to include all shell material directly surrounding the effective height of the fuel element assemblies plus any additional material with a predicted RT of SOF or Nor more over the service life of the vessel.
The ferritic region materials are listed in Table 5. 3-2A.
In the table, as well as l39 throughout this discussion, weld materials are designated according to their weld qualification number.
With the exception of one weld material in the NSS.12 program, base metal, heat-affected zone (HAZ), and weld materials within the vessel beltline region that are expected to experience the most severe changes of material properties due to irradiation are selected for surveillance monitoring.
Due to a scarcity of one weld metal used in the NSS-12 reactor vessel, WF 70,Pishi;hly 45 2
(
" -i l
's-...; heat of ::ld 'i_cj" weld metalP WF 209) 8 L
substituted for this particular material during certain phases of the surveillance program.
Details pertaining to the selection and use of WF 209 appear in Subsection 5.3.1.6.1.3.
The rest of the surveillance materials were selected in accordance with the procedure described below.
d s.3-2 A Tables 5.3-2 provideI~ a tabulation of the materials and their 3
properties which comprise the beltline region of the two Midland units.
. Table 16.4.4-5 indicates which material was finally l 39 selected for surveillance from each of the three vessel regions.
The following were factors in the material selection process:
a.
Unirradiated RT for specimens in the " weak" (axial Nor direction of the vessel) direction 8
b.
Unirradiated Charpy upper shelf energy level for specimens in the " weak" direction c.
Weight percent of trace elements within the materials known to enhance radiation embrittlement (such as copper and phosphorous) d.
Material's end-of-life peak neutron fluence (E >1MeV) at l 33 the vessels 1/4 thickness location 3
e.
Predicted shift in the reference temperature y,
i Revision 39 5.3-5 11/81
MIDLAND 1&2-FSAR ad a h g.,4 h ra fa.-ance da ~ P "' 0"" " A f.
Predicted drop of the Charpy upper shelf energy level 8
In order to comply with Appendix G to 10 CFR 50, beltline region m terials were characterized by establishing the reference
'h!
temperature, RTuoy, in the " weak" direction in accordance with l
the Summer 1972 Addenda of the ASME Code,Section III, Paragraph 28 NB 2300.
In addition to the Charpy tests needed to determine the RT Charpy impact tests were conducted at appropriate
- uoy, temperatures over a temperature range sufficient to define the full Charpy impact transition curve (including the upper shelf energy).
The data reported ineluded fracture energy, specimen's lateral expansion, and percent of shear fracture.
The specimens ware oriented in the " weak" or axial direction of the vessel with the V-notch of the Charpy specimens parallel to the vessel's 8
radial direction, in accordance with the requirements of the Summer 1972 Addenda of the ASME Code,Section III, Paragraph NB 2300.
The predicted adjustment in reference temperature (which is used to define the beltline region, to select surveillance materials, and to help determine the capsule withdrawal schedule) is determined in accordance with Regulatory Guide 1.99, Rev.
1.
This appears in Table 5.3-2 for several beltline region materials.
The predicted drop in the Charpy upper shelf energy 18 level is also, determined using Regulatory Guide 1.99 a reep + a.s dtsc r e~b e d o.s AtPan d * * ^^
j The reference temperature adjustment predictions and USE shif t predictions of Table 5.3-2 are based on conservative fluence calculations.
~
5.3.1.6.1.2 Test Material Origins Specimens are machined f rom as-f abricated metal, weld metal, and 8
HAZ metal.
Material from each heat of steel used in the reactor vessel beltline region is normally set aside and stored for use in the program.
The surveillance test weldments were prepared following the same fabrication history, such as welding parameters, post weld heat treatment, etc, as the beltline region materials of the reactor vessel.
Specimen sections are obtained by cutting a test ring of material f rom each reactor vessel shell course.
This is accomplished before the vessels' outside diameters undergo final machining.
A 4-1/2 to 5 inch thick section is cut from each test ring.
This section is centered about the 1/4 and 3/4 thickness of the i
original shell course by optimizing the original radius of J
5.3-6 Revision 28 5/80
l.
MIDLAND 1&2-FSAR t
l curvature.
Figure 5.3-2 illustrates the method used.
The WF 70 l 33 l /f '
test material was obtained from nozzle dropouts of the NSS-12 reactor vessel.
8 l
The specimen sections also provide ample metal for archive material.
Af ter all finished specimens are loaded into capsules, the archive material is permanently identified, recorded, and
[ 33 retained in the form of rough-machined plates until additional specimens are required.
The exact amount of archive material available cannot be determined until the test specimens have been machined; however, with the exception of WF 70, enough metal for I
l at least two capsules per heat including weld and HAZ will be available.
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,weQs sye 5.3.1.6.1.3 Selection and Use of WF 209 Weld Metal s
and 5.3-2 A TablMP.3-2 11stJ*the pertinent properties of WF 70 and WF 209.
5 4
- t _ ks.. f.
. the same heat of weld wire, but The twoA;; t - 1:
dif fe rent fluxes.
The suitability of WF 209 as a substitute for WF 70 b ::x;; ; _ 2.... o rm..
- _...r-l--+ vs Preuiv u J..f;
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c _. :;; I= I l a. -,*Vg e -F M <....; ou m...., d A Jc. &
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As mentioned in Subsection 5.3.1.6.3, the NSS-12 surveillance program calls for a total of six capsules.
The weld metal specimens in one capsule will be prepared from both WF 70 and
[.
WF 209 in ord'er to obtain a correlation between the two weld
\\
metals.
The other five capsules will contain weld metal specimens prepared from WF 209.
The number and types ot specimens to be contained in each capsule l
arc discussed in Subsection 5.3.1.6.2.1 and are listed in Table 5.3-3.
The specimens in the one capsule containing both WF 70 and WF 209 are also listed in Table 5.3-3.
The number of specimens in this capsule are suf ficient to obtain a correlation 22 between WF 70 and WP 209 while meeting all the re,quirements of ASTM E185-1973.
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,mm l
The remaining WP 70 weld metal will be placed in the reactor vessel surveillance program material archive storage.
Should future testing needs arise which require the use of any of the remaining WF 70 weld metal, details of the test program will be submitted to the NRC for evaluation and approval.
Revision 33 4/01 5.3-6a
/*
MIDLAND 152-FSAR 4
THIS PAGE INTENTIONALLY LEFT BLANK
)
I Revision 22 6/79 5.3-6b
1
.v MIDLAND 1&2-FSAR BA tt)-/fil P a.e d y
5.3.2 PRESSURE - TEMPERATURE LIMITS t~
B&W Topical Report BAW-10046A provides the bases for setting 39 operational limits on pressure and temperature.
This topical report provides detailed assurance that, throughout the life of (41 the plant, operations will comply with requirements of Appendix G l
of 10 CFR 50.
Regula tory Guide 1.99 ( refer to Appendix 3A); andAL omyylwol.s l im. -, Q-39 BAW-10144A)L"'ic' c l'-'a-th: cff;;;
v-une re used to predict the ef fects of neutron irradiation on the beltline region materials.
For assurance of compliance with l41 Appendix H of 10 CFR 50 throughout the life of the plant, refer 39 to Subsection 5.3.1.6.
