ML20215N993
| ML20215N993 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/31/1986 |
| From: | Gridley R TENNESSEE VALLEY AUTHORITY |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8611100190 | |
| Download: ML20215N993 (5) | |
Text
--
eh TENNESSEE VALLEY AUTHORITY SN 157B Lookout Place OCT 311986 Director of Nuclear Reactor Regulation Attention:
Mr. B. Youngblood, Project Director.
PWR Project Directorate No. 4 Division of Pressurized Water Reactors (PWR)
Licensing A U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Youngblood:
In the Matter of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT - ANCHOR POINT MOVEMENT Your. letter dated September 29, 1986, requested clarification of TVA's criteria for anchor point inovement (APM) loads associated with the design basis accident condition (double-ended guillotine break in the reactor coolant loop).
Enclosed is TVA's explanation of this criteria difference.
If you have any questions, please call M. R. Harding at 615/870-6422.
Very truly yours, TENNESSEE VALLEY AUTHORITY R. dr'idley, Rgirector Nuclear Safety and Licensing Enclosure cc: See page 2
(
hhk DOC 00 7
I oO P
l I, I i
An Equal Opportunity Employer 7 yy,,,,, -
..~,------w,
,, - - - -, -,, - - -..,, - -, - ~,
,,,y
,w e,
.-w-
- Director of Nuclear Reactor Regulation OCT 31886 cc (Enclosure):
U.S. Nuclear Regulatory Connaission Region II Attn:
Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. James Taylor, Director office of Inspection and Enforcement U.S. Nuclear. Regulatory Commission Washington, D.C. 20555 Mr. Carl Stahle Sequoyah
-oject Manager U.S. Nuc4 at Regulatory Commission
~1920 Norfolk Avenue Bethesda, Maryland 20814 Mr. G. G. Zech Director, TVA Projects U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 t
S ENCLOSURE Ouestion In the course of the staff inspection held the week of July 7-10, 1986, the team found that anchor point movement (APM) loads associated with a design basis accident (double-ended guillotine break in the reactor coolant loop) were included in the Watts Bar Design Criteria WB-DC-40-31.9 (previously used for the design of Sequoyah supports) but were not included in the new Sequoyah Design Criteria SQN-DC-V-24.1.
The TVA response to the team's observation was that design basis accident (DBA) APM loads were not a licensing issue at Sequoyah. Primary system movements at branch lines resulting from the DBA loading were found to be less than 1/4 inch. Because the APM results in a secondary stress in the attached branch lines in a plant faulted condition, TVA concluded that the ASME Code does not require its evaluation. According to the licensee, this was a Westinghouse-NRC negotiated position.
Furthermore, according to TVA this load case does not occur under the leak-before-break position recently approved by the staff.
The initial staff review of this issue finds no documentation of the Westinghouse-NRC negotiated position. Although secondary stresses in the piping due to plant faulted conditions (e.g. SSE loading) are generally neglected because they are not expected to cause gross structural failure due to local yielding and minor distortions in the pipe TVA should provide the documentation of the basis for concluding that the APMs associated with a DBA event are sufficiently small to produce secondary, self-limiting type stresses.
In addition, TVA should provide written documentation of the Westinghouse-NRC negotiation position identified previously.
RESPONSE
The question was formulated from information received by the NRC staff during the week of July 7, 1986, and two subsequent phone calls. As stated, the question does not accurately reflect TVA's position.
Therefore, the response will first attempt to clarify TVA's position and then to respond to the NRC concern.
Westinghouse postulated breaks in the Sequoyah Nuclear Plant (SQN) reactor coolant loop (RCL) system piping and designed pipe whip restraints to. minimize the effects of this very unlikely event. To minimize movement of the RCL i
system piping, the pipe whip restraints are shimmed to zero or near zero gap in'the hot condition. The effects of a postulated design basis accident (DBA) on RCL supports, pressurizer and steam generator supports, etc., were evaluated. Anchor point movement (APM) of pipe attached to the RCL system is limited by deformation of the RCL system pipe whip restraints since they are l
shimmed with near zero gap. As will be shown later in this response, these movements are small and result in minor distortions of the attached pipe.
d NRC's evaluation and conclusions of the TVA LOCA evaluation are provided in.
section 3.9.1 at pages 3-18 and 3-19 and in section 6.2.1 at pages 6-10 of the SQN SER, and in section 6.2.1 and Appendix C page C-8 of Supplement 1 to the SER. The code of record for SQN piping is ANSI B31.1, 1967. No direct reference is made to evaluating branch lines off the RCL system for LOCA APM in the FSAR, the SER, or the piping' code of record.
The Watts Bar Nuclear Plant (WBN) is similarly designed and evaluated for LOCA effects. However, the WBN evaluation was expanded to include consideration of LOCA APM effacts on~ reactor coolant loop branch lines.
Review of correspondence between Westinghouse and TVA in the 1977 timeframe indicates Westinghouse furnished LOCA APM for WBN.
They further indicated that part of the APM was due to inertial effects and expressed their position that it would be consistent with ASME terminology to conservatively evaluate the entire APM as a primary stress in the faulted condition. We have not found documented records of negotiations between NRC and Westinghouse on evaluation of the WBN primary system branch lines for LOCA APMs. Westinghouse represented TVA in qualifying the Westinghouse RCL, and TVA believed NRC was aware of and concurred in the analysis approach.
The load combinations and allowable stress for the evaluation are documented in Tables 3.9-7 and 3.9-3 of the FSAR.
The RCL for SQN and WBN plants are similar in physical layout, support scheme, postulated pipe rupture locations, and pipe rupture restraints. Therefore, the effects of a LOCA'at SQN will be similar to those established for WBN.
APM at branch line locations in the WBN plant analysis approach 0.5 inches in the broken loop and 0.25 inches in the unbroken loop.
The effects of LOCA and APM on various branch lines attached to the RCL at the WBN plant were evaluated by computer analysis. As indicated in Tables 3.9-7 and 3.9-8 of the FSAR, the resulting stresses were considered to be primary and combined with other faulted condition primary loads.
This evaluation was performed after the RCL branch lines were designed for all other faulted load
,i sources and did not result in any modifications.
It can, therefore, be concluded that SQN RCL branch lines movements associated with a DBA event are sufficiently small.
Furthermore, the instantaneous reactor coolant loop circumferential type breaks will not occur under the leak-before-break (LBB) position recently approved by the NRC staff. TVA has not taken any credit for the LBB in the initial SQN licensing or the new SQN Design Criteria SQN-DC-V-24.1.
l In summary, LOCA APM loads were not evaluated at SQN, given the vintage of this power station. Analysis methods were state-of-the-art for the time period involved and in compliance with applienble codes and licensing bases.
Similar plants of SQN's vintage also did not require evaluation to this loading condition.
?
g ap A
A 3
L
.m
. LOCA APM loads were, however, included in later vintage plant analyses, including sister plant WBN (as a backfit). Even as a backfit to WBN, no physical modifications were required. Consequently, none would be indicated for SQN since RCL arrangement, supports, and pipe rupture restraints are essentially identical.
LBB has already been demonstrated for similar plants with the similar materials and loadings. This, coupled with the fact that TVA plans to implement LBB, provides additional justification for the original licensing position.
et, e
.