ML20030D655
| ML20030D655 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/21/1981 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Hukill H METROPOLITAN EDISON CO. |
| References | |
| TASK-2.K.2.13, TASK-TM TAC-45202, TAC-59974, TAC-59975, NUDOCS 8109140225 | |
| Download: ML20030D655 (9) | |
Text
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WT 2t q Docket No. 50-289
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Mr. Henry D. Hukill, Vice President
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Y and Director - THI-1 C
Hotropolitan Edison Company 5
AUG 2 81981a.- 13 t
P. O. Box 480 C u.s.euummuaussa 5 * **J Middletown, Pennsylvania 17057
Dear Mr. Huk111:
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SUBJECT:
"RESSURIZED THEPJ1AL SHOCK TO REACTOR PRESSURE VESSELS We have reviewed the PWR Owners' Groups responses of May 15, 1981 and the Itcensees' responses of May 22, 1981 to our letter dated April 20 1981 3
concerning the subject issue. The EPRI work which bears on the issue was included in the licensecs' responses. On the basis of our independent review, of the plants where neutron irradiation bas significantly reduced the fracture toughness of the reactor pressure vessels (RPVs), all plants 9
could survive a severe overcooling event for at least another year of full power operation. Ilowever, we believe that additional action should be j
taken now to resc1re the long-term problems.
e This belief is based upon our analyses which indicate that reductious in fracture toughness for some RPVs are approaching levels of concern.
It is also based in part on the fact that any proposed corrective act*
nust allow edequate lead tire for planning, review, approval, procureN and installation. These conclusions were recently discussed with the P1ft i
owners Groups on July 28-30, 1981. At those neatings, the Owners Groups reviewed the programs underway at the three PUR vendors which are designed i
to scope the magniti de and applicability of the generic problen and to be l
N completed by late 1981. The three programs appeared to contain the necessary Eo elenents for resolution of the problem on a generic basis and the NRC plans
>;f to nake full use of the reports due by the end of the year. While the R*
vendors and Owners Groups are to be co e nded and encouraged in addressing l
M8 the generic issue, there is also a need for plant-specific infornation for l
your plant.
o9 Based on current vessel reference temperature and/or syster. characteristics, I
we have identified Ft. Calhoun, Robinson 2, San Onofre 1, Maine Yankee, l
Oconee 1 Turkey Point 4, Calvert Cliffs 1 and Three Nile Island 1 as plants 3
from which we require additional information at this time.
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The staff has used the tine-dependent pressure and tenperature data from the March 20, 1978 Rancho Seco transient as a starting point for our evaluation of this issue because:
(1) it is the nost severe overcooling l
event experienced to date in an operating plant;.(2) it is a real, as L
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fir llenry D. Hukill opposed to a postulated, event; and (3) it was severe enough that it could challenge the RPV when combined with physically reasonable values of f r-radiated fracture toughness and initial crack size.
In future reviews the staff plans to use the stean line break accident or other appropriate transient / accident in order to estinate ain'num operational tires available before plant modifications are required.
Using calculated RPV steel nochanical properties, credible initial flaw sizes, reasonable themal-hydraulic paraneters, and a simplified pressure-tenperature transient similar to that observed during the Rancho Seco event, the staff has concluded that all operating p! ants could safely 4
survive such an event at the present time and for at Icast an additional year of full power operation. Ilovever, because of the required lead times for future actions, the margins in time for long ter:n operation are not large, and there is considerable uncertainty in the probability that sinflar or rore severe transients may occur.
it is clear that positive action ranst be initiated soon for those plants with significantly high transition temperatures. As indicated above, several such plants have been selected by the staff, based on estimates of the current reference temperature for the nil ductility transition (RT
) of the RPVs.