5.3.2.1 Limit Curves Topical Report BAW-10046A provides the following information:
l26 a.
Procedures and criteria used b.
Safety margins c.
Bases used to determine the limits d.
Procedures that will be used to revise the limits i33 The operational limits of pressure and temperature for the 34 C~
following conditions are provided in Subsection 16.3/4.4.9.
a.
Inservice leak and hydrostatic tests b.
Normal operation, including heatup and cooldown 1
1 c.
Reactor core operation The limits on pressure and temperature for the preservice system hydrostatic test are provided in Figure 5.3-1.
The actual 133 material toughness test results are given in Subsection 5.2.3.
l 5.3.2.2 ope ra t ing Procedures Refe r to Subsect ion 5. 2. 2.11 and Sect ion 6.1.
33 5.3.3 REACTOR VESSEL INTEGRITY l
The summary description of the reactor vessel, including major considerations in achieving reactor vessel safety and vessels contributing to the vessel's integrity, is contained in Section 5.3.
B&W is the reactor vessel designer and fabricator.
Revision 41 5.3-13 2/82 1
1
MIDLAND 1&2-FSAR 5.3.3.1 Design S
The ASME Code,Section III, is the primary design criteria for the reactor vessel.
Chapter 5 describes the reactor vessel design, including construction features and arrangement drawing.
Materials of construction are listed in Table 5.2-3.
The design code is given in Table 5.2-1.
Table 5.1-3 gives the design basis values used in the design.
5.3.3.2 Materials of Construction The materials of construction for the reactor vessel are listed in Table 5.2-3.
Special requirements, reason for selection, and suitability of the materials used are included in Subsections 5.2.3 and 5.3.1.
The materials selected have been used extensively in nuclear vessel construction and exhibit well defined properties and serviceability.
5.3.3.3 Fabrication Methods Fabrication methods used in constructing the reactor vessel are described in Subsections 5.2.3.3, 5.2.3.4, and 5.3.1.
The suitability of the fabrication methods is demonstrated by the excellent service history of vessels constructed using these methods.
l l
l Revision 34 5.3-14 6/81 l
l
MIDLAND 1&2-FSAR insulation to protect adjacent concrete and structural members chould facilitate in-place reactor vessel annealing, if required.
3 3
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l l
Revision 33 l
5.3-16 4/81 l
^------------r-.
MIDLAND 1&2-FSAR 5.3.3.4 Inspection Requirements
(
Fabrication inspection requirements imposed on the reactor vessel l 28 are summarized in Subsections 5.2.3.3 and 5.2.3.4, and Table l 33 5.4-5.
Preservice and inservice inspection requirements are summarized in Subsection 5.2.4.
28 5.3.3.5 Shipment and Installation As described in Subsection 5.2.3.4, B&W specifies cleanliness requirements during shipment of the reactor vessel to ensure its arrival at the site in satisfactory condition.
B&W also provides appropriate instructions and consultation to the owner for onsite cleaning and vessel protection.
Temporary protective coatings and/or covers are applied to the vessel during shipment and storage as appropriate for expected environmental conditions.
Water chemistry is controlled during initial fill, testing, and operation of the vessel to prevent an environment that may be conducive to material failure.
5.3.3.6 Operating Conditions The operational limits specified to ensure reactor vessel safety are described in Subsection 5.3.2.1.
These are compared with normal intended and upset operating conditions in Subsection 5.3.2.2.
The design transients for the reactor vessel are specified in Subsection 3.9.1.1.
5.3.3.7 Inservice Surveillance A discussion of the reactor vessel material surveillance program is given in Subsection 5.3.1.6.
The inservice inspection program is discussed in Chapter 16.
5.3.3.8 In-Place Annealing of the Reactor Vessel The design of the reactor vessel does not preclude in-place stress annealing of the reactor vessel if annealing is determined to be required.
3 Space provisions include a three foot annular clearance around the core belt region (after insulation removal) with top access to the annular region.
While there are ongoing industry wide programs aimed at determining methods for in-place annealing, the required time-temperature parameters and specific heating procedures have not been established.
The Midland plant design together with especially designed temporary provisions such as heat source and
\\
Revision 33 5.3-15 4/81
I nse 4
C
=
5.3.3.9 Compliance with 10 CFR Appendices G&H The requirements of 10 CFR 50 Appendices G (as officially corrected on November 20, 1979) and H (with amendments through the September 26, 1979 amendment) have been satisfied in the design and fabrication of the-Midland units except for the following paragraphs:
~.
l i
l miO282-0034a100 l
l
f M $dyk 0
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f 4
=
Appendix G, Part IIIB, Paragraph 4 a.
Requirement
" Individuals performing fracture toughness tests...shall have demonstrated competency to perform the tests in accord with written procedures of the component manufacturer."
b.
Description - At the time of the initial tests of Midland Plant Units 1 and 2 reactor coolant pressure boundary (RCPB) materials, no standard written procedures existed to demonstrate competency of individuals to l
perform tests.
l c.
Comments - Subsequent to initial testing of Midland materials, standard l
documentation was developed to demonstrate materials testing competency.
At the time of the initial tests, it was industry practice not to document material testing qualifications.
However, individuals were experienced I
and competent to perform the material tests in question.
I l
l l
miO282-0034a100
/
Insw 4 G
CPra 3)
~
Appendix G, Part IIIC, Paragraph 2 a.
Requirement "Where seamless shell forgings are used...the test specimens ay be taken from a separate weldment provided that such a weldment is prepared using excess material from the shell forging (s) or plates, as applicable, the same heat of filler metal and same production welding I
conditions as those used in joining the corresponding shell materials."
b.
Description - Weld metal mechanical and impact toughness properties are obtained from test specimens taken during the weld metal and procedure qualifications required by Sections III and IX of the ASME B&PV Code.
The base metal used in the preparation of the required test weldments is P3 (SA508C12 and SA533GRBCll) which is the same ASME Section IX "P" number as the SA508Cl2 vessel forging materials.
The thickness of the test weldment is about 4 inches vs the miniumum reactor pressure vessel wall thickness of about 8 inches.
Experience has shown the heat treatment response and resultant properties of the 4 inch thick test weldment provide and accurate representation of actual pressure vessel weldments.
The welds are made using the same (ASA) automatic submerged arc weld process within the range of procedure weld variables specified and used during actual reactor pressure vessel weld fabrication. The "WF" weld l
l metal designation uniquely describes the filler wire heat and flux lot l
employed during the ASA weld process.
The stress relief cycle employed during the weld metal and procedure qualification was performed at the same temperature (1125 i 25F) used during vessel fabrication.
1 l
miO282-0034a100 l
s.
y T
I n ser -h 0
( Pa-y G b s
The stress relief time is purposely set at 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> to surpass the accumulated stress relief time the actual vessel would see in order to achieve a conservative representation of material properties. The total stress relief time for the Midland Units 1 and 2 reactor vessels were 22-1/2 and 16-1/2 hours, respectively.
The stress relief times for l
particular welds in the vessel will vary due to the fabrication sequence since intermediate post weld heat treatments are required.
