ML19320B603
| ML19320B603 | |
| Person / Time | |
|---|---|
| Site: | Bailly |
| Issue date: | 05/12/1980 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Shorb E NORTHERN INDIANA PUBLIC SERVICE CO. |
| References | |
| NUDOCS 8007140508 | |
| Download: ML19320B603 (1) | |
Text
.
[ga arro,o, UNITED STATES 8 ' 3,,7 ( g NUCLEAR REGULATORY COMMISSION TIT',
y-
- -' /. r REGION lil I
g
[
799 ROOSEVELT ROAD
'\\#
GLEN ELLYN, ILLINOIS 60137 e
- ...+
12 &
Docket No. 50-367 Northern Indiana Public Service Company ATTN:
Mr. Eugene M. Shorb Senior Vice President 5265 Hohman Avenue Hammond, IN 46325 Gentlemen:
The enclosed IE Bulletin No. 80-13 is forwarded to you for information.
No written response is required.
If you desire additional information re-garding this matter, please contact this office.
Sincerely, SYf' ames G. Keppler$ ^ -
Director
Enclosure:
IE Bulletin No. 80-13 cc w/ enc 1:
Central Files Director, NRR/DPM Director, NRR/ DOR PDR Local PDR NSIC TIC Mr. Dean Hansell, Office of Assistant Attorney General 8037140 %
i
O SSINS No.:
6820 Accession No.:
UNITED STATES 8002280661 NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT J
WASHINGTON, D.C.
20555 May 12, 1980 IE Bulletin No. 80-13 CRACKING IN CORE SPRAY SPARGERS Description of Circumstances:
Instances of cracking in core spray spargers have occurred at two BWR facilities.
This trend indicates a need for more intensive inspection of these components
-during subsequent refueling outages.
4 Oyster Creek Nuclear Generating Station 4
j Jersey Central Power and Light Company notified the NRC on October 18, 1978, that a crack had been found in Core Spray Sparger System II during remote visual inservice inspection at their Oyster Creek Nuclear Generating Station.
The crack was located at 208 azimuth and extended at least 180 circumferen-tially around the sparger.
An evaluation of the event by the licensee postu-lated that deformation of the sparger had occurred during fabrication and installation which led to cracking by Intergranular Stress Corrosion Cracking (IGSCC) during service in the BWR environment.
A temporary repair was effected by installing a clamp assembly over the crack.
The licensee's analysis indicated that the crack had relieved the stresses present and therefore precluded further cracking.
The NRC safety evaluation permitted operation until the next refueling outage and reouired inspection of the sparger at that time.
The'NRC was informed by the Jersey Central Power and Light Company on January 16, 1980 that further cracking was discovered in the core spray
.y spargers during an inservice inspection conducted in conjunction with the refueling outage.
A total of twenty-eight cracks 0.001 to 0.002 inches in width and of varying lengths were identified in both core spray spargers.
The licensee stated that they believed the majority of additional cracks were present earlier and not discovered during the 1978 inspection due to inspection equipment limitations.
Near term repair consisted of the application of nine
-additional clamp assemblies in areas of the spargers where cracks were visually observed on the eccessible portion of the sparger and UT indications were present in the ineccessible portion of the sparger and in the junction box
'egion. The licensee analyzed the. flow characteristics of the spargers and determined that adequate flow distribution would be maintained if thru wall cracking.005 inches wide and 180 in length were present.
The licensee stated that the installation of the clamps m.ld assure the sparger would maintain its physical integrity and remain in place.
IE Bulltin No. 80-13 May 12, 1980 Page 2 of 3 The repair measures proposed were determined by the NRC to be adequate until the following refueling outage.
The NRC evaluation stated that actions should be taken to develop and install an improved replacement system at the following refueling outage.
Pilgrim Nuclear Power Station On January 31, 1980 the Boston Edison Company (BECo) informed the NRC that five indications in the upper core spray sparger and two indications en the lower core spray sparger at the Pilgr..n Nuclear Power Station were identified during remote visual inservice inspections.
The indications were confirmed as cracks after hydrolasing and brush cleaning.
The licensees evaluation indicated that the sparger will retain structural integrity throughout the next cycle, although core spray flow distribution may be affected due to through-wall cracks.