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The need to initiate further action at this time is emphasized by the I
recognition that implerentation of any proposed fixes or remedial actions must allest for adequate lead time. Because long-tem solutions may require 0
a year or more, you should explore short-tem approaches as well. Al though clear, concise instructions snould be previded to operators to reduce the likelihood of repressurization during overcooling transients, the HRC staff believes that reliance or operator actions to prevent repressurization during an overcooling tra, *ent will be very difficult to justify as an acceptable long-tem soluti i to the problen.
In accordance with 10 CFR 50.54(f) of the Commission's regulations, you are requested to submit written statements, signed under oath or affirnation, to enable the Comission to detemine whether or not your license should be modi-j fied, suspended or revoked. Specifically, you are requested to submit the following infomation to the MPC within 60 days from the date of this letter:
(1) Provide the RT values of the critical welds and plates (cr for-NDT gings) in your vessel for:
(a) initial (as-built) conditions and location (e.g.,1/4 T) and (b) current conditions (include fluer.cc level) at the RPV inside carbon steel surface.
f tr. Ifenry D. Hukill (2) At what rate is RT increasing for these welds and plate material?
HDT (3) What value of RT for the critical welds and plate caterial do NDT you consider appropriate as a limit for continued operation?
(4) What is the basis for your proposed limit?
(5) Provide a listing of operator actions which are required for your plant to prevent pressurized themal shock and to ensure vessel integrity. Include a description of the circumstances in which these operator actions are required to be taken.
Included in this sumary should be the specific pressure, tmperature and level values for:
a) high pressure injection (HPI) temination criteria presently used at your facility, b) HP! throttling criteria and instruction presently used at your facility anu ) criteria for throttling feedwater presently used at your facility. For each tequired operator action, give the infomation available to the operator and the tino available for his decision and the required action. State how each required operator action is incorporated in plant operating procedures and in training and requalification training prograns.
p You are also requested to submit a plan for Three title Island, Unit No.1 to the NRC within 150 days of the date of this letter that will define
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actions and schedules for resolution of this issue and analyses supporting cor,tinued operation. We request that you include consideration and evalua-
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tion of the following possible actions:
(1) reduction of further neutron radiation damage at the beltline
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by replacement of outer fuel assechlies with dummy assemblie:
or other fuel management changes; (2) reduction of the thernal shock severity by increasing the ECC water temperature; (3) recovery of RPV toughness by it place annealing (include the basis for demonstrating that your plant meets the requirements in 10 CFR 50 Appendix G IV C);
(4) design of a control system to mitipte the initial thermal shock and control repressurization.
For these, as well as for any other alternative approaches, provide implementation schedules that would assure continuance of adequate safety margins.
In the interest of efficient evaluation of your submi' cal, we request that you include with the above plan, a response to tne enclosed request for additional information.
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F "r. Itenry D. liukill Due to the nature of this review, and the past reviest effort that has been expended, we consider the above schedules to be reasonable; however, inform us within 30 days if you anticipate conflicts with previous comitnents with either submittal and a basis for any delay. We also expect participation by the appropriate P!IP. Owners Group and PlSSS vendors in develeping solutions to the problem.
Sincerely, Original signed by Darrell G. Eisenhut, Director Division of Licensing Office of !!uclear Reactor Regulation
Enclosure:
Request for Additional Information cc w/enclosuie:
See next page 1
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. - Metropolitan Edison Company 1-cc w/ enclosure (s):-
Mr. Marvih I". Lewis '
Dr. Walter H. Jordan 6504 Bradford Terrace-881 W. Outer Drive Philadelphia, Pennsylvania 19149 Oak Ridge, Tennessee 37830 Walter W. Cohen, Consumer Advocate Dr. Linda W. Little Department of Justice 5000 Hermitage Drive i
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Anti-Nuclear Group Representing...
Assistant Solicitor
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Knupp and Andrews 24S W. Philadelphia Street P. O. Box P York, Pennsylvania 17404 407 N. Front Street Harrisburg, Pennsylvania 17108
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John Levin, Esq.