The measurement of reference temperature (RTNDT) was n t required by the edition of the ASME B&PV code governing construction of the Midland Unit.s 1 and 2 RCPB components.
However, in virtually every case, it as possible to retrofit the measurement of RT f r the reactor vessel NDT materials. As discussed in Section 5.3.1.5, the only exception is the middle circumferential weld (WF-70) in the Midland I reactor vessel.
c.
Comments - Data obtained from test specimens fabricated as described above provide information which is equivalent to data which would be obtained by complying with 10 CFR 50 Appendix G, Part IIIC, Paragraph 2.
i l
l Appendix G, Part IVA, Paragraphs 1 and 3 Requirement - Paragraph 1 "The materials shall meet the acceptance a.
l l
standards of Paragraph NE-2330 of the ASME Code...."
l Paragraph 3 " Materials for piping, pumps and valves shall meet the l
requirements of Paragraph NB-2332 of the ASME Code. Materials for bolting and other fasteners shall meet the requirements of Paragraph NB-2333 of the ASME Code."
miO282-0034a100
v~
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h S A.e k A L
=r b.
Description - The requirements have been met for all beltline materials except for one weld in Midland Unit 1.
In addition, RCPB vessel materials which are located outside the beltline region were fabricated to an earlier Code which did not require the determination of the RT I"
NDT*
lieu of testing each of the materials, generic upper bound RT values NDT were developed for all materials utilized in the vessel RCPB. Those values are contained in BAW-10046A, Rev 1, which was evaluated by the NRC in a letter from S A Varga to J H Taylor dated June 22, 1977.
c.
Comment BAW-10046A, Rev 1, provides sufficient information for determining the limiting RT f r ferritic RCPB materials.
NDT Appendix G, Part IVB Requirement
" Reactor vessel beltline materials shall have minimum upper-a.
shelf energy...of 75 ft lbs unless it is demonstrated to the Commission by appropriate data and analyses that lower values of upper shelf fracture energy still provide adequate margin for deterioration from irradiation."
l l
- b. ' Description - Based upon Table 3-2 of BAW-10046A, Rev 1, it is estimated l
that the unirradiated upper shelf energy for one beltline weld in Midland Unit 1 may be as low as 66 ft-lbs, c.
Comment - During the initial period of operation of this unit, adequate l
l margin for deterioration from irradiation will be maintained by exceeding i
50 ft-lbs upper shelf energy.
It is predicted that this level will be I
2 maintained for fluence levels of at least 5 x10 N/cm which is equivalent to 15.1 effective full power years (see Appendix 3A and Table 5.3-2).
miO282-0034a100 l
t
Y v
D-1M34e C(
b) t.
If the upper shelf energy level falls below 50 ft-lbs, it is expected that vessel integrity, with adequate margins of safety, will be demonstrated.
This demonstration will use the results of fracture toughness tests on irradiated materials which are similar to the weld in question in Unit 1.
A description of the test specimens and irradiation program for these tests are contained in BAW-1543. Fracture toughness data and appropriate analytical techniques are being developed by the B&W Materials Owners Group.
Appendix H, Part IIR Requirement
" Reactor vessels...shall have their beltline regions a.
monitored by a surveillance program complying with... ASTM-E-185-73..."
In turn, ASTM-E-185-73, Annex A1, required that the materials included in the program shall be those predicted to be most limiting, with regard to setting pressure-temperature limits, for operation of the reactor.
b.
Description - The Midland Unit I surveillance program does not comply with the cited requireisent in that the most limiting base material (and l
?('
l therefore the most limiting heat affected zone (HAZ) material) is not included in the surveillance program and only a limited amount of the most limiting weld material is included in the surveillance program.
r l
The Midland Unit 2 surveillance program does not comply with the cited i
i requirement in that the most limiting weld material is not included in tha surveillance program, but the most limiting base material (and therefore the most limiting HAZ material) is included in the surveillance program.
t miO282-0034a100 s
s
V W
In s a. 4 C C h O.7.);
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c.
Comments - The weld material ir Unit 1 is predicted to be significantiy gole limiting than the, base' material and HAZ, therefore, use of in1ternative base materials will have no effect on the setting of pressure-
.tdmper'ature limits'for the Unit 1 RCPB. The most limiting weld metal in
~~this unit is identified by its weld procedure qualification number (WF-70).,Becau's'e this material is scarce, otly a limited amount of this i
c, i
material is included in the surveillance program.
In lieu of more
\\
-Jcxtensive utilia tion of this material in the Midland Unit 1 surveillance i
prhgram,}simN.armaterial(WF-209)isusedasdescribedinSubsec-
\\
tion 5.3.1;6.1.3.
In additiors) VF-70 material is being irradiated as part of the B&W t
i i
1 Materials Owners Gcoup progran
, described in BAW-1543.
Therefore, 3
s.
t.
s sin addati.on to using. thepbdictedshiftsinRT Regulatory ET s<A
\\
- G%, te 1.99 and the measur{d effects of irradiation on a similar material
(
s.
> f (W-209), it is expected that the Materials Owners Group program will
\\-
<y provide data on the effects of irradiation on the limiting Unit I weld l
material that will be useful in developing pressure temperature limits for operation.
V l
The base materiril ft Unit 2 is predicted to be more limiting than the weld
~
I i
metal, therefore,-use of alternative weld materials will have no effect on the setting of pressure-temperature limits for the Unit 2 RCPB.
l l
)
3 1-Appendix H, Part IIC, Paragraph I
/
x s
Requitcment'
" Surveillance specimens shall be taken f rom locations a.
alongside p fracture tciughness test specimens required by Section III of Appendix 6."
.c.
i miO282-0034a100 s,
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I n sM e C F=7 c 8)
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r b.
Description - This requirement has not been complied with for one beltline weld in the Unit 1 Reactor Pressure Vessel, c.
Comment - As the noncompliance with Appendix G, Part IVA, Paragraph 1, indicates, the fracture toughness tests of Section III of Appendix G were not performed on this weld, therefore, the surveillance specimens cannot be taken from adjacent locations.
l l
l l
l l
l l
miO282-0034a100
MIDLAn*D lt,3-FSAR TAOLE 5.3-2 MATERIAL PROPERTIES 39 BELTLINE REGION OF MIDLAND 162 REACTOR VESSELS Impact ToughnessPropertieN Fluence At Predicted Ef fect of Irradiatio Drop 35 g gFndgof-Life Red uc-Material Beltline Weight 50 Mils Chemistry
.u/cm. E)1MeV Adjusted tion be Adjusted 41 Location) TRegiorg (k
(*t-lb L.E. (ft-lb)(* W (Cu Ident if -
Materi 1 F
Use RT P
Wall location AR?
