ML19290E341
| ML19290E341 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 03/04/1980 |
| From: | Groce R Maine Yankee |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| WMY-80-38, NUDOCS 8003110001 | |
| Download: ML19290E341 (4) | |
Text
e
, MA/nE
- An#EEf '
f J.: I '
sun N,,x E n 0 0,n1.,,
ENGINEERINC OFFICE WE STBOR O. M ASSACH US E TTS O' E 81
$ {~
j 617-366-9011 y
a s
B.3.2.1 mn* 80-38 March 4, 1980 United States Nuclear Regulatory Ccemission Was hington, D.C.
20555 Attention: Office of Nuclear Reactor Regulation
References:
(a)
License No. DPR-36 (Docket No. 50-309).
(b)
" Summary of Meeting with MYAPC on October 18 to Discuss Their Asymmetric IDCA Loads Evaluation in Bethesda, Maryland" from S. B. Hos ford to L. S. Shao.
(c) MYAPC Letter to USNRC, WMY 79-107, dated 9/27/79.
(d) MYAPC Letter to USNRC, FMY 80-1, dated 1/7/80.
(e)
USNRC Letter to PWR Licensees dated 1/25/78.
Subject:
Resolution of Asymmetric IDCA Loads
Dear Sir:
On February 7,1980 the NRC staf f and EG&G Idaho met with technical staff members of Yankee Atom'c Electric Company Nuclear Services Division to discuss the resolution of the Asymmetric IDCA Ioad Issue at Maine Yankee. As requested at that meeting and subsequently reiterated via telecon (February 12, 1980), this letter is written to outline Maine Yankee's current status regarding evaluation of asymmetric IDCA loads.
Maine Yankee is the first operating Combustion Engineering plant in the United States to have installed retrofit pipe rupture restraints specifically designed to limit pipe break blow $own area and thereby greatly diminish maximum asymmetric loads on the primary support system due to a hypothetical Loss of Coolant Accident (IDCA).
Taking credit for the load reduction afforded by pipe rupture restraints, Maine Yankee has quantitatively characterized maximum loadings on affected Safety Class I components, supports and structures.
In direct combination with normal operating and previously considered transient loadings, maximum stresses due to asymmetric LOCA, including thrust load, are within original, conservative FSAR design limits. Therefore, the structural integrity of the reactor coolant system (RCS), supports and associated structures is not comprcznised due to additional asymmetric loads imposed by a hypothetical IDCA, and the original plant licensing bases are not violated, fbOf
U.S. Nuclear Regulatory Commission March 4, 1980 Attention Office of Nuclear Reactor Regulation Page 2 Pipe Rupture Res traints As noted in References (b) and (c), Maine Yankee elected to install pipe rupture restraints to limit the pipe break area resulting from a hypothetical rupture of reactor coolant system piping.
Maine Yankee continues to believe that given an extensive, expensive. state-of-the-art non-linear elastic /plas tic, detailed dynamic analysis eventual system adequacy and the LOCA-mitiga ting ability of the RCS and ECCS could ultimately be shown without added res traints. However, this approach has not considered cost ef fective, nor could it have been completed in the required time frame.
Pipe rupture restraints have been installed at each of six nozzle locations, outside the primary shield wall surrounding the reactor pressure vessel and inside piping penetrations through the primary shield wall. The function of the restraints is to limit pipe motion directed transverse to the axial direction of the ruptured pipe.
In conjunction with pipe motion restraint afforded by existing component supports, total blowdown areas approximately equal to those of currently licensed plant designs are achieved.
(Reactor coolant pumps, steam generators and their supports at Maine Yankee were originally designed to withstand LOCA thrus t loads together with normal operating and SSE seismic loads, maintaining elastic stress levels).
The pipe rupture restraints are designed to accommodate dynamic loads applied through a gapped interface while maintaining elastic stress levels.
Therefore, during normal operating conditions and transients other than pipe rupture, there is no effect on reactor coolant system piping due to restraint ins talla tion. Design loads transmitted through the pipe rupture restraints into existing subcompartment walls were quantified and the integrity of existing structures was analytically assured.
