ML19319D577
| ML19319D577 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/15/1971 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8003170746 | |
| Download: ML19319D577 (15) | |
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,,A M *k UNITED STATES
//31)is ATOMIC ENERGY COMMISSION v bLs r?,.
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yn s 3 Peter A. Merris, Director, Division of Ecactor Licensing CRYSTAL RI\\T,R liUCLEAR GE::ERATI::G PLA:IT, U:'IT 3, DOCRET :;o. 50-302 The inforration cubnitted 'cy the applicant in tha TFA't has beca revie'.ied by the.:aterials Enr ineerin', 73rnnch, Ins, adequate responses to the enclosed request for additional information are rcquired befora ze enn corplate our evcluatien.
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t Edcen c. Crea, I; ire c to r Division of Peactor Statlards Enclocure:
I!atcrials Engineering Request for Additional Inforratica for Crys tal : liv:.: 3 cc s/cncl:
S. II::'tauer, DR R. Boyd, DPJ.,
R. DcYoung, DrL D. S'.:o vho l t, DitL R. :Iaccary, DRS S. Pawlicki, DR3 M. Eolotsky, LRS
- !. Fairtile, DRS H. Faulkner, DRL A. Schwencer, DRI, 8003170
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CRYSTAL RIVI:R NUCLEAR GENERATING PLANT - U'iIT 3 (DOCKET NO. 50-302)
REQUEST FOR ADDITIONAL INFO:C ATION - lL\\TERIALS ENGINEERING, DRS i
REACTOR C00LR;T SYSTEM Fracture Touchnern To evaluate the adequacy of the proposed heatup and cooldown limits for this plant, provide the following informatien:
1.
For all pressure-retaining ferritic componente of the reactor coolant pressure boundary whose 1cuest press ri:ntien tenperature* will be below 250*F, provide the naterial toughness properties (Charpy V-notch impact test curvas and dropweight test NDT terperature, or others) that hcvc been reported or specified for plates, forgings, piping, and weld raterial.
Specifically, for each cenponent provide the following data for natorials (plates, pipes, forgings, castinr,s, welds) used in the construction of the compenent, or your estimates based on the availabic data:
(a) The highest of the NDT temperatures obtained from DWT tests, (b) The highest of the tenperctures correspendine, to the 50 f t-lb s
1 value of the C f racture energy, and y
- Lowest pressurization temperature of a component is the lowest temper-ature at which the pressure within the component execeds 25 percent of the systen normal operating pressura, or at which the rate of tenperature change in the conpenent material exceeda 50 F/hr., under normal operation, systen hydrostatic tests, or transient conditions.
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(c) The lowest of the upper shelf C energy valuca for the " weak" y
direction (UR direction in plates) of the material.
2.
Identify the locatica and the type of the material (plate, forging, veld,.ctc.) in cach component for which the data listed above were obtained.
Where these fracture toughness para-eters cecur in rore than one plate, forging or veld, provide the information requested in 1.(a), (b) and (c) for each of them.
3.
For reactor vessel beltline natorials, including velds, provido the infornation requested in 1. and 2, and in addition specify:
(a) The highest predicted end-of-life transitien tenperature corres-pending to the 50 ft-lb value of the Charpy V-notch fracture energy fcr the "ueak dircction" of tha material (WR directica in plates) and (b) The tinitum upper shelf energy value which will be acccctable for t
continued reactor operatien toward the end-of-service life of the vessel.
4 Furnish the proposed heatup and ccoldown curves which will be used to i
control the pressure and temperatures to which the ferritic material i
of the. reactor coolant pressure boundary will be exposed during operation of the plant until the scheduled removal of the first l
natcrial capsule.
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, ~. CRYSTAL RIVER 3 REACTOR COOLAL'T SYSTCI Reactor Vee,sel ?faterial Surveillance Proaren 1.
For the predicted ::DT torperature shift of 250 r (TSAR, page 4-25) at least 5 capsules are required by the AEC proposed " Reactor vessel IIaterial Surveillance Progran T.equirenents", 30.55a, Appendix H, pub-lished in the rederal Re-ister en July 3, 1971 Each of these sur-veillance capsulca should include specinens from the base netal, heat-affected zone and the ucid netal, as recornended in the AST:! E-185, Sectica 3.3.
Sectica 4.3.3 ef the TS!i refers to t r.c report DN -lC006 for the descriptien of the curveillance pre ra-canaiatin~ of 6 capsules, caly 3 of t.Gich contain weld natal specinens.
In effect,the proposed surveillance progran consis ts of caly 3 capcules centaining, the required nunber and type of 1"n,ct specinens.