RT C Ug CUg ca tion 8)
T 0)
F) i'F) t)
(6)
I. D.
394T/V
(*
(
ft)
(Yt-lb)
Unit I sg LM 8.0GX
&8 All zo a tl 39 ASM 155 SA 508 C1.2(5) Nozzle 30 70 70 140 30 0.13 0.006 Jul0 N
M M
E M
II
?422:212)
Belt N
$m s.W ai 19 WF 336 Weld NB/US
-30 30 A0 77
-30 0.03 0.004 4x10 4 40x10 W
- too Jff8 M4'3 III SA 508 C1.2(5) Upper 20 40 40 140 20 0.02 0.008 Jul0 Wu10 4euro 7K70 M89 Lkft8J l41 18 18 ABZ 196 (532598)
Shell
'**4' O(6)(20)(6) M M $ 10 I4I I4I M(4) 18 x10 ' MM7 MW Mii) 1 yq p g
g')
WF 70 Weld US/LS M
[.
gl E.20 [ M[J 50]
M E903 Co 413Co o21'J /.w/e* /.ev,m'se 141 I I 18 18 ACA 19 7 SA 508 C1.2 tower 20 10 10 127 20 0.02 0.010 10 10
.W8d 4T7Z
.3M F MHr/03 39 (522366)
Shell II'3I o,3s o 048 s.9 s.84*
WF 209-I Weld SJrve11-20 70 70 76 20 4 49.6-ete Jul0II.L-98x10II M1287 MTJO'/ 4e"Ol).4t~C88) l 41 lance Unit 2 AAY 195 SA 508 C1.2(5) Nozzle 0
40 40 154 0
0.01 0.005 10 10
<50
<50 J513
..&fd3Y 18 18
( 3 P-28 70 )
Belt 18 18 WF 447 Weld NB/US 10 70 40 77 10 0.05 0.010 10 3g
<50
<60 Mt3 6fM III 0
D D'ZB 2 4 3 SA 500 C1.2(5) Upper 10 70 70 96 10 0.03 0.016 10 10 Jef St.
L&f 9&
.3t"19
.7f 73
( S P-410 6 )
Shell III 1
19 WF 336 Weld US/LS
-30 30 30 77
-30 0.03 0.004 10 '
x10 4t 4fD M QO )r19 W 62.
I I D
18 BFF 247 SA 508 C1.2 tower 10 70 70 98 10 0.01 0.012 10 10 M*6F-36* 7 f.
21* f f F79 l41
( 4 P-36 2 3)
Shell (see Ag**d '
% #,'d *** *
- III Surve illance material.
39 I,2) tn accordance with Regulatory Guide 1.9%and fluence at 4ftT vessel wall location.
'3I 4
lmpact properties are determined from speciaa weldsent eaterial and not frca a. t.va1 Midland Unit I surveillance material,
!'I Abt available.
'S 50 1.2 = A-508-64, Class 2 (Code Case 1332-4.
- g (6
- ' O h4 #***'E*""
"I~*
- f 4 "'88 O * *1 8 A
/
dt Estamated from BAW-10046A, Rev Is NCh ds Yd f h b **K "1 M op e* *.
e I7I Based on BAW-10144A, Evalwation of Atypical G ht94-4 8 E8'* f 4-8,
O m-(s') De n f o rs d ha m,
- a. s 6.o 1 // d r,,',. y,
a 4,,,D
- e. M a's 4*a s.
- 6) Os*Lfore d
-C.,
y }.e,,,, I e,,,4, g ;e,4,,,,,_,,, 4,,
,,4*g.
(oo) Ba.ssd wro
- =. 4s-LIs
- e. ea M4.'s 4,*.,,, a w g gu p (u,,;4
.g _ s,),
(ss) A s f. die.a.h sd by B A W - oSH P, of ap
.J, go p. jgs w ; g
,,,.4 M g,,.e s~a to '# d/c e. ' QS l E'W
,w
.+ Fy W.;
A.,.4 s s ).
(! 1) P r e p r s'e }. a.,w.y ~ s < < d A W ~/ CU P sP*yLV~Y Cs 3 ) (*en = e s. 4 a.4 e c/ v. /., e.
9 MIDLAND 182-FSAR s.
9
?
e TABLE 5.3-1 HAS BEEN DELETED C
O 5
L.
e O
'e 9
Revision 39 11/81
MIDLAND 1&2-FSAR TABLE 5.3-2A REACTOR VESSEL BELTLINE MATERIALS 39 Heat Number.
1 Base Fluu Tensile Base Metal.
Let i
Metal "
or Weld No.881 Beltline Propert.es t
or We?d'8 Filler (Welds Region Chemical cosaposition (Wt.Y)
UTS Y.S.
i 1.D. No.
Wire Only)
Location C
Mn P
S Si Ni Cr No Cu (ksi) (ksi) 141 UNIT 1 Nozzle Belt 0.24 0.56 0.006 0.007 0.24 0.73 0.32 0.61 0.13 97.4 76.9 ASK 155 122x212 WF 336 442002 8873 NB/US 0.12 1.20 0.004 0.016 0.32 0.46 0.09 0.30 0.03 80.5 63.0 Upper Shell 0.20 0.66 0.008 0.011 0.19 0.74 0.42 0.60 0.02 86.3 60.5 ABZ 196 532598 WF 70(N
-7tt95(1) 46Ml(F)US/LS h
h h
dl b h
M 07 85.5 69.0
{0.07
.67] Q.021) {0.012} {1.00] {0.10} (0.07} @.46] (0.41} (92.0] {4.3}
ical ACA 197 522366 Lower Shell 0.20 0.60 0.010 0.015 0.19 0.76 0.33 0.61 0.02 84.5 58.9 WF 209-1 7MG6-(N SGMSurveillance 9:1rF 4,64*
evete
+:90T ev44r 9,46-D-it 4,99 9790 87.5 72.5 i
Weld o.09
/, (o?.
c.0 8 A o.ost o.59 c,59 c.10 o.'do c.35*
l 39 UNIT 2 AAY 195 3P-2870 Nozzle Belt 0.21 C.67 0.005 0.009 0.28 0.75 0.38 0.61 0.01 85.9 62.8 WF 447 442002 8064 NB/US 0.08 1.37 0.010 0.013 0.45 0.62 0.097 0.41 0.05 83.3 64.5 BZS 243 SP-4106 Upper Shell 0.18 0.61 0.016 0.010 0.25 0.75 0.32 0.52 0.03 90.6 62.4 WF 336 442002 8873 US/LS 0.12 1.20 0.004 0.016 0.32 0.46 0.09 0.30 0.03 80.5 63.0 Surveillance i
Weld BFF 247 4P-3623 Lower Shell 0.22 0.65 0.012 0.016 0.30 0.74 0.33 0.55 0.01 89.9 66.4 I
"8 Base metal, SA 508 64 Class 2. Code case 1332-s.
888 Welds made by the automatic submerged arc process.
83'Linde 80 weld flux.
- *' Based on BAW-10144 A, Evaluation of Atypical We&d-Metert b/elelsuas h.
$ Se e. 6 /8 W -l St/ P, Tk. 6 /t 10 C psye 8-Z8) Soe e.L,ss is 4 *E8,s -
& 8 a.s a. d a n om e. 4 t.re's.l o <, d
- W n't. a. S****s3 o m SL M
- N ** wI'*-
(f) 4 ea 8#9(J-/ Sit P, Ta.6/s F ( ra.9t 8-Esp),
Table 5.3-2A
- i " 41 h) (*A) = e,sJianan,/eaf v a. I e t.