Potential adverse effects on existing concrete walls due to the restraints' becoming a thermal conduction path, and loss of ability to perform inservice inspection due to pipe rupture restraint installation were identified as critical design constraints. The thermal-conduction design is such that concrete overheating will not occur. The pipe rupture restraint design has been reviewed by the Yankee Atomic Electric Company ISI Coordina tor. Pipe rupture restraint installation does not adversely affect the ability to perform inservice inspection.
Structural Evaluation Report A preliminary draft of " Evaluation of LOCA-Related Loadings on the Reactor Coolant System Components, Supports and Piping at Maine Yankee",
Enclosure (A), is provided for your information in order to assist in expediting review of the final report. This version, although preliminary, contains detailed descriptions of methodologies and analytical models used to define components of the asymmetric LOCA loads and their application to the structural model for subsequent analysis. Evaluation criteria and techniques for assessing the effects of total applied loads are described for each reactor coolant system component support and for ECCS and branch piping.
Detailed analyses are undergoing interns 1 review in accordance with Yankee Atomic Electric Company Operational Quality Assurance procedures, and therefore are not presented at this time. However, results of analyses are
U.S. Nuclear Regulatory Commission March 4, 1980 Attention Office of Nuclear Reactor Regulation Page 3 available to the extent that applied loading and resulting safety margin magnitudes are qualitatively assessed. Therefore, engineering judgement with a high degree of confidence is used in concluding structural integrity.
Final report submission is tentatively scheduled for the end of April,1980.
Evaluation As previously discussed, (Reference (b) and (d)), original design analysis methods and results, together with determination of applicability to Maine Yankee of results obtained for similar plants subjected to identical loading conditions, were to be used extensively in conservatively developing a basis for concluding system structural adequacy at Maine Yankee. This approach was success fully utilized to determine that small additional subcompartment asymmetrical pressurization LOCA-related loads resulting from postulated breaks outside the primary shield wall do not compromise components ', supports ' or subcompartment halls ' structural integrity.
Load definition for reactor coolant system components and supports subjected to loadings associated with postulated pipe rupture inside the reactor cavity was accomplished using more detailed analytical investigations.
Classical thermal hydraulic and structural analysis techniques were employed to conservatively predict upper bound RCS component support reaction loads and transmitted forcing functions applied to remaining, unbroken RC3 loops. Analytical models and analysis methodology, described in Enclosure (A), have been presented to the NRC staff and EG&G Idaho at a recent meeting held at the Maine Yankee plant site.
Preliminary results of bounding, dynamic analyses indicated that total RPV support loads due to normal operating, maximum seismic and total asymmetric LOCA loads are within original FSAR limits. This is attributed to the load reduction due to the installed pipe rupture restraints as well as an indication of the high degree of conservatism employed in the original design analyses.
Further, reactor internals loadings are substantially less than those previously considered in detailed dynamic structural analyses performed for Maine Yankee. There is no increase in reactor pressure vessel displacements relative to those resulting from original design LOCA loads. Therefore, the magnitude of loadings applied to ECCS and RCS branch piping; and control element assemblies due to reactor motion is unchanged.
Conclus ion Maine Yankee has evaluated the consequences of a hypothetical reactor coolant pipe rupture as requested in Reference (e).
Based upon the calculations and studies performed, Maine Yankee concludes that:
1.
no loss in structural integrity of essential components and supports due to the accident considered is anticipated, 2.
the ability of safety systems to mitigate the consequences of the accident under consideration as defined in the Maine Yankee FSAR remains unchanged.
U.S. Nuclear Regulatory Commiss ion March 4, 1980 Attention Office of Nuclear Reactor Regulation Page 4 6
Maine Yankee concludes that plant safety has been adequately demonstrated and cons iders the issues identified in Reference (e) satis factorily resolved.
Should you have any questions, or wish to discuss the contents of this letter, please feel free to call.
Very truly yours, MAINE YANKEE ATOMIC POWER COMPANY b
obert H. Groce Senior Engineer - Licensing AVR/kaf