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Pennsylvania Public Utilities Comm.
John E. Minnich, Chairman Box 3265 Dauphin Co. Board of Commissioners Harrisburg, Pennsylvania 17120 Dauphin County Court. house i
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. U.S. iluclear Regulatory Commission
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Uashington, DC 20006 e
Metropolitan Edison Company.
,s Mr. Thomas Gerusky Ms. Jan2 Lc2 Bureau of Radiation Protection R.D. 3, Box 3521 Department of Environmental Resources Etters, Pennsylvania 17319 P. O. Box 2063 Harrisburg, Pennsylvania 17120 Karin W. Carter, Esq.
505 Executive House P. O. Box 2357 Harrisburg, Pennsylvania 17120
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EIS C0ORDINATOR Curtis Building (Sixth Floor)
Mr. C. W. Smyth,
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l Metropolitan Edison Company.
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ANGRY 32 South Beaver Street York, Pennsylvania '17401 William S. Jordan, III, Esq.
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York College of Pennsylvania Country Club Road York, Pennsylvania 17405 Mr. Donald R. Haverkamp Senior Resident Inspector (THI-1)
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o Enclosure REQUEST FOR ADDITIONAL INFORMATION 1.
Geometry Geometrical description including design and as-built (when available) dimensions of the core, assemblies, shroud / baffle, thermal shield, downcomer, vessel, cavity, and surrounding shield and/or support structure.
2.
Material Description Region-wise material composition and material isotopic number densities (atoms / barn-cm) for the core, near-core regions and RPV, suitable for neutron transport calculations.
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3.
Neutron Source Present and expected EOL:
a) Assembly-wise and core power history (EFPY).
b) Rod-wise and core power history (EFPY) for peripheral assemblies.
c) Core average axial power history. distribution.
4.
Vessel Fluence a) Description of available calculations of the vessel fluence ' including fluence values, locations, and corresponding power histories (EFPY),
including 1/4T,1/2T and 3/4T through the RPV.
b) Description of available capsule-inferred vessel fluences including fluence values, locations, and corresponding power histories (EFPY).
5.
Surveillance Capsules a) Capsule materials, radial and axial dimensions and locations.
i b) Capsule fluence measurements, together with the accumulated power history (EFPY) and a description of the lead factors used to extra-polate the measurements to the peak wall fluence location.
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6.
Vessel Welds Axial and azimuthal locations of vessel weld-seams with respect to the core. Overlay of current fluence map with weld locations.
Identify the critical welds, vertical and circumferential, and give the weld wire heat numbers. Give weld chemistry for the critical welds. For each weld wire heat number, report the estimated mean copper content, the range and the standard deviation, based on all the reported measurements for that weld wire heat. The welds may be surveillance weldments for your vessel or others, nozzle dropouts that contain a weld, weld metal qualification data, or archive material.
In the absence of any information, assume that copper content is at its upper limit (0.35 percent when using P.G.1.99, Rev.1) and that the nickel content is high.
7.
Systems Analysis a) Provide a list of transients or accidents by class (for example:
excessive feedwater, operating transients which result from multiple failures including control system f ailure's and/or. operator error, steam
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line break and small break LOCA) which could lead to inside vessel fluid temperatures of 300 F or lower. Provide any Failure Modes and Effects Analyses (FMEAs) of control systems currently available or reference any such analyses already submitted. Provide the ana' lysis of the most limiting transient or accident with regard to vessel thermal shock con-siderations.
Estimate the frequency of occurrenc'e of this event and provide the basis for this estimate. Discuss the assumptions made regarding reactor operator actions.
b)
Identify the computer programs used to calculate the limiting transient or accident.
Indicate the degree to which the computer program.
used have been verified and any other additional verification required to demonstrate that the computer program r.odels adequately treat the identi-fied important physical models (i.e., ECC mixing, heat transfer, and repressurization).
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