4 3
I
?
f 2600 1
'400 THE ACCEPTABLE PRESSURE AND TEMPERATURE COMBINATIONS ARE BELOW AND TC j
2200 THE RIGHT OF THE LIMIT CURVE. THE REACTOR NUST NOT BE MADE CRITICAL U
THE PRESSURE TEMPERATURE COMBINATIONS ARE TO THE RIGHT OF THE
~ CRITIC TT LIMIT CURVE. MARGINS OF 25 PSIG AND 10*F ARE INCLUDE 0 FOR CRITICAllTY POS$tBLE I UuENT ERROR.
(guit E.
1600 3
a 1400
[
PolNT PRESS.
TEMP 1200 A
395 70 B
556 175 1000 C
556 240 0
3 O
l 1000 248 800 E
2250 318 f
F 0
31 e
600 S
H 2250 35 2
o B
C E
400
(
l 200 l
F K
80 120 160 200 240 280 320 360 Indicateo Reactor Cociant System Temperature Tc
'F fj, f} urs Aro / /
A' f**'AW f,
e, < ~ d m en 4.
CONSUMERS POWER COMPANY MIDLAND PLANT UNITS 1 & 2 gy
/ dp,.Jik A*4 FINAL SAFETY ANALYSIS REPORT Reactor Coolant System, 3eeese6ee.
16.3/4.4-25 e
Gee 4deauELimitations - Unit 1 Technical Specifications Figure 16.3.4-4A 4/81 Revision 33
2600 2
541. HEATUP SOF HR 00 THE ACCEPTABLE PRES $URE AND TEllPERATURE COMBINAtl0NS ARE BELOW AND HE RIGHT OF THE Lluli CURVE. THE REACTOR uuST NOT BE MADE 2000 - CRITl UNTil THE PRESSURE TErPERATURE COGBINAil0NS ARE TO THE y
RIGHT OF CRITICALITT LIEli CURVE. WARGINS OF 25 PSIG AN
'F l
1800 - ARE INCLUDEO POSSIKE INSTRUNENT ERROR.
3 CRITICALITY 1600 (gggy
[
PolNT PRESS.
TEBP.
1400 A
420 70 3
B 556 13T 1200 C
556 240 a-0 1850 290 H
5 1000 E
2250 284 F
0 2
G 150 84
'G H
1150 290 600 I
22 324 J"
400 200 i
e i
80 120 ISO 20 0 240 288 360 Inaicateo Reactor Coolant System Temperature Tc
'F f
/
4f$
6V f
bL fv6Vc' L
b aa e d s.*~ +.
y CONSUMERS POWER COMPANY MIDLAND PLANT UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Reactor Coolant System, Normal Operations Heatup Limitations Unit 2 16,3/4.4-26 Technical Specifications Figure 16.3.4-2B 4/81 Revision 33
2600 s
NAI COOLDOWN 550F-.-280F 100 F.'HR 400 200 F -.150 F 50F/HR 150F ~ 70F 10F/HR 1
2200 E ACCEPTABLE PRESSURE AND TEMPERATURE COMBINATIONS ARE BELOW en AND THE RIGHT OF THE Llulf CURVE.NARGINS OF 25 PSIG AND
'F 1800
- ARE IN DED FOR POSSIBLE INSTRUNENT ERROR.
3 1600
- _ POINT PRE TEMP.
2 A
180 h 70 l
5 1400 B
460 H2 C
556 1200 56 215 E
1100 224 E
=
F 2250 276 1000 5
800 600
]
C D
400 B
200 A
t t
f I
t f
80 120 160 200 240 280 320 Inoscated Reactor Coolant System Temperature Tc-l IS OPERATING WITH NO RC PUNPS OPERATING.
[f [4 U#S M8 *g/
i 1'
WHEN THE DECAY HEAT RENOVAL SYSTEN (OHR)
T THE IN0lCATED OHR SYSTEN RETURN TENPERATURE TO THE REACTOR VESSEL SHALL BE USED.
g gp [
4 4 g M g4A 2.
A NAllNUM STEP TEMPERATURE CHANGE OF 75'F f[
g IS ALLOWABLE WHEN RENOVING ALL RC PUNPS FROM OPERATION WITH THE DHR SYSTEN OPERATING.
THE STEP TEMPERATURE CHANGE is DEFINED AS THE RC TEMP (PRIOR TO STOPPING ALL RC PUNPS)
NlNUS THE OHR RETURN TEMP (AFTER STOPPING CONSUMERS POWER COMPANY MIDLAND PLANT UNITS I & 2 FINAL SAFETY ANALYSIS REPORT Reactor Coolant System, Normal 16.3/4.4-27 operation cooldown Limitations Unit / d Technical Specifications Figure 16.3.4-3A 4/51 Revision 33
2600 MAX. C00LOOWN 550F -+ 280F 100F/HR 280F + 150F SOF/HR 2400 150 F -.- 70 F 10F/HR j
0 220 2000 THE CEPTABLE PRESSURE AND TEMPERATURE NBINA NS ARE BELOW AND TO THE RIGHT OF THE 1800 g
LIMIT CURV MARGINS OF 25 PSIG AND 10*F ARE b
1600
- IN LUDE0 FOR P IBLE INSTRUMENT ERROR.
/
E 1400 E
1200
- POINT PRESS.
TEMP.
b2 A
180 70
[.,
C 556 215 B
556 121 g
2 0
2250 24 800 k
600 B
_C 400 200 A
I t
t t
l 80 120 160 200 240 280 320 7
Indicated Reactor Coolant System Temperature, Tc-T *F T
un wo'//
yt,;, -Pry p re vo d< d hs I.
WHEN THE OFCAY HEAT REMOVAL SYSTEM (OHR)
IS OPERAilNG Wiin NO RC PUMPS OPERATING.
h dlll4, M df M M4 l
THE IN0lCATED OHR SYSTEM RETURN TEMPERATURE TO THE REAC(OR VESSEL SHALL BE USED.
2.
A MAXIMUM STEP TEMPERATURE CHANGE OF 75'F CONSUMERS POWER COMPANY l
IS ALL0 FABLE WHEN REMOVING ALL RC PUMPS FROM OPERATION WITH THE OHR SYSTEM OPERATING.
MIDLAND PLANT UNITS 1 & 2 THE STEP TEMPERATURE CHANGE IS DEFINED AS THE FINAL SAFETY AN.LLYSIS REPORT RC TEMP (PRIOR TO STOPPING A!L RC PUMPS)
MINUS TiiE DHR RETURN TEMP ( AFTER STOPPING ALL Reactor Coolant System, Normal RC PUMPS).
Operation Cooldown Limitations
(
Unit 2 16.3/4.4-28 Technical Specifications Figure 16.3.4-3B 4/81 Revision 33
~
1
)
MAX. HE AVUP = 50F HR E
I 2400 X. COOLDOWN 550F + 280F = 100F HR
- 7 l
280F 150F = 50F HR 2200 150F - - 70F = 10F HR 2000
/
y THE ACCEPTABL RESSURE AND TEMPERATURE COMBINATIONS ARE BELOW
[
1300 AND TO THE RIGHT THE LIMIT CURVE. MARGINS OF 25 PSIG AND 5
iOF ARE INCLUDED FOR SSIBLE INSTRUMENT ERROR.
