ML20234B742
| ML20234B742 | |
| Person / Time | |
|---|---|
| Site: | 05000000, North Anna |
| Issue date: | 07/02/1976 |
| From: | Pike W Office of Nuclear Reactor Regulation |
| To: | Muller R Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML20234A777 | List:
|
| References | |
| FOIA-87-40 NUDOCS 8707060184 | |
| Download: ML20234B742 (84) | |
Text
{{#Wiki_filter:, - _ - - . j :' s jf# " %, UNITED STATES 2+4 NUCLEAR REGULATORY COMMISSION y j j WASHINGTON, D. C. 20555 e \\,...../ JUL 2 2 Docket Nos. 50-338 and 50-339 ) Ragnwald Muller, ACRS Staff J Enclosed is a list of the NRC staff members who will l be available for the July 7,1976, Subcommittee meeting concerning the North Anna Power Statien Units 1 and 2 application for operating licenses. Walter J. Pike, Project Manager Light Water Reactors Branch No. 3 I Division of Project Management l l O B707060184 870610 PDR FOIA THOMASB7-40 PDR
se JUL 2 TJ/S ENCLOSURE NRC STAFF AVAILABLE FOR THE JULY 7,1976 ACRS SUBCOMMITTEE MEETING ^ NORTH ANNA 4 Z. Rosztoczy J. Knight -~ S. Pawlicki I. Sihweil _ ( C. Stepp K. Desai F. Litton S. Bhatt U. Potapovs S. Chan D. Jeng S. Isreal S. Kim P.. Meyer l M. Dunenfeld. D. Thatcher D. Mcdonald B. Ma nn F. Allenspach M. Fields J. Kudrick J. Boegli C. Ferrill L. Bell K. Campe L. Heller W. Bivins A. Cardone J. Fairobent T. Murphy F. Liederbach J. Martin J. McMillen W. Jensen T. Johnson R. Hofman R. Bosnak A. Dromerick W. Pike D. Vassallo wvam ya t e w. ..i u +
diJARD'S DESK ~ ADVISORY COMMITTEE OM REACTOR SAFEGUARDS J MEETING OF THE NORTH AMMA SUBCO:t1ITTEE ') ROOM 1046 1717 H ST. NW' ~ WASHINGTON, D.C. ' JULY 7, 1976 NAME AFFILIATION David Okrent Aros %~har ACRS Me.mber H. Etherington M _ "9. Plesset Ac9s Member ~
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H. S. Isbin 7cF9 Ma~ Me*~ s 'M _ Bush ACRS Member J. Merkle ACRS Consultant Ivan Catton ACRS Consultant l T. Theofanous ACR'S Consultant W. C. Lir;inski ACRS Consultant L T. Wilson .ACRS Consultant D. Canonico ACRS Consultant e w o,
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1 / lt. ARD'S DESK t 4 ADVISORY, COMMITTEE ON REACTOR SAFEGUARDS l MEETING OF THE 8 'k NORTH ANNA SUBCOMMITTEE ROOM 1046 - 1717 H ST. NW WASHINGTON, D.C. JULY 7, 1976 1 = NAME AFFILIATION j ) ~- yEPCO. yxa v wra.n - u ._Mr.L W. F. Bennett .Vircinia Electric and Power Comca:- '~ ._.Mr.. Shm Brown 1 l - u -r. ' v,- nc e 4 + s- ~ l l -=.+a..- i y A nnnm j - w _. r' Annne pr 4 ..n R _ Ry 1\\ri = i l 1 -D. W. Apni clel 1 ..J. M. Davie; l .J. M. McAvov .F. M. Alligood F. C. Prince W. B.'Rodell b r W. L. Parker W. R. Runner I J. L. Perkins C. M. Robinson I. Kaplan N
GuAnbis cesn ADVISORY COMMITTEE OU RE7tCTOR SAFEGUARDS MEETING OF THE h0RTH ANNA SUBCOMMITTEE ,RGOM 1046 - 1717 H ST. NW WASHINGTON, D.C. JULY 7, 1976 NAME_ AFFILIATION ISun Shipbuilding & Dry Dock Co.' Dr. John Harrison B r i f i~s h W e l d i n i I n s t i t u t e ^ Mr. Willian Pell'i'n'i Sun Shiobuildinc c Drv Doch Co. Mr. Richar: bicic;hi Sun Shipbuilding & Drv Dock Co. Mr. Richard Hagan Sun Shipbuilding &'Ory Dock Co. Mr. John Runzer Sun Shipbuilding & Dry Dock Co. Mr. ]eter Hepp Sun Shipbuilding & Dry Dock Co. t I l i )
yur,ypo uuwss e? 8 - VEPCO P-2 NORTH ANNA SUBCOMMITTEE MTG ' i. ' ,.y-.g AFFILIATION 'NAME 3; W. H. Chamberlain ' STONE'& WEBSTER 4
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v! L /. i .r'.- ,_.:w. main l f fire pump bldg. .screenwe lls { I i .w.a t e r i hydrant (typ) 6) bldg. 12" 6 N C ~ R l_ ] D 0 O O O O r Y .c=, c=" l 10" hese rack (typ) b turbine area 0' turbine area _ unit 2 unit 1 IC* c. 4" r =. c) CP t> a u?' ~ warehouse fD[Ijr service building 9 C=3 7 i m = w s au. buildir g aux fw ph reactor reactor f containment containment i /~ fuel J unit i O blde b unit 2 9 12" Y 9 I N f u el oilh { stg tank 12" j { i ~- ~ ~ ' ~ _ ~ ~. ~ ser v. w ate r pump house unsts 1A2
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t $t i' f NAl-ACRS 1.4-1 7-7-76 l STATEMENT OF THE ISSUE: What is the Ability of the Station to Withstand Temporary Loss of All A.C. Power? I VEPCO POSITION: North Anna Iower Station has the ability to withstand a l temporary loss of all A.C. power. Before addressing this subject, houcver, it should be emphasized that off-site and on-site power sources have been designed and constructed to be extremely reliable. For example, availab111ty of of f-site power is insured by (3) redundant reserve station service transformers which provide a very strong tie with the Vepco electrical grid. In addition, availability of on-site power is insurad by (4) redundant diesel generators which provide an uninterruptable power source in the event of an arec wide system " blackout". Not