ML20010E925
| ML20010E925 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 08/21/1981 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Groce R Maine Yankee |
| References | |
| NUDOCS 8109090123 | |
| Download: ML20010E925 (8) | |
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I AUGUST 21301 Docket No. 50-309 mb' lpt[( NL J. I /v t fir. Robert 11. Groce I] AUG 2 81981>.) 4 Senior Engineer, Licensing 6 !!aine Yankee Atomic Power Company g\\d u.a g p 1671 Worcester Road 1 Framingham, Massachusetts 01701 ]
Dear Mr. Groce:
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SUBJECT:
PRESSURIZED THERMAL SHOCK TO REACTOR PRESSURE VESSELS 1 Ve have reviewed the PUR Owners' Groups responses of May 15, 1981 and the licensees' responses of May 22, 1931 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 basis of our independent } review, of the plants where neutron irradiation has significantly reduced the fracture toughness of the reactor pressure vessels (RPYs), all plants could survive a severe overcooling event for at least another year of full power operation. flowever, we believe that additional action should be i taken now to resolve the long-term problems. This belief is based upon our analyses which indicate that reductions in fracture toughness for sone RPVs are approaching levels of concern. It is also based in part on the fact that any proposed corrective action nust 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 neetings, the Duners Groups reviewed the programs underway at the three PWR vendors which are designed to scope "te magnitude and applicability of the generic problem and to be completed by late 1981. The three prograris appeared to contain the necessary elements #cr resolut'on of the problem on a generic basis and the !!RC plans to make full use of the reports due by the end of the year. While the vendors and Owners Groups are to be corronded and encouraged in addressing the generic issue, there is also a need for plant-specific information for or j T your plant. So 38 Based on current vessel reference temperature and/or systen characteristics, r i we have identifled Ft. Calhoun, Robinson 2, San Ono ce 1, Maine Yankee, r Oconee 1. Turkey Point 4, Calvert Cliffs 1 and Three Mile Island 1 as plants i n I 36 fron which we require additional infomation at this time. 88 o< The staff has used the tire-dependent pressure and teoperature data from om the March 20, 1978 Rancho Seco transient as a starting point for our l evaluation of cnis issue because: (1) it is the rest severe overcooling i event experienced to date in an operating plant; (2) it is a real, as l -.-.
1 m y l Mr. Robert H. Groce, opposed to a postulated, event; and (3) it was severe enough that it could challenge the RPV when combined with physically reasonable values 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 mininun operational times available before plant nodifications are required. Using calculated RPV steel mechanical properties, credible initial flaw sizes, reasonable thermal-hydraulic parameters, and a simplified pressure-temperature transient similar to that observed during the Rancho Seco event, the staff has concluded that all operating plants could safely survive such an event at the present time and for at least an additional year of full power operation. However, because of the required lead times for future actions, the margins in time for long term operation are not l large, and there is considerable uncertainty in the probability that sin 11ar or more severe transients may occur. It is clear that positive action nust he initiated soon for those plants with significantly high transition 1 temperatures. As indicated above, several such plants have been selected by the staff, based on estimates of the current reference temperature for I the nil ductility transition (RT ) of the RPVs. j NDT 1 j The need to initiate further action at this time is erphasized by the recognition that implementation of any proposed fixes or remedial actions
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must allow for adequate lead time. Because long-term solutions may require a year or more, you should exploro short-tern approaches as well. Although clear, concise instructions should be provided to operators to reduce the l likelihood of repressurization during overrooling transients, the NRC staff i believes that reliance on operator actions to prevent repressurization i during an overcooling transient will be very difficult to justify as an acceptable long-term solution to the problem. In accordance with
- CFR 50.54(f) of the Commission's regulations, you are I
requested to submit written statements, signed under oath or affirmation, to 2 enable the Commission to determine whether or not your license should be modi-fled, suspended or revc.ed. Specifically, fou are requested to submit the following information to the HRC within 60 days from the date of this letter: I (1) Provide the RT values of the critical welds and plates (or for-IIDT j gings) in your vessel for: (a) initial (as-built) conditions and location (e.g.,1/4 T) and l (b) current conditions (include fluence level) at the RPV inside carbon steel surface. i \\ l J a
l Mr. Robert H. Groce (2) At what rate is PT increasing for these welds and plate material? NDT (3) Vhat value of RT for the critical welds Md plate material do NDT you consider appropriate as a limit for certinued operation? (4) yhat is the basis for your proposed limit? (5) Provide a listing of operator actions which are required for your plant to prevent pressurized thermal shock and to ensure vessel integrity. Include a description of the circunstances in which these operator actions are required to be taken. Included in this summary should be the specific pressure, teaperature and level values for: a) high pressure injection (HPI) temination criteria presently used at your facility, b) HPI throttling criteria and instruction presently used at your facility and c) criteria for throttling feodwater presentl= used at your facility. For each required operator action, give the inforration available to the operator and the time available for hir decision and the required action. State how each required operato' action is incorporated in plant operating procedures and in trairing and requalification training prograns. You are also requested to submit a plan for Maine Yankee to the NRC within 150 days of the date of this letter that will define actions and schedules for resolution 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 assenhlies with dumy assemblies or other fuel nanagement changes; (2) reduction of the themal shock severity by increasing the ECC water temperature; (3) recovery of RPV toughness by in-place annealing (include the basis for demonstrating that your plant neets the requircaents in 10 CFR 50 Appendix G IV C); (4) design of a control system to mitigate the initial thernal shock and control repressurization. i For these, as well as for any other alternative approaches, provide inplementation schedules that would assure continuance of adequate safety margins. } } In the interest of efficient evaluation of your submittal, we request that you include with the above plan, a response to the enclosed request for additional informatinn.
