ML20141D663
| ML20141D663 | |
| Person / Time | |
|---|---|
| Issue date: | 03/27/1986 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Turk C ARKANSAS POWER & LIGHT CO., BABCOCK & WILCOX OPERATING PLANTS OWNERS GROUP |
| References | |
| NUDOCS 8604080205 | |
| Download: ML20141D663 (2) | |
Text
_
March 27,1986 k
Mr. Charles H. Turk, Chairman DISTRIBUTION: [KET JFI)
Analysis Committee NRC PDR E6 GT FDR WPaulson B&W Owners Group PD#6 Rdg FJMiraglia JPage Arkansas Power & Light Company 0 ELD EJordan Misc. File P. 0.. Box 551 BGrimes JPartlow Little Rock, Arkansas 72203 RIngram ACRS (10)
SUBJECT:
REACTOR VESSEL CLOSURE REGION THERMAL STRESS DURING NATURAL CONVECTION C00LDOWN - GENERIC ISSUE NUMBER 79
Reference:
(1)
J. H. Taylor to R. C. DeYoung, Unanalyzed Reactor k
Vessel Thermal Stress During Cooldown, dated March 18, 1983.
Dear Mr. Turk:
This letter confirms a recent telecon between you and Mr. Joel Page of the NRC staff. Reference I descr. bed a concern regarding thennal stress in reactor vessel flanges and studs in a B&W reactor vessel during natural circulation cooldown. The B&W Owners Group provided an analysis of this concern by letter dated October 15, 1984, including B&W Document Nos.
32-1151155-00 and 77-1152846-00.
We have performed a preliminary review of these documents.
In order to con-tinue the review, certain deta'ls regarding geometric parameters of reactor i
vessel components are needed. We feel that this information may be contained in some of the references (pp. 102 - 103) of Document No. 32-1151155-00.
Please provide two copies of References 7 through 15, 19 (Report No. 2), 27 and 32.
Additionally, the following information is needed regarding materials proper-ties:
For the SA-508-C1.2 (3/4 Ni-1/2 Mo-1/3 C -V) steel used for the FA-177 r
reactor vessel shell (1) Nil ductility reference tuperature, RT
, determined according toNB-2331SubsectionNBDivisionIIIofDIheASMEBoilerand Pressure Vessel Code.
(ii) " Adjusted reference temperature", as defined in 10 CFR 50 Appendix G paragraph II-E, which is the nil ductility reference temperature adjusted for irradiation effects.
For SA-540-Gr.823-C13 bolting material used for the preload bolts of the FA-177 reactor vessel (i) Charpy V-notch test results (C tests) preferably including y
lateral expansion results.
_8604080205 060327 1if C
2-1, It would be appreciated if this information could be provided within 45 days in order to stay on schedule with our review.
The reporting and/or recordkeeping requirements contained in this.. letter.
affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely,
- +.LGLuh staxa yg 19.F. SAL 4* '
John F. Stolz, Director PWR Project Directorate #6 Division of PWR Licensing-B 1
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Subject:
Unanalyzed Reactor Vessel Thermal Stress Duri; molcown b
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Dear Mr. DeYoung:
The purpose of this letter is to describe an issue which b n k
been processed as a preliminary safety concern at B&W.
"his concern has evolved out of further consideration of the
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St. Lucie upper head voiding event and has the potential u
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be generic to all Pressurized Water Reactors.
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L The concern.that thermal stresses, beyond those considered in the original design, may develop in the RV flanges and stels 4-due to large axial temperature gradients across the reactor A
vessel flanges.
These gradients could develop as a result of non-uniform cooling of the reactor coolant within the ve-
-1.
G These non-uniform effects in the reactor coolant may occur once N
the reactor coolant pumps are secured and the decay heat removal system has been actuated in the normal cooldown mode or daring a g
natural circulation cooldown.
During either mode, a relatively
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stagnant area exists in the upper head region.
Because of this stagnant flow region, there is poor thermal mixing between the 3
c fluid in the head and the fluid in the plenum and nozzle regions i )
of the reactor vessel.
Possible axial gradients between
'.50* and 200*F could produce thermal stresses in the vessel flange area 74 or in the studs during either of these two cooling modes that
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might exceed code allowables when added to stresses already con-sidered.
.c BtM's initial evaluation of this concern included calcule* inn the temperature profiles of the reactor vessel head and mge and d,
the shell flange during natural circulation cooldown.
The-
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average estimated cooldown rate of the reactor vessel head for c'
j B&W 177 FA plants after the reactor coolant pumps were secured
- '4 was about 2*F/hr and decreased to less than l'F after approximately 28 houro.
The calculations assumed a large stagnant volure of j
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' : %g Mr. R. C. DeYoung March 18, 1983 28y fluid in the region from the plenum cever to the top of tha x/
reactor vessel head.
Thus the thermal calculations took no Y
credit for additional cooling that might result from convection h
in the fluid in the upper head region since r.o data was avail--
W' able as a basis for establishing this cooling.
Only a very small amount of heat transfer was calculated to occur between the stag-nant fluid and the lower temperature coolant flowing below.
The b
balance of the heat loss was by conduction through the in w1ated tA 1 head to the containment.
Temperature differences of up to 200'F M
developed between the vessel shell at the nozzle belt reginn and p&)
the shell flaage and upper head region.
Only a small temperature
'.W 1 difference was calculated to exist in the metal from the top of dome to the bottom of the head flange.
As a result of these Aq temperature gradients, high thermal stresses were postulated in the radius at the outer transition between the vessel flange forging and the nozzle belt forging and in the closure studs.
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{4 Although fluid temperature differences of greater than 150*F have been experienced in several plants during cooldown, the extent to which these fluid differences cause meta 1 temperature oiiferences p~
specifically in the flange area is not known.
Furthermore, the cooldown rate of the stagnant area as calculated is believed to be conservative.
This implies that if the gradients are unaccepc-h able they may be reduced simply by slowing down the cooldown rate t:W of the reactor coolant during the latter portion of the ecoldown operation after the reactor coolant pumps are secured.
,k W
W1 As a part of the B&W investigation of this matter, we contacted Ng:
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EPRI/NSAC to determine whether they had investigated actual gradi-ent effects in connection with the St. Lucie event.
They had done some work using the MARC code which is a general purpose bM finite element code.
This unpublished EPRI investigation ased El k"s cooldown rates of the coolant in the vessel head region in the H
neighborhood of 20*F/hr.
These significantly higher ratar produced improved temperature gradients and the resultant vessel st: nes Q
were not excessive.
Stresses in the reactor vessel stun ra e.I not investigated.
This phenonenon is not likely to be a serious near terra ty a
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concern but it does represent an unanalyzed situation w t t u potential for margin reduction over plant life.
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The NRC Staff may have m oss to data from the St. Lucie or similar events that v0M 9naole a better assessment of actual g
cooldown rates of '.R-a
.1 in the upper head region or u perature g& c profiles of the me;n krc studs in RV head closure and nonie belt regions.
In lieu of such information it would seem tnat the most direct approach to further this investigation la to obtain p
actual reactor vessel and head flange metal temperatures during WN a natural circulation cooldown.
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R. C. DeYounc 42;,4 March 15t, 1983
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BW Sn appriced the D W opercting plant cuners of thi:.'
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towar<1 resolution.
Any input the NRC can provide to holp '
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- achnic-T evaluntion of thic issuc would be np'preciated, f/,[
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If you have any questions on this matter please call me m
$W (804-165-7417) or T.
T, B.516wi n (904-385-3142) of my ct:."~
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