withstanding the fact that both sources of power meet the highest standards, it could be hypothesized that all A. C. power is lost. Even dth this everly stringent essumption; the results of such an occurence are controllable. It is a design feature that systems and components necessary to protect the reactor, in the event of loss of power, depend on energy sources other than A.C. power for the actuation. Consider the following results if all A.C. power is lost, while the reactor is at 100% power: 1. Saf e Shutdown of the reactor is assured even without A.C. power, because the reactor protection system does not depend on A.C. power; but utilizes D.C. power from the redundant 125v d-c Batteries for actuation. Therefore, the reactor will be shutdevn and auxiliary feedwater flow fnitiated. 2. Sufficient Removal of Store: and Residual Heat is assured, even without A.C. power, because the mcchtnically operated code safety valves on the main stream line utilize spring force for actuation. Therefore, excess energy will be removed from the system. 3. Adecuate Addition of heat sink is assured, even with no A.C. power becau'se the turbine driven auxiliarv feedwater pump does not depend on electrical pcwer, but utilizes stored steam for actuation. Therefore, enough water will be added to remove core residual heat.
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NAl-ACRS 1.4-2 7-7-76 In conclusion, the station has the ability to temporarily withstand III. loss of all A.C. power. During a period of time, of at least several hours following the event, the system has the ability to safely maintain hot Sufficient time exists to restore normal A.C. powered cooling shutdown. systems. 8 s l i I l s l { I
( .'.I - 3 [, e 7 5 l.., s ; ; -r INTRODUCTORY REMARKS BY SAM C. BROWN, JR. VICE-PRESIDENT, POWER STATION ENGINEERING AND CONSTRUCTION DEPARTMENT, VEPCO BEFORE THE NORTH ANNA SUBComi1TTEE I ACRS, JUIY 7, 1976 Gentlemen: My name is Sam C. Brown, Jr., Vice-President, Power Station Engineering and_ Construction, Vepco. Until about a week ago, we were prepared to appear here today and report to you that there is reasonable assur-ance that the steam generator supports at North Anna can be used safely without any modification in either the structures or in the conditions under which they will operate. Based on data obtained within the past week, however, we have concluded that the toughness characteristics of some of the steel in the structures would not be adequate for use in the containment environment in which we had planned to use them. This af ternoon we will describe for you both our earlier data, which we found so reassuring, and the more recent, disappointing results. We will also describe our proposed plan for dealing with this toughness problem. That plan consists of insulating the supports in order to maintain them at temperatures where their toughness will be adequate. Before our presentation begins, however, I want to address a question that may well have occurred to you. How have we come so far down the road without having detected any problems with the material properties of our steam generator supports? It is importo.nt to realize that when these structures were designed, they were designed in accordance with all applicable codes. Those codes imposed no fracture toughness requirements for structures such as these. The steels that were specified for the North Anna steam generator supports were steels commonly used throughout the nuclear industry. As our litigation with Sun Ship Building and Dry Dock Company progressed, se set out to test any of the steels in the structures that were available to us. There are two types of steels used in these structures - - A-36 and A-572. We found tha't we had on hand at the. North Anna site certain A-36 material that was originally included in the steam generator supports before the rewelding was undertaken. Portions of this A-36 steel had been removed during the repairs, and
a w t .Q i . we were able to perform our tests on these removed portions. You will see the results this 5f ternoon.Ve were quite pleased with them. Indeed we will demonstrate that insofar as those A-36 steels were concerned wa came extremely close to satisfying the presently applicable Section NF even though the supports had been designed before that Section had ever been proposed. As far as we knew at that time, we did not have any A-572 steel on hand for testing. But based on material in the literature, we believed it was reasonable to conclude that the A-572 steel would have toughness characteristics equal or superior to those of A-36. Accordingly, we took the position in our earlier presentations to the NRC that our supports would perform adequately under the conditions planned to exist at North Anna. Within the last two weeks, however, we discovered that in fact we did have some A-572 material that could be tested. 