ML20132E422
| ML20132E422 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/30/1985 |
| From: | Perlis R NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML20132E426 | List: |
| References | |
| CON-#385-044, CON-#385-44 OL, NUDOCS 8508010811 | |
| Download: ML20132E422 (10) | |
Text
6l}t) 7/3045 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKETED BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
GEORGIA POWER C0.
Docket Nos. 50-424 0FFICE OF SECRETM <
et al.
)
50-425 00CKETg ERVKf (0L)
(Vogtle Electric Generating Plant, Units 1 and 2)
NRC STAFF RESPONSE TO APPLICANTS' MOTION FOR
SUMMARY
DISPOSITION OF CONTENTION 11 (STEAM GENERATORS)
I.
INTRODUCTION On July 5, 1985, Applicants filed a Motion for Summary Disposition of Joint Intervenors' Contention 11 which alleges that Applicants have not demonstrated that no unacceptable radiation releases will occur as a 1
result of certain specified steam generator tube failures. The Staff responds to Applicants' Motion herein; for the reasons presented below and in the attached Affidavits of Jai Raj Rajan and William LeFave, the Staff submits that the Motion should be granted.
II. LEGAL STANDARDS GOVERNING
SUMMARY
DISPOSITON The Staff previously set forth the applicable legal standards governing motions for summary disposition in its July 26, 1985 " Response to Applicants' Motion for Summary Disposition of Contention 10.3 (Cables in Multiconductor Configurations)" (at pp. 1-3).
In order to avoid h $h 3se
unnecessary repetition, that discussion is incorporated by reference herein.
III. APPLICANTS' MOTION In its order of September 5,1984, the Board admitted portions of Joint Intervenors' Contention 11, and reworded that contention as follows:
Applicants have not demonstrated their basis for confidence that no unacceptable radiation releases will occur as the result of steam generator tube failures occasioned by vibration-induced fatigue cracking and by bubble collapse within the Vogtle steam generators.
LBP-84-35, 20 NRC 887, 908 (1984). The Staff will examine the issues of vibration-induced fatigue cracking and bubble collapse serfatim. For the reasons presented below, neither issue poses a problem for the safe operation of the Vogtle facility and, accordingly, the Staff supports Applicants' Motion for Summary Disposition of Contention 11.
A.
Vibration-Induced Fatigue Cracking Vibration-induced fatigue cracking is a failure mechanism in which i
excessive vibration of a flexible member such as a steam generator tube can result in a cyclic deflection or bowing of the tube between its supports. This cyclic deflection can generate stresses causing the tube walls to undergo degradation or fatigue and eventually, after a finite j
number of cycles, to crack and fail.
If vibration of the tubes is kept within certain limits, the tube walls can tolerate, without cracking, an infinite number of cycles. This limit is defined as the endurance limit I
of the tube.
If vibrations are kept below the endurance limit, i
I
e vibration-induced fatigue cracking and failure should not occur. Rajan Affidavit, 11 3-4.
The Westinghouse Model F steam generators which are installed at Vogtle are accurately described in detail in the Affidavit of Carl Hirst attached to Applicants' Motion. The NRC Staff has reviewed the Westinghouse U-tube type steam generator (including the Model F) and has determined that vibration-induced fatigue cracking should not be a problem.
This conclusion is based upon the following:
a)
Operational experience with the Westinghouse U-tube type steam generators to date has not revealed any vibration-induced fatigue cracking. This experience, which includes Model F steam generators, isdocumentedinNUREG-0886("SteamGeneratorTubeExperience",
Feb.1982) and NUREG-0606 (" Unresolved Safety Issues Summary",
Aug. 17, 1984).
b)
The NRC Staff has reviewed the analytical results and model test data obtained by Westinghou*se to evaluate the secondary flow in the Westinghouse Model F steam generators. This review indicates that the vibratory stresses in the tubes in the various regions of the steam generator are likely to be well under the endurance limit. The Westinghouse evaluation included the parallel and cross flow excitation and the three vibration mechanisms: vortex shedding, fluid-elastic excitation and turbulence. The results indicate that vibration-induced fatigue cracking should not be a concern with Model F steam generators.