Provide the proposed capsule 'tithdrweal schedule, and describe how the fracture toughness of the veld netal will be enitored throughout the life of the reactor vessel.
' CRYSTAL RIVER 3 REACTOR C0013"T SYSTEM Sensitized Stainicss Steel 1.
Describe the plans which were folicued to avoid partial or local severe sensitization of austenitic stainicss steel during heat treatncnts and welding operations for core structural load bearing members and component parts of the reactor coolant pressure boundary.
Describe velding n2thods, heat input, and the quality controls that were enployed in weldine, austenitic ctainless steci compencats, 2.
If nitronen was added to stainicas steel types 304 or 316 to enhance its strength (as pernitted by ASME Ccde Case 1423 and USAS Case 71),
provide justificatten that such raterial t-ill not increase tendency to sensitization at heat affected senes during welding,.
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5-CRYSTAL RIVCR 3 REACTOR C00L.'J.;T SYSTE:t Electroclac 1lcidinq 1.
Since tha process of electroslag ucidinr, will be used in the f abricatien of conpencnts within the reactor coolant boundary (i.e., stcan generator shells and pressurizer shell), d: ascribe the process variables and the quality control procedures applied to achieve the desired material prop-crties in the base metals, heat affected zones, and welds.
2.
Indicate whether electrealag ucidinc: vill be used in fabricatien of any other components, particularly the.,e rede f rem stainicss steel.
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CRYSTAL RZVER 3 REACTOR COOL'.:TI SYSTPt Puno F1'.-/: eel Intecrity Provide the folleuing infornatica regardiac, the prinary coc] ant punp flywhccls:
1 State the type of natcrial used for the punp flywheci and its mini-nun specified yield strer.r,th.
Indicate the nil-ductility transitien (NDT) tenperature specified for the raterials, as obtained from drop-uci;ht teste (C'.:T), the nininun acceptabic Charpy V-notch (C ) upper y
chcif ener;y level in the verk directica (1:R orientacien in plates),
and the frccture teep,hnccs of the -a terial at the nornal operating tenpera ture of the fly checl.
2.
Describe the nondestructive exar.inations to be perforced on the finished flywheel.
State if the fly' heels vill be subjected to a w
100 percent value'etric ultrasenic inspection usina precedures and acceptance criteria equivalent to these specified for Class A vessels in Sectica III of the AS:E Soiler and Pressure Vessel Cede.
Describc the surface inspecticas to bc perforned on flywheel bores, keyvays and drilled holes.
3.
State the design stress specified for the flywhcc1 as a percenta;c of the mininum specified yield stren3th, for the normal operatine, speed and the design overspeed condition.
4 State if the calculated cc-hined prinary stresses in the flywheel, at the normal operatina speed vill include the strasses due to the inter-ference fit of the wheel on the shaf t, and the stresses due te centrifucal
- forces,
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State the highest anticipated overspeed of the flywhcci and the basis for this assumption.
6.
State the estimated naxinun rotational speed that the flywheel attains
' in the event the reactor coolant piping breaks in either the suction or discharne side of the pamp.
In addition, describe results of any studies directed touards:
(1) determining the naximum speed the punn or notor can reach due to physieni limitations (e.g., the speed at which the pump impeller seizes in the year rints due to r,rowth from centrifugal forces or the speed at which rotor parts conc loose and l.
grind or bind to prevent further increase in speed); (2) establishing speed and torque for various pipe break sizes; (3) devisinr, neans to 1
disenacge the notor from the punp in the event of purp overnpced; (4) verifying that pump fra,,ents generated at na::itum speed do not ponctrate the pump casing and that any nissiles leavine in the blou-doua jet do not panetrate containnent; (3) establishing failurc speeds 1
for notor parts and whether they will penetrate the rotor frsne and 1
if~so with what energy; and (6) defining a mininum rotor sciaurc tino.
- 7. - State the rotational speed that will be specified for the preoperational overspeed tests of the. flywheel.
8.
Describe the inservice inspection progran proposed for the flywheels, i
state.the areas to be inspected, access provicions, type and frequency of inspections, and the acceptance criteria.
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9 Provide the nanc of the manufacturer of the notot and sketches of the arrangement of the flywheel en the motor.
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' CRYSTAL RIVER 3 REACTOR C00lR;T SYSTDI Leakace Detection Sveten Provide the follouing information regarding the reactor coolant leakage detection systen:
1.
Describe the methods that will be used to determine coolant :cahage from the.rcactor coolant system.
Provide suf ficient detail to indi-cate that a ninimum of two systens with diverse modes of operation c<ll be installed in the plcnt.
2.