1 We found some A-572 beams that had not been needed in repairing the steam generator supports but that were from the same heats of material cs some of those that had been used in the repairs. We immediately arranged to have this A-572 steel tested for toughness. The results showed that, contrary to our belief, the toughness of the A-572 steel was decidedly inferior to the A-36. About the same cine that we found the A-572 material, we discovered in deposing a Sun Ship employee that Sun Ship might have in its possession some A-572 material from our supports. We called them and agreed to exchange our A-572 test results. While we were satisfied that the supports will perform as designed under normal conditiens, the conservatism that is desirable under the unlikely accident condition requires that we make a design modification. Based on collective results, we were informed by our consultants that steps should be taken to increase the operating temperatures in which the steam generator supports would be used. We reported the results of our tests to the NRC immediately and, as I have said, we will discuss them thoroughly with you this af terncon. I have discussed this background with you because I want there to be no question about Vepco's good faith or its sincere desire to see that these supports have been built and will be used in a way that will not endanger the public health and safety.
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I v.. ? u Jiu . n... LOWER STEAM GENERATOR SUPPORT FR AME M ATERI AL S NORTH ANNA 1 2 MINIMUM SPECIFICATION CERTIFIED MIL MATER 1 REQUIREMENTS TEST REPORTS PROPERTY A-36 A-572 A-36 A-572 YlELD POINT 36.0 42.0 37.3-54.8 56.5-33.O i TENSILE STRESS 58-80 60.0 65.3-80.0 83.5-91.5 ELONGATION 17.0 17 0 21,5-33.0 24.5-25.0 i 45.3-60.9 58.1-62.6 REDUCTION OF AREA SPECIFICATION: ASTM A-36 & A-572 GR 42 ADDED REQUIREMENTS: UT PLATES & SHAPES t 3" ASME III. SUBSECTION NF 1 MINIMUM SPECIFICATION M ATERI AL REQUIREMENTS PROPERTY A - 36 A - 572 YlELD POINT 36.0 42.0 TENSILE STRESS 58 - 80 60.0 ELONGATION 17.0 17.0 l
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._...-( 1 at i Repair, inspection and Quality Assurance Steam Generator and Reactor Coolant Pumo Repair Program _ Repair of Steam Generator and Reactor Coolant Pump Supports ), The removal and replacement of all welds in these support structures We say proven, because hi s was accomplished using proven procedures and tec n que. h which are widely these were not unique-or experimental procedures but t ose d use to produce used in the industry and have been demonstrated by repeate the desired benefits. Exhibit I summarized these along with the resulting benefits that were achieved by their use: 9 l l
2 t l Pump Supports Inspection of Steam Generator and Reactor Coo ant 2. decided in developing the repair procedure, it was At the outset, icle method (MT) as the primary to rely upon examination by the magnetic partAgain, this no manns of assuring weid quali ty. industry; its capabilities are is an accepted method and is widely used in the l to provide assurance wall known. The introduction of ultrasonic test techniques was llar tearing were not present and to that base metal defects such as lame in preventing such defects. i confirm that the procedures used were effect ve i s Exhibit #2 lilustrates the extent of such inspect on. Insoection Results_ i d 3 separate The completed welds in the Unit I supports rece ve d in addition to the "in-process" examinations by the magnetic particle methoConsidering that the m examinations carried out during welding. re considered very good with defects l are welding process was ussd, the resu ts we f weld Inspected. The being found in less than 1% of the total lengths otine; that is, slag nature of the defects was also considered rouThere were no gross undercuts, and some linear indications. defects to cause concern. d two separate The completed welds in the Unit 2 supports receive "in-process" magnetic particle method in additica to the examinations by the f final post These examinations were conducted before and a examination. tment on Uni t #2 weld heat trnetment. The number of defects found af ter post weld heat trea 1, with about 13% of the final MT on Unit was greater than that experienced at about 4% with a repair depth greater than l welds requiring some repairs but on y removal of welds. ed the In some cases the pursuit of these indications caus 1".