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c)
The first operating plant with Model F steam generators was instrumented to monitor vibration of the tubes during actual operation. The vibration data under plant operating conditions revealed no excessive vibration and were consistent with the analytical and model test results.
Rajan Affidavit, 11 5-9.
Based on the Staff's evaluation of the analytical and model test results, as well as the operational data generated for the Model F generator, the Staff has concluded that vibration-induced fatigue cracking is highly unlikely to occur in Westinghouse Model F steam generators. J_d.,110.
Further, while the Staff does not expect vibration-induced fatigue cracking to cccur in Model F generators, even if a crack were to occur, no appreciable releases of radiation should result.
The Vogtle technical specifications provide an additional level of protection, by specifying permissible primary-to-secondary system leakage limits for the steam generators. Any time a leakage rate is exceeded, the plant must shut down and appropriate remedial action must be initiated.
These limits are set at such a low level that even in the unlikely event a fatigue-induced crack does propagate through a wall and cause a leak, the leak should be detected, the plant shut down, and remedial measures initiated before any appreciable amounts of radiation could be released due to the leak.
Rajan Affidavit, 1 11.
Joint Intervenors have also raised a concern about fretting.
Fretting wear of steam generator tubes can result from vibratory impacts or rubbing between the tubes and other metal parts such as support plates. Fretting is a distinct issue from vibration-induced fatigue
r cracking, and as such is not properly raised in the Contention.
In any event, fretting is not a matter of concern for Model F generators.
Fretting wear was observed in the preheat region of Westinghouse Model D stea:n generators. The problem was caused by a high level of flow turbulence in the preheat region. The Model F generator has no preheat region and has been designed to preclude the possibility of fretting wear by reducing the level of flow turbulence. The Staff has reviewed the Model F design and agrees that fretting should not be a concern for Model F generators. Rajan Affidavit, f 12. Fretting is thus neither part of the admitted contention nor should it pose a problem at the VogtIc facility.
In conclusion, analyses, tests and operational evidence indicate that the Model F generator used at Vogtle should not experience vibration-induced fatigue cracking. Even in the unlikely event a crack should occur, the facility's technical specifications should lead to detection of the existence of the crack and plant shutdown before the public health and safety is adversely.affected. For these reasons, summary disposition of that portion of Contention 11 that deals with vibration-induced fatigue failure is appropriate.
B.
Bubble Collapse Bubble collapse is a type of water hamer occurrence induced by condensation of a steam pocket in the steam generator feeding or the feedwater piping leading to the steam generator. Bubble collapse water hammer was one of several types of water hammer evaluated by the NRC Staff inresolvingUnresolvedSafetyIssue(USI)A-1.
In March 1984, the Staff published its re olution of USI A-1, in NUREG-0927 (" Evaluation of Water
Hammer Occurrence in Nuclear Power Plants - Technical Findings Relevant to Unresolved Safety Issue A-1") and NUREG-0993 (" Regulatory Analysis for USI A-1, ' Water Hammer'"). LeFave Affidavit, 1 4.
One of the results of the resolution of USI A-1 was adoption of the design considerations and preoperational test recommendations described in Auxiliary Systems Branch (ASB) Branch Technical Position (BTP) 10-2 (a copy of which is attached to the LeFaye Affidavit). The Staff has determined that compliance with BTP ASB 10-2 provides an acceptable means of protection against water hammer events. LeFave Affidavit, 1 5.
Applicants have followed the guidance of BTP ASB 10-2 in the design of the Vogtle steam generators.
In addition, the Applicants have committed to perform preoperational testing to verify that the unit operating procedures do not result in steam bubble collapse water hammer in the main or auxiliary feedwater system. LeFave Affidavit,16.
The potential for bubble collapse water hammer in the feedring of the Model F steam generator used at Vogtle has been minimized by the use of J-tubes and a welded thermal sleeve (to prevent draining the feedring when no flow is present), thereby preventing steam bubble formation.
In addition, by introducing auxiliary feedwater (AFW) flow into a separate steam generator nozzle in lieu of the feedring, the colder auxiliary feedwater will not contact the warmer feedwater in the feedring or any steam in the feedring, thereby minimizing the possibility of condensation-induced bubble collapse (water hammer). The probability of steam bubble collapse water hammer has been significantly reduced in the Vogtle design by these measures. Jd.,116-7.