Describe the cathods used to provide positive indications in the control room of leakage of coolant from the reactor coolant systen to the containe.ent.
3.
Discuss the adequacy of the leak detectica subsysten 'Qich depends on reactor coolant activitr for detcetion of chcnaes in leahage during the initial period of plant creration when the ecolcat activ-ity may be low.
4
.L'ith reference to the proposed maximun allowable leakage rate f rom unidentified sources in the reactor coolcat pressure boundary, furnish the following information:
(a) The length of a through-wall crack that would leak at the rate of ~ the proposed limit, as a function of wall thickness.
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(b) jaul ratio of that length to the length of a critical through-i-
p vall crack, based on the application of the principles of fracture techanics.
(c).Tne machenatical ecdel and data used in such analyses.
5.
Specify the proposed naxi=um allouabic total leakage rate for the
. reactor coolant pressure boundary, and the basis for the proposed limit.
Furnish the ratio of the proposed licit to:
(a) The normal capacity of the reactor coolant rakcup systen.
(b) The noreal capacity of the containnent uater renoval systen.
6.
Provide the sensitivity (in gp=) and the response time of cach leak detcetica systen.
For the containment air activity monitors, provide i
the sensitivity and the response tire as a function of the percentace of failed fuel rods or of the corrosion product activity in the reacter
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coolant, as applicable, j
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Estinate the anticipated nornal total leakage rates and major leahac,e
- sources en the basis of operational experience from other plants of L
similar design.
S.
. Describe the adequacy of.the proposed Icakage detection systems to differentiate between identified and unidentified leaks from components J
uithin the primary reactor containment and indicate which of these systems
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provide.a neans for locating the general area of a leak.
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Provide the time interval in thich reactor will be shutdoun if either the total or unidentified leakage rate lin.t is exceeded.
10.
Describe the proposed tests to denonstrate sensitivitics and opera-bility of the leakage detection systems.
11.
The air sanpling system, RM-Al, associated with the containment 5
purge valves appears to possess 10 tines core continuous crarating sensitivity in detecting containncnt atnosphere radioactivity than air sanping systen, E:!-AS, proposed to detect primary coolcat Icakage into the containacnt r.ttesphere.
Ccusider adcptinn Systen R.'!-Al to detect primary ecolant leakap.c ce to addina a cotatinuous readinq air particulate conitor to System R:!-A6.
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i REACTOR COOLANT SYSTEM Inservice Inspection Program 1.
Describe the design and arrangement provisiens for access to the reac' tor coolant pressure boundary as required by Sections IS-141-and IS-142 of Section XI of the ASME Boiler and Pressure Vessel Code -
I Inservice Inspection of Nuclear Reactor Coolant Systen.
Indicate the specific provisions made for access to the reactor vessel for 8
examination of the welds and other compenents.
1 2.
- Section XI of the ASMF Boiler and Pressure Vessel Coda reco nizes the problems of c::anining radioactive areas where access by personnel will be inpractical, and provisions are incorporated in the rules for the cxanination of such areas by renote maans.
4 In sc=e cases the equipnent to be used to perfern such examination is under development.
Provide the f ollcuing information with respect to your inspection program:
-l (a) Describe the equipnent that will be used, or is under developnent for use,. in perforcing the reactor vessel and nozzle inservice inspections.
(b) Describe the systen to be used to record and compare the data from the baseline inspection with the data that will be obtained fron subsequent inservice inspections.
(c) Describe the procedures to be follcwed to coordinate the develcpment of the remote inservice inspection equiprent with the access previsions' for inservice inspection afforded.by the plant design.
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- CRYSTAL RIVER 3 CONTAI:7'IENT Leahace Testine Protran 1.
Describe the containnent lenhage testing progran and indicate the degree of corpliance with the AEC preposed " Reactor Containrent Leahage Testing for 'Jater Ccoled Power Reacters", 5 50.54 (n), Ap-pendix J, published in t:1c Federal Recister en August 27, 1971.
2.
Describe the design features of the contain ent airlocks that will pernit testing of the airlocks at the calculated peak containr. cat pressure. Describe t:1e te st r.athod that *:ill ac u ;ed to veri?"
leak tightness of air lock dcors, dcor penetratiens and door -aske ts.
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' CRYSTAL RIVER 3 E::CI!:EERED SAFETY FEATCRES Inservice Insnection Tronran 1
Describe the inservice inspection pro,.ran for fluid systens other than those composing the reactor coolant pressure boundary, including items to be inspected, accessibility requircrente, and the f requenty and types of inspection. The fluid systens to be censidered are applicabic enginecrett safety systcas, reactor shutdown systens, cooling water systems, sad the radioactive vaste trentrent syctens.
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