5 y
1600 O
a-
\\
1400 POINT PRESS. l TEMP.
I 1200 00 80 b
E B
556 108 t
l E
C 556 236 g
1000 1600 262 0
5 E
2500 309 800 l
600
/
B C
400*
A 200 t
t t
t t
t 80 120 160 200 240 280 320 360 Indicated Reactor Coolant System Temperature Tc
- F
("j f & Y i, W
prav[/4./
4Py a %.c a M ay a. M,
16.3/4.4-29 CONSUMERS POWER COMPANY MIDLAND PLANT UNITS 1 & 2 En see Vib< les k a., d FINAL SAFETY ANALYSIS REPORT Reactor Coolant System, M m d/g 73, c M
[
~
^7-- " - - "- -; Limitations rap 3 nit 1 He 4y as d Ca /k J
Technical Specifications Figure 16.3.4-2A 4/81 Revision 33
7 C
2600 MAX. HEATUP 50F/HR 2400 MAX. COOLDOWN 550F -.- 280F 100F/HR 280F + 150F SOF/HR 2200 150 F -= 70F 10F/HR 2000 1800 5
THE ACCEPTABLE PRE URE AND TEMPERATURE BINATIONS ARE b
1600 BELOW AND TO THE RIGH F THE LIMIT C E.
MARGINS OF 25 E
PSIG AND 10F ARE INCLUDE OR POSS E INSTRUMENT ERROR.
l 1400 E
1200 a
POINT PRESS.
T P m
1000 A
300
/ 80 j
8 556 108 800 C
5 236 0
00 284 600 400 A
200 I
t t
t t
280 \\
320 80 i20 iGO 200 2A0 NM ieo, cot.o R..cto, cooi.ot 5,stes T..o u. tor. Tc. r A,,
S;, ur e
- w. //
b<
p r e *
- d' 0' ay w
a/ w = 4.
16,3/4.4-30 CONSUMERS POWER COMPANY MIDLAND PLANT UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT
- h Renttor Coolant System, Inservico Leak}down Limitations Unit /p &
Iydrostatic Test Heatu Cool E
Technical Specifications Figure 16.3.4-4B 4/81 Revision 33
MIDLAND 1&2-FSAR REACTOR COOLANT SYSTEM TABLE 16.4.4-5 l33 o
REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE'I I
Capsule Vesselt2 LeadGU Withdrawal'8 l
Number Location Factor Time EFPY Removal Interval NSS 12 f.S MDI-A WX-Top 9.6 999@
End of first cycle (10.9')
J.o MDI-D WX-Bottom 9.6 4,9@
End of third cycle (10.9')
Gr. V seve 4A MDI-B XW-Top 6.9
+c&b End of 4664k cycle (26.S*)
4./
MDI-E XW-Bottom 6.9 Whe>9 End of ninth cycle (26.5*.)
MDI-C ZY-Top 6.9 Standby (26.5')
MDI-F ZY-Bottom 6.9 Standby (26.5')
41 NSS 13
/. 3 MDII-A XW, Top 6.9 4v44 End of first cycle (26.5*)
V. 7 MDII-C XW, Bottom 6.9 4s4Mk End of fifth cycle (26.5')
4.1 MDII-B Z Y, Top 6.9
&cek End of ninth cycle (26.5')
MDII-D ZY, Bottom 6.9 Standby (26.5*)
iI The schedule may be modified by amendment, if necessary, after l
evaluation of each capsule.
In addition, the schedule is approxi-mate - the actual withdrawal schedule will depend on scheduled refueling or any unforeseen shutdown occurring near the scheduled withdrawal.
t2iSee M Figure 5.3-6.
_.4 4,'els a c sS (33To 1/4 fflocation (43Based on 460 EFPD Cycle 1 and subsequent cycles of 310 EFPD 16.3/4.4-31 Revision 41 2/82
e MIDLAND 182-FSAR m
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Revision 41 16.3/4.4-32 2/82
MIDLAND 1&2-FSAR REACTOR COOLANT SYSTEM BASES value of the predicted adjusted reference temperature at the end of 5 EFPY.
The reactor vessel materials have been tested to determine their initial RTnoy.
The results of these tests are shown in Table 5.3-2.
Reactor operation and resultant fast neutron (E>l MeV) irradiation will cause an increase in the RTway.
Therefore, an adjusted reference tempern.ture, based upon the fluencef copper, and phosphorus content of the material in question, caA be 39 predicted using BAW-10046A, Regulatory Guide 1.99, and BAW-10144A.
The heatup and cooldown limit curves, Figures 16.3.4-2A&B and 16.3.4-3A&B, respectively, include predicted adjustments for this shift in RT o t at the end of 5 EFPY, as well N
as adji2stments for possible errors in the pressure and temperature sensing instruments.
The actual shift in RTwoy of the vessel material will be established periodically during operation by removing and ovaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specisens installed near the f nside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside 39 curface are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent cection of the reactor vessel.
The heatup and cooldown curves must be recalculated when the ARTwor determined from the surveillance capsule is different from the calculated ART for nog the equivalent capsule radiation exposure.
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MIDLAND 1&2-FSAR REACTOR COOLANT SYSTEM r
BASES 16.3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Table 16.5.7-1.
During heatup 13 3 and cooldown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.
These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady-state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.
The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel.
These stresses are additive to the pressure induced tensile stresses which are already present.
The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.
Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.
The heatup limit curves, Figures 16.3.4-2A&B, are composite 133 curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 100F per hour.
The cooldown limit curves, Figures 16.3.4-3A&B, are composite curves which were prepared 133 based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresser at the outside wall.
The heatup and cooldown curves were prepared based upon the most limiting o
Revision 33 16.3/A-21 4/81
I MIDLAND 182-FSAR
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FIGURE 16.4-1 HAS BEEN DELETED AND THE INFORMATION HAS BEEN PROVIDED IN du b see ha,
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Revision 41 16.3/A-23 2/82 I
MIDIJLND 152-FSAR N
BASES FIGURE 16. 4 2 yAS BEEN DELETED
. Revision 33
- 16. 3/A -24 4jg3
A Rasponses to NRC Questions Midland 1&2
\\
Ouestion 121.17 (3.2)
Compare all tests, da ta, me thods, proposed programs, etc as presented in the Midland Units 1 and 2 PSAR, technical specifications, and any other referenced sources (such as topical reports) on a point-by-point basin with the requirements of Appendix G, Fracture Toughness Requirements, and Appendix H, jg Reactor Vessel Materials Surveillance Program Requirements, of 10 CFR, Part 50.
Identify all areas of noncompliance to the ap pendixes.
Response
The areas of noncompliance are due to the Midland Units 1 and 2 reactor vessel being f abricated to the 1968 edition of the ASME Code,Section III while Appendix G to 10 CFR 50 references the 1972 summer addenda of the Code.
Significant changes occurred to 15 the Code during this pe riod of time.
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..,..m..mo In accordance with Enclorure 1 of S. A. Varga's letter dated March 30, 1979, we are requesting formal exemption from Appendixes G and H of 10 CFR 50 for the f:11:11:3" areas of 32 nonconf o rma nce i d a% +i f i s, d h
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Requirement - " Individuals performing fracture tough s
tes ts...shall have demonstrated competency to per m
tests in accord with written procedures of e
e co nent manuf acturer."
b.