The Vogtle design also follows the recomendations of BTP ASP 10-2 regarding the shortest possible horizontal run of inlet piping to the steam generator feedring and auxiliary feedwater nozzle, to prevent highpoint steam pockets. During normal plant operation, the AFW inlet piping is kept full of water by the main feedwater system via bypass piping. The discharge end of the inlet piping in the steam generator is below the normal operating water level. Additionally, four check valves in series are provided between the AFW inlet nozzle and the AFW pump recirculation lines to minimize the possibility of backleakage under no flow conditions. Thus, steam bubble formation in the AFW inlet nozzle and piping is not expected to occur. LeFave Affidavit, 1 8.
In conclusion, the Staff has taken the position that the generic water hamer " unresolved safety issue" is now resolved and that, if ASB BTP 10-2 is followed, the public health and safety will not be endangered by steam generator tube degradation due to bubble collapse water hammer.
The Staff has reviewed the Vogtle design. features intended to minimize the potential for bubble collapse water hammer. As indicated in Section 10.4.7 of the Vogtle Safety Evaluation Report, the Staff has concluded that the Vogtle design meets the Staff's criteria with respect to the prevention of bubble collapse water hamer as set forth in ASB BTP 10-2, and that the design is acceptable. LeFave Affidavit, 11 5,9.
For these reasons, the Staff has conluded that steam bubble collapse water hamer does not present a concern for the Vogtle facility, and therefore supports Applicants' Motion for Sumary Disposition of Contention 11 as it relates to this issue.
IV. CONCLUSION For the reasons presented above, the Staff submits that Applicants' Motion for Sumary Disposition of Contention 11 should be granted.
Respectfully submitted, 1
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Robert G. Perlis Counsel for NRC Staff Dated at Bethesda, Maryland this 30th day of July,1985 i
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i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 00CKETED U5hRC BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 1f3 AUG -1 A10 :49 In the Matter of
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GEORGIA POWER C0.
Docket Nos. 50-424 ObbkfTf G.i[
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et al.
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50-425 BRAsca
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(0L)
(Vogtle Electric Generating Plant, )
Units 1 and 2)
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CERTIFICATE OF SERVICE 1
I hereby certify that copies of "NRC STAFF RESPONSE TO APPLICANTS' MOTION FOR
SUMMARY
DISPOSITION OF CONTENTION 11 (STEAM GENERATORS)" and Supporting Documents in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commission's internal mail system, this 30th day of July,1985.
Morton B. Margulies Esq., Chairman
- Mr. Gustave A. Linenberger, Jr.*
Administrative Judge Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board Panel Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Washington, D.C.
20555 Dr. Oscar H. Paris
- Bradley Jones, Esq.
Administrative Judge Region 1 Counsel Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission i
Panel Suite 3100 U.S. Nuclear Regulatory Commission 101 Marietta Street Washir.gton, D.C.
20555 Atlanta, GA 30303 i
i Bruce W. Churchill, Esq.
Douglas C. Teper David R. Lewis, Esq.
1253 Lenox Circle Shaw, Pittman, Potts & Trowbridge Atlanta, GA 30306 1800 M Street, N.W.
Washington, D.C.
20036
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Atomic Safety and Licensing Laurie Fowler, Esq.
Board Panel
- 218 Flora Ave. NE U.S. Nuclear Regulatory Commission Atlanta, GA 30307 Washington, D.C.
20555
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Docketing and Service Section*
Atomic Safety and Licensing j
Office of the Secretary Appeal Board Panel
- U.S. Nuclear Regulartory Commission U.S. Nuclear Regulatory Commission i
Washington, D.C.
20555 Washington, D.C.
20555 James E. Joiner, Esq.
Ruble A. Thomas Troutman, Sanders, Lockerman, Southern Company Services, Inc.
2
& Ashmore P.O. Box 2625 127 Peachtree Street, N.E.
Birmingham, AL 35202 Candler Building, Suite 1400 Atlanta, GA 30043 4
l Tim Johnson Executive Director i
Campaign for a Prosperous Georgia j
175 Trinity Avenue, S.W.
i Atlanta, GA 30303
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Robert G. Perlis t
Counsel for NRC Staff i
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