ML20030D653

Requests Addl Info Re Pressurized Thermal Shock to Reactor Pressure Vessels,Per Review of PWR Owners Group 810515 & Licensees 810522 Responses to NRC

ML20030D653
Person / Time
Site:	Calvert Cliffs
Issue date:	08/21/1981
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
References
NUDOCS 8109140199
Download: ML20030D653 (8)

Text

C AUEUST C 1 1531

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lir. A. E. Lundvall

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Vice President, Supply j

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Baltimore Gas 8 Electric Company C ggcgu"$s 4

P. G. Box 1475

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SUBJECT:

PRESSURIZED THEPML SOCK TO REACTOR PRESSURE VESSELS 1 e have revicued the PVP Owners' Grouus responses of liny 15, 1991 and the licensees' responsas of 4ay 27,1931 to our letter dated April 20, 1981 concerning the sub,tect issue. The EPRI work which bears on the issue was included in the licensees' responses. On toe basis of our independent revleu, of the plants where neutron irradiatinn has significantly reduced the fracture toughness of tbt reactor pressure vessels (RPVs), all plants could survive a severe overcooling evant for at least another year of full f

powar operation. However, we believ9 that additional action should be el

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taken now to resolve the long-term problens.

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This belief is based upon our analyses which indicate that reductions in fracture toughness for sore RPVs are approaching levels of concern.

It is also based in part on the fact th3t any proposed corrective action rust allos adequate lead tir.e for planning, review, approval, procurepent and installation. These conclusions were recently discussed with the Pim Owners Grouns on July 28-30, 1981 At those rvntings, the Owners Croups rr_vioued the prograas underaay at the three PW vendor-s which are designed to scope the naanitude an.1 applicability of the generic problea and to be i

T conpleted by late 19R1. The three programs appeared to contain the necessary 9

ele ents for resolution cf the oroblem on a generic basis and the imC plans W

to nake full use of the reports due by the end of the year. While the 88 vendors and Duners Groups are to be comended and encouraged in addressing the gennric issue, there is also a need for plant-specific infornation for g

your plant.

u(D 85 Dased on current vessel reference tenperature and/or systen characteristics, g

we have identified Ft. Calhoun, Robinson 2, San Onofre 1, Maine Yankee, c-Oconee 1. Turkey Point 4, Calvert Cliffs 1 and Three Mile Island 1 as nlants from which we require additional infomation at this tim.

The staff has used the tine-dependent pressure and te perature data from the March 20, 1078 Rancho Seco transient as a starting point for our evaluation of this issue because:

(1) it is the rvst severe overcooling event experienced to date in an operatinq plant; (2) it is a real, as l

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opposed to a postulated, event; and (3) it was severe enough that it could l

challenge the RPV when combined with physically reasonable values of ir-radiated fracture touchness and initial crack size.

In future reviews the f

staff plans to use the steam line break accident or other appropriate i

transient / accident in order to estinate nintrun operational times available before plant modifications are required.

l tising calculated RFV steel rechanical properties, credible initial flaw sizes, reasonable thermal-hydraulic paranaters, and a simplified pressure-temperature transient sinilar to that observed during the Rancho Seco event, the staff has ccncluded 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 l

for future actions. the margins in time for long tem operation are not large, and there ii considerable uncertainty in the probability that similar or more severe transients may occur.

It is clear that positive action must be initiated snon for those plants with significantly high transition temperatures. As indicated above, several such plants have been selected I

e' by the staff, based on estimates of the current reference teraperature for 1

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 recognition that inplementation of any proposed fixes or remedial actions must allow for adequate lead time. Because long-term solutions may require l

a year or more, you should explore short-term approaches as well. Although i

clear, concise instructions should be provided to operators to reduce the i

likelihood of repressurization during overcooling transiLnts, the HRC staff believes that reliance on operator actions to prevent repressurization during an overcooling transient will be very difficult to justify as an L

acceptable long-tem solution to the problem.

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In accordance with 10 CFR 50.54(f) of the Conmission's regulations, you are

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requested to submit written statements, signed under oath or affimation, to enable the Connission to determine whether or not your license should be nodi-fled, suspended or revoked. Specifically, you are requested to submit the following infomation to the NRC within 60 days frcm the date of this letter:

(1) Frovide 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 l

(b) current conditions (include fluence icvel) at i

the RPV inside carbon steel surface, l~

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Mr. A. E. Lundvall (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 raterial 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 thernal 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 sunnary should be the specific pressure, temperature 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 feedwater presently used at your facility. For each required operator action, give the information available to the operator and the time available for his decision and the aquired action. State how each required operator action is incorporated in plant operating procedures and in training I

and requalification training programs.

f You are also requested to submit a plan for Calvert Cliffs Unit No.1 to the NRC within 150 days of the date of this letter that will define actions and schedules for resolution of t

1ssue and analyses supporting continued j

operation. We request that you in.lude consideration and evaluation of the following possible actions:

(1) reduction of further neutron radiation damage at the beltline by replacement of outer fuel assemblies with dumy assemblies or other fuel management changes; (2) reduction of the thernal shock severity by increasing the ECC water tenperature; 3

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(3) recovery of RPV toughness by in-place annealing (include the basis

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for demonstrating that your plant meets the requirerents in 10 CFR 50 Appendix G IV C);

(4) design of a control system to mitigate the initial themal shock and contrbi repressurization.

