ML20079M705

Forwards Superheater Tube Leaks in Steam Generators of Fort St Vrain HTGR, Containing Results of Evaluations of Potential Causes for Tube Leak.Insufficient Evidence Exists to Identify Causes of Leaks

ML20079M705
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/20/1984
From:	Lee O PUBLIC SERVICE CO. OF COLORADO
To:	Jay Collins NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML20079M708	List: ML20079M705 ML20079M717
References
P-84028, TAC-54598, NUDOCS 8401270395
Download: ML20079M705 (4)

Text

__

l PUBLIC SERVICE COMPANY OF COLORADO P. O. B O X 840 DENVER, COLORADO 80208

• • ^" " '"" Jantary 20, 1984 Fort St. Vrain Unit No. 1 P-84028 Mr. John T. Collins M IM ) N 3 3 Regional Administrator  !

r egulatory Commission JAN 2 3124 bil Ryan Plaza Drive, Suite 1000 _

g Arlinoton, Te,xas 76011 DOCKET NO.: 50-267

SUBJECT:

FSV Steam Generator Tube Leaks

REFERENCE:

1) NRC Letter E.H. Johnson to

.- 0.R. Lee (G-83412) dated November 18, 1983

2) PSC Letter J.K. Fuller to James Miller (P-80064) dated March 31, 1980

, 3) LANL Letter C.H. Anderson to Robert Clark (G-82014) dated January 5, 1982

4) PSC Letter H.L Brey to Robert Clark (P-82061) dated

_ .[, March 29, 1982

Dear Mr. Collins:

Publ'ic Service Company of Colorado (PSC) shares the concern expressed in your November 18, 1983 letter regarding the possibility of a steam generator:_ tube degradation mechanism which could lead to additional tube leaks in the future. Subsequent to the tube leak in module B 3 of the loop 2 steam generator at Fort St.Vrain(FSV),PSC conducted an engineering evaluation to determine whether there may be a degradation mechanism active in the steam generators.

8401270395 940120 PDR ADOCK 05000267 P + PDR gdof

.s i\ l

c

, P-84328 Page 2 PSC and GA Technologies. Inc. (GA) evaluated potential causes for this tube leak even though the plugging of subheader M of module B 3 did not result in loss of rated capacity cr performance of the FSV plant as described in the FSAR. ,pThe result of these evaluations is reported in Attachment 1. PSC and GA concluded that there is not enough evidence to identify conclusively the mechanism that caused the tube failure. This lead to the general statement in the Reportable Occurrence that the tube leak apperrs to be random in nature.

PSC has periodically evaluated whether it is possible or practical to use h0E to detect any degradation that may be occurring in the tubes of the steam generators. Excerpts from References (2), (3) and (4),

enclosed with this letter as Attachments 2, 3, and 4, contain the results of the most recent review made by PSC and the NRC of the use of steam generator tube NDE as part of the FSV ISI program development. However, examination of the FSV superheater 2 (SH 2) tubes in the area where the two leaks were found (one in module B-2-3 and one in module B-1-1) is for all practical purposes impossible with available NDE techniques and methods. Steam generator modules would have to be removed from the PCRV for access to the tube bundle.

This would involve an extended plant outage, be prohibitively expensive, &nd result in substantial radiation exposure to personnel. '

Attachment E shows the approximate location of the two tube leaks and Attachments 6 and 7 show the complexity and congestion inside the module. Eddy current, magnetic particle or ultrasonic methods for obtaining quantitative data are not feasible on external tube surfaces because the clearances between the tubes or between the tubes and the outer shroud of the module are too small to accommodate state of the art probes or crystals, even if the tube bundle were removed from the PCRV and made accessible. Inspection of the internal surfaces of the SH 2 tubes with fiber optics, eddy current or ultrasonic methods is also impractical unless tubes on a removed tube bundle were destructively cut to reach the site of the tube leak. Remote internal use of these methods is also inhibited where the feedwater and steam subheaders are divided into three smaller tubes in the helium outlet plenum of the module.

The only areas where NDE is practical are not wholly representative of the tube leak area. These areas are external to the Prestressed Concrete Reactor Vessel (PCRV). The subheader tubes in these areas are made accessible for NDE in the process of capping the subheaders containing the leaking tube (s). Metallographic examinations of specimens taken from the feedwater and steam subheaders that were connected to the leaking tubes in modules B-1-1 (November 1977) and B-2-3 (December 1982) have been performed by GA. The results of these examinations showed very little degradation and no discernible difference in the condition of the internal surfaces of the module B-2-3 subheaders and the internal surfaces of module B-1-1 subheaders.

P-84028 , ,

Page 3 Based upon the above discussion and the evaluations previously performed by PSC and GA, PSC will perform a metallographic examination of specimens taken from the accessible subheader tubes that are connected to the inaccessible tubes which contain the leak, immediately following each future tube leak. The results of these examinations will be compared to those obtained from the specimens for the B-2-3 and the B-1-1 tube leaks. PSC will also evaluate the size and elevation of all future tube leaks to determine if additional evidence or circumstances can help to identify a cause or trend in the degradation of the tubes of the FSV steam generators.

After the occurrence of any future steam generator tube leak (s), PSC will notify the NRC of the estimated size and elevation of the leak (s) and of the results of the engineering analyses performed that may help to identify the mechanism that caused the leak (s) to occur.

PSC considers this action to be adequate at this time because:

a. The consequences of steam generator tube leaks have been analyzed in the FSAR.

b. The steam generators are designed with enough tube surface area to allow plugging of the equivalent of one feedwater/ steam subheader in each of the twelve modules without reducing the rated capacity of the Fort St. Vrain plant. There are now only two modules (B-1-1 and B-2-3) that have or.e plugged feedwater/ steam subheader.

c. Steam generator tube leaks can be detected while they are very small in size, and they can be located as in the two previous incidents.

d. Nearly five years of operation elapsed between the occurrence of the first and the second tube leaks. The plant has been generating for over six months since the second tube leak was repaired with extremely low or undetectable moisture levels in the primary coolant which directly confirms current steam generator tube integrity.

This is acceptable performance for a plant designed and licensed to accommodate a large number of steam generator tube leaks, and is not indicative of a significant problem.

E

' ' ~

P-84028 , ,

Page 4 PSC believes that the information provided herein and the above PSC ccamitments respond to your request for additional NDE following future tube leaks in the most practical and technically effective manner possible. Please direct any questions to Mr. M.H. Holmes at (303)571-8409.

Very truly yours, 8

0. R. Lee, Vice President Electric Production ORL/AHW:pa Attachments (7) i