~ 3 i However,_It is not unusual on large welded structures such as these to find defects such as slag stringers, slag pockets and other small defects which had been aggravated into cracks or tears during the post weld heat treatment The important point if, that all of these were or stress re11sving process. found and repaired satisfactorily. A special re-examination of one structure was made to verify our contention that the HT examination results reported by Sun Ship were not valid and represented non-relevant Indications. The inspections and examinations by the magnetic particle method perfonned during the repair give adequate assurance that the welds in these structures are sound and will perform their intended function. The ultrasonic examination of certain welds to verify absence of Although no evidence of lamellar lamellar tearing adds to this assurance. tearing was found, some of the UT reflectors turned out to be weld defects. All This was not unexpected because of the sensitivity of the UT procedure. i In some cases additional reportable reflectors were reviewed and evaluated. UT was required to more accurately assess' the significance of the reflectors. These On three welds the reflectors appeared to have planar characteristics. l were removed, weld repaired and re-examined and determined to be acceptab e. Therefore, the inspections and non-destructive tests performed during the repair cycle provide assurance that the required weld quality is present and the welds comply with the technical requirements. ) 3 Quality Assurance The following positive quality control actions were implemented durIng the repalr program: A detailed plan which specified all the technical require-1. This ments needed to make the repairs was developed. technology available. speelfication incorporated the best fL----- _ _
4 The repair techniques spectified were formulated by drawing upon the expertise available within ~ the organization of our Architect-Engineers, in addition to the results of consultation with other l f abricators of large structures. The use of a controlled fabrication process whereby a 2. weld " traveller" system was implemented. In this manner, work methods were prescribed in wri ting, outlining each step in the repair procedure for each weld. Such items as weld techniques, weld bead sequencing, hold points for in process NDT, pre-heat and post-heat requirements were included. Operations were not left to the discretion of the worlenan, but were pre-planned by welding engineers and all such actions verified during in-process inspections. Extensive in-process inspections to verify conformance to 3 requirements as the work progressed. I t has been estimated that over 150,000 separate inspections were performed during the repair effort. 4. The implementation of a three level quality assurance i surveillance plan during the repair effort to verify adequate / work performance at all levels. The first level was the quality control surveillance Imposed by the organization responsible for performing the repair work. They were responsible for the day to day 1 I J
5: Inspections, documentation and work verification. Also, these organizations had full quality assurance organizations with programs complying with Appendix B, 10CFR50. The second level was a planned audit program carried out by the Virginia Electric and Power Company quality These audits were conducted Assurance Department Engineers. Seventy-on a regular basis throughout the repair activity. three such audits were conducted and the results confirmed ~ that acceptable quality programs were being implemented. The third level was the independent overview provided by They our consultants, Southwest Research