For these, as well as for any other alternative approaches, provide implementation schedules that would assure continuance of adequate safety nargins.

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 infomation.

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e fir. A. E. Lundvall Due to the nature of this review, and tr.e past review effort that.has been expended, we consider the above schedules to be reasonable; however, infom us within 30 days if you anticipate conflicts with previous cornitments with l'

either subrtittal and a basis for any delay. We also expect participation *

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by the appropriate PilR Owners Group and NSSS vendors in developing solutions to the problem.

Sincerely, Original signed by Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation

Enclosure:

Request for Additiona'i Infomation cc w/ enclosure:

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omay OF'FICIAL RECORD COPY usa m imi u-uo WC FORM 318 (10-80) NRCM 024o

DISTRIBUTION:

Docket File KKniel NRC PDR NAnderson L PDR RJohnson TERA JClifford flSIC -

BSheron ORB #5 Rdg DCrutchfield ORB #4 Rdg EThrom ORB #1 Rdg WHazelton ORD#3 Rdg RKlecker Gray Files (8)

Glainas B&W Owners Group NRandall CE Owners Group Jftartore Westinghouse Owners Group TNovak llDerton JStolz DEisenhut SVarga RVollmer RAClark Silanauer CTrammell RMattson DNeighbors Tliurley SNowicki RSnaider FSchroeder RJacobs 1

OELD (S)

HSilver AEOD flGrotenhuis IE (7)

PWagner ACRS (10)

BRequa ORAB DJaffe SEPB GYissing 11 Conner RIngran Cilarwood PKreutzer RDiggs Ellylton NHughes CParrish PWoolley HSmith omce >

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nac rwu m oow nneu ono OFFICIAL RECORD COPY uso m a i_3y m

Saltiacre Gas and Electric Company cc:

James A. Biddison, Jr.

Ms. Mary Harrison, President General Counsel Calvert County Board of County Commissioners Baltimore Gas and Electric Company Prince Frederick, MD 20768 P. O. Box 1475 Baltimore, MD 21203 U. S. Environmental Protection Agency Region III Office George F. Trowbridge, Esquire Attn:

EIS Coordinator Shaw, Pittman, Potts and Trowbridge Curtis Building " Sixth Floor) 1800 M Street, N. W.

Sixth and Malnut Streets Washington, D. C.

20036 Philadelphia, PA 19106 Mr. R. C. L. Olson, Prf rchal Engineer Mr. Ralph E. Architzel Nuclear Licensing Analy;is Unit Resident Reactor Inspector Baltimore Gas and Electric Company NRC Inspection and Enforcement Room 922 - G8E Building P. O. Bos 437 P. O. Box 1475 Lusby, MD 20657 Baltimore, MD 21203 Mr. Charles B. Brinkman Mr. Leon B. Russell Manager - Washington Nuclear Operations Plant Superintendent Combustion Engineering, Inc.

Calvert Cliffs Nuclear Power Plant 4853 Cordell Avenue, Suite A-1 Maryland Routes 2 & 4 Bethesda, MD 20014 Lusby, MD 20657 Mr. J. A. Tiennan, Manager Bechtel Power Corporation Nuclear Power Department Attn: Mr. J. C. Judd Calvert Cliffs Nuclear Power Plant Chief Nuclear Engineer Maryland Routes 2 & 4 15740 Shady Grove Road Lusby, MD 20657 Gaithersburg, MD 20760 Mr. W. J. Lippold, Supervisor Combustion Engineering, Inc.

Nuclear Fuel Management Attn: Mr. P. W. Kruse, Manager Baltimore Gas and Electric Company Engineering Services Calvert Cliffs Nuclear Power Plant P. O. Box 500 P. O. Box 1475 Windsor, CT 06095 Baltimore, Maryland 21203 Public Document Room Mr. R. E. Denton, General Supervisor Calvert County Library Training & Technical Services Prince Frederick, MD 20678 Calvert Cliffs Nuclear Power Plant Maryland Routes 2 & 4 Director, Department of State Planning Lusby, MD 20657 301 West Preston Street Baltimore, MD 21201 Mr. R. M. Douglass, Manager Quality Assurance Department Administratori Power Plant Siting Program j

Fort Smallwood Road Complex Energy and Coastal Zone Administration P. O. Box 1475 Department of Natural Resources Baltimore, MD 21203 Tawes State Office Building Annapolis, MD 21204 l

Mr. T. L. Syndor, General Supervisor 1

Operations Quality Assurance Calvert Cliffs Nuclear Power Plant Maryland Routes 2 & 4 Lusby, MD 20657

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.g Enclosure REQUEST FOR ADDITIONAL INFORMATION 1.

Geometry Geometrical description includf rg 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.

3.

Neutron Source Present and expected E0L:

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, s

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 dimensfors 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 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 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 f ailures and/or.cperator error, steam line bre(- and small break LOCA) which could lead to inside vessel fluid

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temperatu. es of 300 F or lower.

Provide any Failure Modes and Effects Analyses (FMEAs) of control systems currently available or reference any such enalyses already submitted. Provide the analysis 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 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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