ML20030D654
| ML20030D654 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/21/1981 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | William Jones OMAHA PUBLIC POWER DISTRICT |
| References | |
| NUDOCS 8109140204 | |
| Download: ML20030D654 (8) | |
Text
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AUGUST 2 1 1981 x
Docket No. 50-285 ei Q
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"g3 2 8198W }3 Mr. W. C. Jones T
Division fian'iger, Production 3-Operations 1 V '8' J
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1' 1623 Ilarney Street Omaha, Nebraska 68102 WM g
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Dear lir. Jones:
SUBJECT:
PRESSURIZED Ti!ERMAL SHOCK TO REACTOR PRESSURE VESSELS We have reviewed the PWR Owners' Groups responses of May 15, 1981 and the licensees' responses of May 22, 1981 to our letter dated April 20, 1981 concerning the subject issue. The EPRI work which bears on the issue was included in the licensees' responses. On the b' sis of our independent review, of the plants where neutron irradiation has significantly reduced the fracture toughness of the reactor pressure vessels (RPVs), all plants
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could survive a severe overcooling event for at least another year of full l
power operation. Ilowever, we believe that additional action should be g
taken now to resolve the long-tern problems.
TThis belief is based upon our analyses which indicate that reductions in P
fracture toughness for sone RPVs are approaching levels of concern.
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It is also based in part on the fact that any proposed corrective act in rust allow adequate lead time for planning, review, approval, procurement and installation. These conclusions were recently discussed with the PWR Owners Groups on July 28-30, 1981. At those meetings, the Owners Groups revicwed the prograns undeceay at the three PWR vendors which are designed to scope the magnitude and applicability of the generic problen and to be mia) cry'pleted by 1 ste 1981. The three prograns appeared to contain the necessary Eo elements for resolution of the problem on a generic basis and the NRC plans J
to nake full use of the reports due by the end of the year. While the c *=
vendors and Owners Groups are to be ccmended and encouraged in addressing 80 the ceneric issue, there is also a need for plant-specific infomation for
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9 your olant.
OU Based on current vessel reference tenperature and/or systen characteristics, h@
we have identified Ft. Calhoun, Robinson 2, San Onofre 1, flaine Yankee, Oconee 1. Turkey Point 4, Calvert Cliffs 1 and Three Nile Island 1 as plants zu from which we require adt lonal infornation at this time.
The staff has used the tii o-depeudent 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 most severe overcooling event experienced to date in an operating plant; (2) it is a real, as
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tir. W. C. Jones opposed to a postulated, event; and (3) it was severe enough that it could challenge the PPV when combined with physically reasonable valuer of ir-radiated fracture toughness and initial crack size.
In future reviews the staff plans to use the steam line break accident or other appropriate transient / accident in order to estimate minimum operational times available before plant nodifications are required.
Ilsing calculated RPV steel rechanical properties, credible initial flaw sizes, reasonable thernal-hydraulic parametcrs, and a simplified pressure-temperature transient similar to that observed during the Rancho Seco event. the staff r.es concluded that all operating plants could safely survive such an event at the present time and for at least an additional year of full pcuer operation. However, because of the required Icad times for future actions, the margins in tine for long tem operation are not large, and there is considerabic uncertainty in the probability tnat similar or more severe transients may occur.
It is clear that positive action must be initiated soon for those plants with significantly high transition temperatures. As indicated above, several such plants have beer selected r
by the staff, based on estimates of the current reference temper 4ture for the nil ductility transition (RT
) of the RPVs.
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NDT The need to initiate further 3ction at this time is emphasized by the recognition that implementation of any proposed fixes or remedial actions oust allow for adequate lea f time. Because long-tem solutions may require a year or more, you should explore short-tera approaches as well. Although clear, concise instructions should be provided to operators to reduce the likelihood of repressurization during overcooling transients, the 11RC staff believes that reliance on operator actions to prevent repressurization during an overcooling transient will be very difficult to justify as an acceptable long-term solution to the problem.