Desc rip n - At the time of the initi tests of Midland P1 t Units 1 and 2 RCPB ma ials, no standard written proce es existed to de strate competency of
. 6 individuals to o rm tests.
c.
Comments - Subsequent ital testing of Midland materials, standard do tation was developed to demons trate materia testi compe tency.
At the time of the inital tes it was in try practice not to document mater testing qualif t tions.
- However, individuals re experienced and co tent to perform the mater tests in question.
ap pe nd i x G,
art IIIC, Paragraph 2 a.
Requirement - "Where seamless shell forgings are use 4
or where the same welding process is used for Y
A Revision 41 Q&R 3.2-5 2/82
R23ponaso to NRC Quastions F
0"i I
Midland 1&2 b
spa J / g/,. A I
longitudinal and circumferential welds 1 ate the st specimens may be taken from a separate weldment p
vided that such a weldment is prepared using exces ma rial from the shell forging (s) or plates (as app cable), the same heat of filler material, and he same roduction welding conditions as those used joini the corresponding shell materials."
b.
Descrip on - Consistent with common industry
- actice, test spe mens were obtained from weldments m de with 41 the same at of filler material and the sa production welding co itions as those used in joinin the correspondi shell materials, but the ma rials joined were not fro excess shell material.
c.
Comments - Dat obtained from test spe mens fabricated as described in
- tem b will provide i ormation that is equivalent to dat which would be ob ained by complying with 10 CFR 50 Ap ndix G, Paragrap III.C.2.
Appendix G, Part IVA, Paragra 3
a.
Requirement
" Material for ping (i.e., pipe, tubes, and fittings), pumps, a va es (excluding bolting materials) shall meet the r uirements of Paragraph G 3100 of the A Code."
b.
Description - The reacto e
lant pressure boundaries of the Midland plant were esigr. d and constructed in accordance with the r uireme s of an addenda or edition of the ASME de, Sect 1 n III, issued before the summer 1972 addenda 16 c.
Comments - This i ue is discusse in detail in BAW l
10046A, Chapter Appendix G, Part IVA, aragraph 4 a.
Requireme s
" Materials for bolting an other fastener with nominal diameters exceeding 1 inch shall meet th minimum (Charpy) requirements of mils later expansion and 45 ft-lbs...."
b.
Des iption - Bolting materials meet the requi ments of th applicable ASME Code edition.
c.
omments - See BAW 10046A, Section 3.2.1.
Revision 41 Q&R 3.2-6 2/82
Rnsponses to 11RC Ounstions Midland 162 Appendix Part IVB a.
Requ ement
" Reactor vessel beltline mater' ls shall have a nimum upper-shelf energy of 75 f bs unless it 16 is demons ated...."
b.
Description -
e of the Unit 1 weld nts has been determined to ha an unirradiated pper shelf somewhat below 75 foot-poun (see Table 5
-2).
l39 c.
See BAW 10046A, Section
.1.3 Appe ndix !!, Part IIC, Paragraph 1 a.
Requirement - "Survei ance speci ns shall be tdken from locations alon ide the fractu toughness test 16 specimens required y section III of pendix G.
The specimen types s 11 comply with the re irements of section IIIA of ppendix G."
b.
Description Because of e scarcity of one w d metal used in th 11SS-12 reactor vessel, WF-70, a si lar (same he of filler wire) weld metal, WF209, i substit ed for this particular material during c tain phase of the surveillance program.
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4 d *6 JP Revision 41 O& R 3. 2-6a 2/82
-f.
MIDLAND 153-FSAR 1
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Revis ion 41 Q&R 3.2-6b 2/82
Responses to NRC Question:
Midland 1&2 c.
Comments - This is discussed in detail in Subsectio 3.1.6.1.3.
A response to the substitution is g en in 39 Qu tion 121.19.
Appendix H, Part C, Paragraph 2 a.
Requirement
" Surveillance capsules co aining the surveillance s cimens shall be locate
...so that the neutron flux rec ved by the specime is...not more than three times a high as that re ived by the vessel inner surface."
b.
Description - As indica d in gure 5.3-6, the lead l 39 factors to the vessel inn s
face exceed three.
c.
Comments - This requireme
'll be changed with a new revision to 10 CFR 50, A endi H.
The Midland lead factor should then be ceptabl Appendix H, Part IIC, Parag ph 3 16 a.
Requirement e required number of rveillance capsules and eir withdrawal schedules re as follows:....
C b.
Descripti
- Although the number of capsule used in NSSS 12 nd 13 programs meet or exceed the min' um requir
.ent, the withdrawal schedules for these apsules spec led in Table 16.4.4-5 are somewhat differen 1 39 tho specified in Appendix H.
16 c.
omments - The philosophy behind tb^
schedule has been acceptad t i d RC. ) withdrawal
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Revision 39 Q&R 3.2-7 11/81
Recponses to NRC Quantions Midland 1&2
(.
Question 121.11 (5.2.3.1, 5.3.1)
(
Clarify the discrepancies between Table 5.2-3 and Table 5.3-2, with specific reference to materials of construction.
Also clarify the discrepancies between Table 5.3-1 and Table 5.3-2, with specific reference to USE and'RTNOT of weld deposits.
8 Re sponse The indic ed corrections have been made to Tables 5.2-3j fr. 5-2';
and 5.3-Table 5.3-1 has been deleted, and pertinent 39 informat n has been incorporated into Table 5.3-2.
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Revision 39 Q&R 5.2-3 11/81
l l
l R0cpan300 to NRC Qusation3 Midland 1&2 Question 121.12 (5.2.3.3, 5.3.1, 3A.l.99, 5.3.2)
)
R3ferences to Topical Reports BAW-10046P and BAW-10046 are not cppropriate.
BAW-10046A, for calculation of pre-and post-irradiation properties of reactor vessel materials (using the NRC staff recommendations attached therein), is an acceptable reference.
Amend Section 3A.1.99 to reflect the staff rccommandations.
l Rcsponse Topical Report BAW-10046A, Rev.
1, presents B&W's method of complying with the fracture toughness requirements of Appendix G to 10 CFR 50.
Although the topical report does contain design curves for predicting the irradiation effects on the properties of the reactor vessel materials, the more conservative design curves of Regulatory Guide 1.99 are used to predict the adjustment to the referenced temperature.
Rasponse to Regulatory Guide l'.99 has been revised in Appendix 3A in response to this question-
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l Q&R 5.2-4 Revision 8 4/78 l
Responses to NRC Questions Midland 1&2
(
Ouestion 121.10 (5.3.1)
In reference to Request 121.5, we require the following additional information on Midland plant, Units 1 and 2 reactor vessels:
1.
A schematic sketch of each reactor vessel showina all welds (lonoitudinal and circumferential), plates and/or forgings in the beltline region.
Welds should be identified by a shop control number (such as a procedure qualification number), the heat of filler metal, type and batch of flux, etc.
Each plate and forging should be identified by a heat number and material type.
2.
For each of the above welds, and for welds in the vessel 8
material surveillance program, an identification of the welding process should be provided.
3.