_ = _ = 1 I w~r o e5 y Mr. Robert H. Groce 4 Due to the nature of this review, and the past review effort that has been expended, we consider the above schedules to be reasonable; however, infonn us within 30 days if you oriticipate conflicts with previous comaitments.ith either submittal and a basis for any delay. We also expect participation by the appropriate PWR Owners Group and HSSS vendors in developing solutions to.the problem. i Sincerely, t original signed M r Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Information cc w/ enclosure: i See next page f 1 4 I 4 i 1 4 4 4 omcE > g, 0RBJ4:DL AD-0R L D ......N............. SURNAME. ...........m.......... ... k v l...... Lq.L.1............ /2 /B.1....... . 3.&!.l.1.. f DAM ) sac reau sia 0o-8c3 nacu cao OFFICIAL RECORD COPY usa m ini-=,.no ~..-.
DISTRIBUTION: Docket File KKniel NRC PDR NAnderson L PDR RJohnson TERA JClifford NSIC BSheron ORB #5 Rdg DCrutchfield ORB #4 Rdg EThrom ORB #1 Rdg WHazelton ORB #3 Rdg RKlecker Gray Files (8) Glainas B&W Owners Group NRandall CE Owners Group JMartore Westinghouse Owners Group TNovak HDenton JStolz DEisenhut SVarga RVollmer RAClark SHanauer CTrammell RMattson DNeighbors TMurley SNowicki RSnaider FSchroeder RJacobs OELD (5) HSilver AE00 MGrotenhuis IE (7) PWagner ACRS (10) BRequa ORAB DJaffe SEPB GVissing MConner RIngram CHarwood PKreutzer RDiggs EHylton NHughes CParrish PWoolley HSmith
s - i .. Maine Yankee Atomic Power Company e e cc: E. W. Thurlow, President Mrs. L. Patricia Doyle, President Maine Yankee Atomic Power Company SAFE POWER FOR MAINE Edison Drive Post Office Box 774 Augusta, Maine 04336 camden, Maine 04843 Mr. Donald E. Vandenburgh First Selectman of Wiscasset Vice President - Engineering Municipal Building Yankee Atomic Electric Company U. S. Route 1 20 Turnpike Road Wiscasset, Maine 04578 Westboro, Massachuse tts 01581 Stanley R. Tupper, Esq. John A. Ritsher, Es auire Tupper and Bradley Ropes & Gray 102 Townsend Avenue 225 Franklin Street Boothbay harbor, Maine 04538 Boston, Massachusetts 02110 David Santee Miller, Esq. Mr. Rufus E. Brown 213 Morgan Street, N. W. Deputy Attorney General Washington, D. C. 20001 State of Maine Augusta, Maine 04330 Mr. Paul Swetland Resident Inspector / Maine Yankee Mr. Nicholas Barth c/o U.S.N.R.C. Executive Director P. O. Box E Sheepscot Valley Conservation Wiscasset, Maine 04578 Association, Inc. P. O. Box 125 Mr. Charlbs B. Brinkman Alan, Maine 04535 Manager - Washington Nuclear Operations Combu,stion Engineering Inc. Wiscassett Public Library Association 4853 Cordell Avenue, Suite A-1 .High Street Bethesda, Maryland 20014 Wiscasset, Maine 04578 Mr. John H. Garrity, Director Mr. Torbet H. Macdonald, Jr. Nuclear Engineering & Licensing Office of Energy Resources Maine Yankee Atomic Power Company State House Station #53 Edison Drive Augusta, Maine 04333 Augusta, Maine 04336 Robert M. Lazo, Esq., Chairman Atomic Safety and Licensing Board U.S. Environmental Protection Agency U.S. Nuclear Regulatory Commission Regior 1 Office l Washington, D. C. 20555 ATTN: EIS C0ORDINATOR JFK Federal Building - Dr. Cadet H. Hand, Jr., Director Boston, Massachusetts 02203 Bodega Marine Laboratory University of California Bodega Bay, California 94923 Mr. Gustave A. Linenberger Atomic Safety and Licensing Board State Planning Officer U.S. Nuclear Regulatory Commission Executive Department Washington, D. C. 20555 189 State Street Augusta, Maine 04330 _y 7,
1 Enclosure REQUEST FOR ADDITIONAL INFORMATION 1. Geometry Geometrical description including design and as-built (when available) dimensions of the core, assembif es, shroud / baffle, thermal thield, downcomer, vessel, cavity, and surrounding shield and/or stjport 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. 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) Descrip fon of available calculations of the vessel fluence including fluenct values, locations, and corresponding r ower 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. 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. / r
a 1 . y 6. Vessel Welds Axial and azimuthal locations of vessel weld-seams with respect to the core. Overlay of current fluence m'ap with weld locations. Identify the critical welds, vertical and circunferential, 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 qualification data, or archive material. In the absence of any informt tion, assume that copper content is at its upper limit (0.35 percent. hen 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 f ailures including control system failurei and/or aperator error, steam line break and small break LOCA) which ;suid 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 gnalysis of the most limiting transient or accident with regard to vessel thermal shock con-siderations. Estimate the frequency of occurrence 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 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). l t 4 N .}}