In accordance with 10 CFR 50.54(f) of the Comission's regulations, you are reonasted to submit written st?tements, signed under oath or affirmation, to enchle the Comission to determine whether or not your license should be rodi-fied, suspended or revoked. Specifically, you are requested to submit the vnllowing infomation to the NRC within 60 days from the date of this letter:
(1) Provide the RT values of the critical welds and plates (or for-NDT gings) in your vessel for:
(a) initial (as-built) conditions and location (e.g.,1/4 T) and Ib) current conditions (include fluence level) at the RPV inside carbon steel surface.
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a lir. W. C. Jones (2) At what rate is RT increasing for these welds and plate material?
NDT (3) What value of r.T for the critical welds and plate material de 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 circunstances in which these operatcr actions are required to be taken.
Included in this summary should be the specific pressure, temperature and level values for:
a) high pressure in.jection (HPI) temination criteria presently used at your facility, b) HPI throttling criteria and instruction presently used at your facility and c) criteria for throttling feedwater presently used at your facilit;y. For each required operator action, give the information available to the operator and the time 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 programs.
Ycu are also requested to submit a plan for Ft. Calhoun to the NRC within
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15C days of the date of this letter that will define actions and schedules a
for resolutier. of this issue and analyses supporting continued operation.
We request that you include consideration and evaluation of the following possible actions:
(1) reduction of further neutron radiation damage at the beltline by replacement of outer fuel assenblies with durrty assemblies or other fuel management changes; (2) reduction of the themal shock severity by increasing the ECC water tenperature; (3) recovery of RPY toughness by in-place annealing (include the basis for demonstrating that your plant neets the requirements in 10 CFR 50 Appendix G IV C);
(4) design of a control systen to mitigate the initial thermal shock and control repressurization.
For these, as wall as for any other alternative approaches, provide implementttion schedules that would assure continuance of adequate safety marcins.
In the interest of efficient evaluation of your subnittal, we request that you include with the above plan, a response to the enclosed request for additional infomation.
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n ffr. W. C. Jones Due to the nature of this review, and the past revf m Offort that has been expended, we consider the above schedules to be reasonable; however, inform us within 30 days if you anticipate conflicts with previous commitments with either subnittal and a basis for any delay. We also expect participation by the appropriate PW2 Owners Group and fiSSS vendors in developing solutions to the problem.
Sincerely, ori.1inal &ned by Darrell G. Eisenhut, Director Division of Licensing Office of t!ucicar Reactor Regulation
Enclosure:
Request for Additional Informtion cc w/ enclosure:
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r Omaha Public Power District cc:
Marilyn T. Shaw, Esq.
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, D. C.
20036 Mr. Emmett Rogert Chairman, Washington County Beard of Supervisors Blair, Nebraska 68023
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U.S. Environmental Protection Agency Region VII ATTN: Regional Radiation Representative 324 East lith Street Kansas City, Missouri 64106 Mr. Frank Gibson W. Dale Clark Library 215 South 15th Street Omaha, Nebrr..,k: 50102 Alan H. %irshen, Esq.
Fe11ran, Ransey 6 Kirshen 1166 Wooc' men Tov.er Onaha, Nebraska 68102
r. Denni s Kelley U.S.N.R.C. Resident Inspector D. 3. Box 68 Fort Calhoun, Nebraska 68023 l',r. Charles B. Brinkman Manager - Washington Hucicar Operations C-E Power Systems Combustion Engineering, Inc.
4853 Cordell Avenue, Suite A-1 Bethesda, Maryland 20014 e
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m 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 RPY, suitable for neutron transport calculations.
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) Descripticn of available capsule-inferred vessel fluences including fluence values, locations, and corresponding power histories (EFPY).
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Surveillance Capsules a) Capsule materials, radial and axial dimensions and locations.
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.
Yessel 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 wel ds. 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 quali ication data, or archive material.
In tne absence of any information, assume that copper content is at its upper limit (0.35 percent when using R.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 failures and/or.qperator error, steam 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 analysis of.the most limiting transient or accident with regard to vessel thernal shock con-siderations.
Estimate the frequency of occurrence of this event and provide the basis for this estimate. Discuss the assumptions made regardi'ng reactor operator actions.
b)
Identify the computer programs used to calculate the limiting transient or accident.
Indicate the degree to which the computer programs used have been verified and any other additional verification required to demonstrate that the computer program models adequately treat the identi-fied important physical models (i.e., ECC mixing, heat transfer, and repressurization).'
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