A listing of the following information on all beltline materials (weld, plate and/or forging):
chemical composi tion (particularly Cu, P,
and S content), drop weight T RT
, USE and tensile properties.
(If any of these fracture toughness requirements have not been determined, use Branch Technical Position - MTEB 5-2 to estimate their value.)
(
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4.
The maximum end of life fluence at the vessel I.D.
a nd 1/4t locations for each weld in the beltline region.
Response
O&R Figures 5.3-1 and 5.3-2 are schematic sketches of the Midland Units 1 and 2 reactor vessels showing the beltline region welds and forgings.
The welds are identified by their weld qual if icat ion test number and the base metals by their forging 39 identification numbers.
Table 35.3-2 rovidedbthe specific heats, weld flux, chemistry, mechanical, and impact toughness data for the beltline region and surveillance materials, a d 7.5-2 A Revision 39 O&R 5.3-3 11/81 1
R2cponaca to NRC Qusations Midland 1&2 Question 121.6 (5.3.2, 16.0)
Provide pressure-temperature limits, as required by General Design Criterion 31 to assure adequate safety margins against non-ductile behavior or rapidly propagating failure of ferritic caterials of the reactor coolant pressure boundary, for each of the following operating or test conditions:
(1) Preservice hydrostatic tests, (2) Inservice leak and hydrostatic tests, 3
(3) Heatup and cooldown operations, and (4) Core operation.
Rnsponse The requested pressure-temperature limits for preservice hydrostatic tests are provided in Figure 5.3-1.
l 39 Pressure-temperature limits for inservice leak and hydrostatic 3
tests, for heatup and cooldown operations and core criticality, and for core operation are provided in Subsection 16.3/4.4.9.
] 39
)
Revision 39 Q&R 5.3-4 11/81
Respon0cc to NRC Queotions Midland 1&2
('
Question 121.20 (5.3.1) (RSP)
Babcock and Wilcox has informed us (IE Preliminary Notification of Event of Unusual Occurrence PNO-78-141A) that certain B&W supplied reactor vessels may have been fabricated with weld wire that was not within specification.
This weld wire is characterized by an atypical composition having high silicon and low nickel.
B&W has stated that a portion of weld filler wire heat Number 72105 was misidentified by the weld wire manufacturer.
This weld wire was used to weld beltline materials ABZ-196 and ACA-197 ( weld qualification number WF-70) in the Midland Plant Unit I reactor vessel.
B&W has determined that the ductile-to-brittle transition temperature of the faulty weld wire 14 may be 50 to 100F higher than that measured on a properly fabricated WF-70 weld.
It is our position that Consumers Power Company verify the presence or absence of this weld wire in the Midland Plant Unit I reactor vessel.
If it is found that this weld wire was used to fabricate the reactor vessel, then you must reevaluate the technical specification pressure-temperature limit curves to show conformance with the requirements of 10 CFR 50, Appendix G, Fracture Toughness Requirements.
Re s ponse
,a. El f* I' Subsections 5.2.3.3.1, 5.3.1.5, and 5.3.2; Tables 5.3-3 and Figures 16. 3.4-2 A, 16.3.4-3A, and 16.3.4-4A have been bevised in 39 response to a presumed presence of the " atypical" weldment.
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l Revision 39 Q&R 5.3-9 11/81
ROOp n3GO to NRC Quacticn3 Midland 1&2 Qu stion 211.182 (5.3) (RSP)
^g Your response to request 211.103 does not meet our requirements with respect to check valve leak testing.
The proposal to test two valves in each of the core flood and low pressure injection lines is acceptable for these systems.
However, we require that et least two check valves in each of the high pressure injection lines be tested also.
This should be done by classifying these I 17 s
valves as AC in accordance with Section XI of the ASME code.
s Modify your response accordingly.
I l
Rorponse The responses to Questions 211.22 and 211.103 have been revised in response to this question.
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ev sion 17 Q&R 5.3-10 1/79
i,,
ROCpana;3 to NRC QuOstion2 Midicnd 1&2 l
Question 121.22 (16.0)
!1&
6 F gure 4-1, " Fast Neutron Fluence (E > 1 MeV! as a Function of 2,';
Full Power Service Life," Figure 4-2, "Effect of Fluence and Copper on Shift of RTum for 'leactor Vessel Steels Exposed to i 26
'550sF Temperature," and Tnble 4-1, "Reacto; Vessel Toughness," of i
the Midland Plant Technical Specifications have been left blank.
l Su; ply this information.
21 Respons_e_
1
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Figure 16. 4-1 has been ' deleted from the FSAR. The requested t
i'i information is provided in gubsection 16. 3/4. 4.pi ;;f r--
j te " ' " L, _ a ~,. _ _ 2?? i.: ::J.2 Figure 16. 4-2 nas been
' deleted and subsection 16.3/4.4.9 has been revised to include 39 i
1 the - information from Figure 16.4-2 by a reference to Regulatory GuMe 1. 99. Table 16.4-1 h'a included in Tables 5. 3-2 h.s been deleted and the information is
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Q&R 16.0-1 Revision 39 11/81
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Enclemure 2 I
l RESPONSE TO DRAFT SER OPEN ITEMS 123.1 and 2 123.1 To demonst:4te compliance with the beltline material test requirements of Faragraph III.C.2 of Appendix G, 10 CFR Part 50, indicate whether the beltline test materials comply with the requirements of Paragraph III,C.2 of Appen-dix G, 10 CFR Part 50.
If the applicant cannot comply with the requirements of Paragraph III.C.2, state why and provide data to demonstrate that the weld metal test materials are metallurgically equivalent to the beltline welds which they represent. Parameters to be compared from the weld metal test and the actual beltline welds are welding parameters (eg, wire type, weld process, flux type, etc), base material specificacion, weld thickness and postweld heat treatment.
Response
Subsection 5.3.3.9 addresses the degree of compliance of the Midland Plant with Appendices G and H of 10 CFR 50.
Paragraph III.C.2 of Appendix G is addressed specifically in that subsection.
(
3S$42 a)
Indicate the estimated irradiation damage which WF-70 weld can sustain and have its upper shelf energy remain above 50 ft-lbs.
b) To demonstrate that Beltline Weld WF-70 of Midland Unit I complies with Paragraph IV.B of Appendix G, 10 CFR Part 50, provide CVN impact data which demonstrates that the estimates of 123.2a) are conservative.
CVN impact test data to be provided must be from weld test specimens removed from surveillance capsules. The weld test specimens must be from materials which were prepared using the equivalent welding parameters (weld wire, flux type, chemical composition and weld process as Weld WF-70 and had an unirradiated CVN upper shelf energy similar to WF-70.
The applicant must provide actual CVR upper shelf energy data from the unirradiated WF-70 and unirradiated and irradjated weld test specimens. The l
applicant must provide the chemical composition and wire and flux heat j
identification of the weld test specimens.
l
Response
a) Appendix 3A and Table 5.3-2 have been revised to address the prddicted irradiation damage which WF-70 weld can sustain and maintain its upper shelf energy above 50 ft-lbs.
b) Appendix 3A contains information concerning the evaluation of VF-70 several years prior to its predicted drop to 50 ft-lbs.
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