Institute. participated in the program starting with the review and comment on the repalr specification and continuing on with audits of work performance and witnessing of non-destructive I tests. The steam generator and reactor coolant pump support welds were replaced using carefully thought out procedures and a closely controlled quality program. The quality assurance program implemented during the. repair of these structures f All the evidence complies fully with the requirements of Appendix S,10CFR50. supports our view that the welds are sound and are in canpliance with the I technical requirements. JLP 7-1-76
\\ EXHIBIT 1 PROCEDURE BENEFIT AWSandASMEqualffia.dwelding..................VideIndustryusagehas procedures demonstrated that the use of such procedures will produce sound welds Stringent preheat and pos t hea t.................. Minimizes hydrogen induced cracking, control reduces localized restraint stresses, retards cooling rate in weld metal and heat affected base metal which improves metallurgical structure " Buttering" weld passes applied................. Provides a buffer of ductile weld to base material before welding metal between base metal and cavities cavity weld Welding technique sheets specified............... Reduces weld shrinkage s tresses l weld bead sequence Veld joint sequence speci fied................... Provides optimum access, reduces i ~ overall welding stresses and l assists in dimensional control ] l Machanical peening of all weld.................. Distributes and reduces residual ) j layers except buttering, root welding stresses cnd cover passes (Unit #1 only) l Pos t weld heat trea tment of completed............ Reduces res idua l welding s tresses, assemblies (Unit #2 only) toughens weld heat-affected zones Use of low hydrogen type coated welding......... Minimizes hydrogen induced cracking, i electrodes (E-7018) all position type electrode in common usage with well documented properties Strict control of electrode issuance............ Minimizes hydrogen induced cracking and maintenance of heaters to keep electrodes free from moisture July 1, 1976
i ..p ~ J EXHIBIT 2 INSPECTIONS CARRIED OUT DURING REPAIRS f VISUAL INSPECTION (VT) Surfaces and edges of excavations were verified to be smooth, (A) uniform, f ree f rom fins, tears, cracks and other defects. (B) Each pass of weld metal was visually examined. (C) All canpleted welds were examined MAGNETIC PARTICLE (MT) (A) Excavations examined by MT before welding. Initial weld layer to base metal and each subsequent t" (B) of weld thickness or h, t, or 3/4 or weld thickness whichever was more restrictive. (C) After canpletion of welding Unit af ter 8 hour preheat temp soak -y> y 72 hours after the 8 hour soak Unit after completion of welding 3 2 24 hours after PWHT ULTRASONIC (UT) Ultrasonic examination of selected high stressed main 1 (A) member welds. t July 1,1976 --__N___--__
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~ + [ j EVALUATION OF MATERIALS-j IN THE. STRUCTURES 1. DATA FROM MATERIAL IN THE STRUCTURES A. CHARPY V-NOTCH TEST DATA, A36 MATERIAL B. DROP WElGHT DATA, A36 MATERIAL C. CHARPY V-NOTCH TEST DATA, A572 MATERIAL D. DROP WEIGHT DATA, A572 MATERIAL E. CHEMISTRY DATA FOR SAMPLES MECHANICALLY TESTED (CHARPY AND DROP WEIGHT TESTED) 'F. CHEMISTRY DATA FOR MATERIAL IN THE STRUCTURES NOT MECHANICALLY TESTED G. TEST DATA OBTAINED BY SUN SHIP, KNOWN ]rl j VEPC0 II. RELATED SUBJECTS A. BRIEF OVERVIEW OF MATERIALS IN THE STRUCTURE AND PROCUREMENT REQUIREMENTS B. RELEVANCE OF TEST SAMPLES, A36 MATERIAL C. RELEVANCE OF TEST SAMPLES, A572 MATERIAL D. NDTT DATA FROM SPECIAL ASME TASK GROUP ON FRACTURE TOUGHNESS, A36 MATERIAL E. RELEVANCE OF VEPCO DATA BASE, A36 MATERIAL F. CONCLUSIONS, A572 MATERIAL b O 2 _
