ML20040C912
| ML20040C912 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/21/1982 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8201290366 | |
| Download: ML20040C912 (4) | |
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/BALTIMORE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203 i
000 ARTHUR E. LUNOVALL. JR.
9 V".E PRES 6DE*ef
$4 sPPLY O\\
Office of Nuclear Reactor Regulation (b
U.S. Nuclear Regulatory Commission 8
Washington, D.C. 20555 RErgpjg g
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Attn:
Mr. D. G. Eisenhut, Director
, jay g g 7g Division of L, censing
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Subject:
Calvert Cliffs Nuclear Power Plant W
7, Unit No.1, Docket No. 50-317
/b Pressurized Thermal Shock pa Gentlemen:
In your letter of December 18, 1981, you provided the results of your evaluation of our "60-day response" on pressurized thermal shock (PTS) and requested additional information. Our response is attached.
As we have had the results of scoping studies developed through our participation in the Combustion Engineering Owners Group for less than a week, we have
- not had time to complete our review. We will forward the results of our analyses by the end of this month.
BALTIM R
' ELECTRIC MPANY a
BY: A. E. Lundy 11, Jr.
Vice Presid nt - Supply STATE OF MARYLAND:
TO WIT:
CITY OF BALTIMORE :
Arthur E. Lundvall,Jr., being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a Corporation of the State of Maryland; that he executed the foregoing for the purposes therein set forth; that the statements made therein are true and correct to the best of his nowledge, informa, tion, and belief; and that he was authorized to execute the same on be alf of sa:.d r ration.
WITNESS my Hand and Notarial Seal:
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My Commission Expires:
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82012G0366 820121 DR ADOCK 05000317 h1(
_ Enclosure cc:
- 3. A. Biddison, Esquire G. F. Trowbridge, Esquire Mr. D. H. Jaff e - NRC Mr. R. E. Architzel - NRC L
Values 1.
RTNDT CEN-189, " Evaluation of Pressurized Thermal Shock Effects Due to Small a.
Break LOCA's with Loss of Feedwater for the Combustion Engineering NSSS," presented a more complete explanation of the basis for use of a lower initial RTNDT value than that allowed by MTEB 5-2. The analysis of f -20 degrees F (the the SBLOCA's was performed with an initial RTNDT 2e upper bound) for the limiting welds, however, future analyses will be based upon the more realistic methods described in the CE report. An updated material properties map for Calvert Cliffs units was included in Appendix B of CEN-189.
b.
We will perform our initial analyses using 0.30% copper concentration. We are still evaluating the available data to determine the most appropriate, realistic concentration of copper for use in PTS analysis; howeve?. for the work in progress 0.30% is an acceptable bounding value.
c.
The estimated nickel content of longitudinal welds is based upon the nickel content of the weld wire used, as described in CEN-189.
d.
Reported vessel ID fluence is the peak fluence.
2.
Rate of Increase of RTNDT No changes in core configuration which would result in increased fluence at the limiting welds are contemplated.
l Limit and Basis for the Limit 3.
RTNDT l9DT s not an appropriate limiting condition for We wish to reemphasize that RT i
operation.
We will cooperate in the effort to establish an appropriate and technically defensible criterion. Clearly, undue emphasis on a single index of material behavior (which is only one aspect of the PTS issue) does not adequately accoun for system characteristics which reduce or preclude challenge to reactor vessel integrity.
4.
Operator Actions We fully appreciate the significance of the PTS issue.
It is and has been receiving attention at the highest levels of this Company. Our Off-Site Safety Review Committee is following the issue closely to monitor developments which may be significant to safety. In addition to our participation in Combustion Engineering Owners Group (CEOG), we have retained MPR Associates and Dr.
W. E. Cooper of Teledyne Engineering Services to augment our internal technical resources.
Finally, we are cooperating with the Electric Power Research Institute (EPRI) to conduct additional realistic, plant specific analyses.
As you mention in your request for additional information, operators at Calvert Cliffs are generally knowledgeable concerning PTS. Our "60-day" submittal included excerpts from Calvert Cliffs procedures which provide guidance to the operators to prevent overpressurization. These procedures are factored into our operator training program and provide adequate assurance of reactor vessel integrity for near term operation.
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~c-Pressurized thermal shock is a complex phenomenon which ' involves many aspects of plant design and operations. We have been'and are participating in a j
j program with - the; CEOG which includes the ' review of - plant - operating procedures'for the proper inegration of specific guidance on pressurized thermal shock. Results of this review are expected to be available in the first half of-1982, at which time they will be -used to upgrade Calvert Cliffs operating =
procedures and. training programs. No specific. training lessons or procedures have yet been added to address the PTS issue, however, our current training and procedural guidonce provide adequate assurance that proper operator action wi!!-
be.taken to prevent pressurized thermal shock to the reactor Wssel in the interim.
Specifically, thermal shock is included in the " Plant Performance" section of
.the licensed operator training program.- This training includes heat-up and -
cooldown limits associated with the reactor vessel and the possible. consequence of ' exceeding these limits.
Detailed information is' presented regarding the =
types of forces expected and the effects of those forces on the reactor vessel, i
Minimum pressurization temperature (MPT) curves are reviewed with particular attention given to_ the bases of the curves and the 'effect of extended plant operation on their slope und break points.
We reiterate that our current training and procedurt.1 guidance provide adequate ' assurance that proper operator action will be taken to. prevent pressurized thermal :, hock to the.
reactor vessel in the near term.
In addition to our efforts within-the CEOG, we are working with the Electric P(wer Research Institute to perform additional plant-specific studies and to
. develop new analytical methods for addressing PTS.. As the results of this' program becon.e available, we will evaluate them for possible inclusion in our procedures and training program.
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