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l CONCLUSIONS 1572 MATERIAL f 1. TE MAXIMUM NDTT AS DETERMINED BY DROF EIGHT TESTS IS 100 F. 2. NOTCH TOUGEESS VALUES AE LOER THAN FOR A36, BUT NOTCH TOUGE ESS IUCREASES TO VALUES OVER 20 FT. LES. AT APPROXIMATELY 200 F. FOR THE MATERIALS SHOUING TE LOEST NOTCH TOUGENESS'. 3 DUE TO THIS NE"I DATA THE OPERATING TEIGERATURE OF THE A572 EMBERS IN TEE STRUCTUES MUST BE l INCREASED. l e h i e l \\ 1
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D E R E T D I D N S E N E T O T C L U T S S I A N T T I N N N M E E T U T P E T T H U A k T 4 T B E E S M E E R I E EN D T F F L P O O R L R MO T O H O I S E T H O E L FI U T T T F T 'O R A D A R S O O E E E E P W R E R R P H H P E Z O E D T T D T M N T N M SM N I E M U E N A E U Z H H I A B G G R G B G G S U O I N R T O E U U E Y S D F A M T C N M A O O G L S S A E E A O I R R R G E R G E G G C H H N E D AA F E T T E E E T T I E L S E MS L L R S N N R R T T NE R T A A T R R E N N A O AD E A E A E E R E E V S T V V A I I T R S T E S T I I G O O B B C T O R C R L N R E S S S M M E T S O R O L K A V O S E S S A P U T O O N A A WP N C O P M O HS R U E O P O R R S O O I S O U R C C S L S H C C T T I E T T S ) ) A L V ) ) ) ) ) ) ) E a b A a b c d e a b E A E IT C E H R AA H H U F I T T D R E O A AU V W I N R S F T 1 2 3 4
E R E P M U R R B O O R T T G G A AT E E U R RR L L T E EO A N NP R R M T R E EP E E G GU G V V P O A S E O O M O M M L S S E L R T A AC S S T F E E T O O E N T TT O R R H H S SE H C C G G P E E U U M MF M M M O O O O O O O R R T M R R R R R H H F F F F F T T T E A R R R R H H H H T F / / / / U U U 'U M T T T T 5 DHO B B B B l 0 0 0 0 0 0 0 0 0 NPE = 0 2 6 4 2 4 0 4 4 t S n 3 5 8 3 8 8 0 8 8 AMC i 8 7 5 3 6 7 3 1 e A b 3 P_ 3 1 m E Ta WT = = I E O F R R R R H H H H LF O / / / / F U U U U T T T T 0 B B B B I 7 T = 0 0 0 0 0 0 0 0 0 E 5 0 6 2 7 0 0 0 0 A t S n 7 8 9 6 8 0 9 1 0 9 A i 8 e 5 9 9 9 1 1 E C b 2 3 2 3 m H T a = = = = = = = = = 3 4 3 2 E Q O Z i 4 9 9 q 9 $aE gsI $$$gE u
1 1 MAXIMUM ALLOWABLE TEMPERATURE NORTH ANNA 1&2 ST. GEN. SUPPORTS 1 \\ 50 - l 0.9 ACTUAL MIN Y.S. A572 GR. 42 W14X605 GEAM 45-, E7) 40 -\\ 0.9 ACTUAL MIN Y.S. x A572 GR. 50 N W14X426 BEAM e 3 E % ' d.E ALLOWABLE N !n w 35 - y% U:! y ^0.9 ACTUAL MIN Y.S. ^ ~ 31.3 K /__ BEAM 42 _ _ _g_ _ _f,_GR.50 C E 30 -% ~30,4 KSl 'N '~ ~~ ~~ ~~ BEAM 23 (A-36) l f g. 8_ K S_,l _ _ _ l __ _ _ y GR. 421 BEAM 88 l l 25 - l I l 455' 575' ,j 20 i i 100 200 300 400 500 600 700 TEM PER ATURE 'F k
MAXIMUM ALLOWABLE TEMPERATU RE NORTH ANNA 1&2 ST. GEN. SUPPORTS 50 - 45-ui 40 - x 0.9 ACTU AL MIN Y.S. A 3" PL. zy 35 - M r h!. ksi 3 F BEAM (PL.) 213 3g _ I I O.9 ACTUAL MIN Y.S. A 1" PL. l I 25 - __.E' 23. 4 K S I l BE AM PL. 21 lp595* I, 20 100 200 300 400 000 soo 700 TEMPERATURE 'F
i MISCELLAE003 ITEMS S.G. SUPPORTS ITFN MAT'L MIN. SPECIF. MIN. ALI4W. DESIGN ALLCW. Y.S. (o.9xY.S.) STRESS TEMP (KSI) (KSI) (KSI) (oF) k 40 155 139 5 105 > 6000 3 S.G. FEET BOLTS (UNBRAKO) 11/2"-12I9" UPPER PEDEST. Ah90 130 117 17.2 > 600* BOLTS 2"-8X10" 0 VERT. SUPP. A516 38 3k.2 27.5 520 BLOCKS GR. 70 VERT. SUPP. 43ho lho 126 108 4800 PLATE 0 SEAR BIOCKS A-36 46.8(ACWAL) 42.1 28.6 540 LUBEITE PL. ASm 55 49 5 13.o > 6000 B22-863 MAG. BRZ. 800 maaIIs
- ?aoPRICmY MATERIAL SPECIFIED SNUBBER 43ho 90 81 50 3
> 6000 CLEVIS
L 1 1 1 HOT 7t!NCTIONAL TEST - STEAM GENERATOR SUPPORT INSUI.ATION A. STEADY-STATE CONDITION 1. Confirm thermal analysis Thermal profile selected beams 2. Modify insulation coverage (if required) B. TRANSIENT CONDITION 1. Record tecrature time-historv selected beams (heat-up and cool-down sequence) i f I
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