However, core spray flow delivery tc the shroud interior would not be expected to decrease.
A loose parts analysis was presented which addressed (1) corrosion, (2) flow blockage, and (3) control rod interference.
To support power operation in Cycle 5 with the core spray sparger in its present condition, BECo has reanalyzed ECCS taking credit only for core spray reflood, taking no credit for core spray heat transfer.
The submission by BECo is currently under review by the staff.
The analysis is expected to cover a full spectrum of core spray failures.
It is expected that the limiting condition will be the failure of recirculation suction line.
A MAPLHGR limit reduction will likely be imposed during Cycle 5 to compensate for the assump-tion of no core spray heat transfer.
Based on results from other sparger inspections and previous pipe cracking experience, cold work and sensitization during fabrication and installation stresses are considered to be the major factors in causing the observed cracks at the Pilgrim Station.
The cracks are hypothesized to be initiated and propagated by intergranular stress corrosion (IGSCC).
.g A meeting was held with representatives from GE in Bethesda, Maryland on March 13, 1980 to discuss core spray sparger cracking at BWRs.
At the meeting GE provided the following information:
1.
In February 1979,-GE issued to BWR licensees Service Information Letter (SIL) No. 289 that recommended inspection of the core spray spargers for visual indications of cracking. To date, 19 of 21 plants inspected have no observed cracking.
Cracks have been found at 2 facilities (Pilgrim and Oyster Creek).
2.
The key contributors to IGSCC vary from plant-to plant, although stresses from cold work and sensitization during fabrication and installation are considered prime factors leading to IGSCC at Pilgrim and Oyster Creek.
Because the cause of cracking it not yet confirmed by metallurgical analysis, GE is developing tooling to extract sparger samples to verify the postulated _ cracking mechanism.
IE Bulletin No. 80-13 May 12, 1980 Page 3 of 3 3.
GE is evaluating methods of improving the sparger inspection techniques, and is considering a modification to the SIL, if warranted.
The staff agreed that improved inspection techniques should be developed and metallurgical examinations should be performed to determine the mode of failure.
The staff asked GE to keep them informed of progress in these areas.
Actions to be Taken by Licensees:
For all boiling water power reactor facilities with an operating license:
1.
At the next scheduled and each following refueling outage until further notice, perform a visual inspection of the Core Spray Spargers and the segment of piping between the inlet nozzle and the vessel shroud.
Remote underwater TV examirations are acceptable if adequate resolution can be demonstrated. The viewing in situ of 0.001 in. diameter fine wires is considered as an acceptable means of demonstrating suitable rusolution of the TV examinations.
Such techniques as the use of oblique lighting, and the ability to light from each side independently are considered useful in enhancing the image of cracks to facilitate detection.
2.
In the event cracks are identified during examination of the core spray sparger system, the location and extent of the indications shall be recorded and reported to the NRC.
Supplementary examinations using volumetric methods may be performed to aid in characterizing the extent of cracking in nonvisible locations.
An evaluation shall be submitted to NRR for review and approval prior to return to operation.
3.
Any cracking idantified in the core spray cooling system shall be reported to the Director of the. appropriate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification.
i 4.
A written report of the results of the examinations including any
...m corrective measures taken shall-be submitted within 30 days of the cc:aple tion of the examination to the Director of the NRC Regional Office with a copy to tha %C Office of Inspection and Enforcement, Division of Reactor Operations In:.pection, Washington, D. C.
20555.
Approved by GAO, B180225 (R0072); clearance expires 7-31-80.
Approval was given under a blanket clearance specifically for identified generic problems.
a IE Bulletin No. 80-13 Enclosure May.12, 1980 RECENTLY ISSUE 0 IE BULLETINS Bulletin Subject Date Issued Issued To No.
80-12 Decay Heat Removal System 5/9/80 Each PWR with an OL Operability 80-11 Masonry Wall Design 5/8/80 All power reactor facilities with an OL, except Trojan
.80-10 Contamination of 5/6/80 All power reactor Nonradioactive System and facilities with an Resulting Potential for OL or CP Uomonitored, Uncontrolled Release to Environment 80-09 Hydramotor Actuator 4/17/80 All power reactor Deficiencies operating facilities and holders of power reactor i
construction permits 80-08 Examination of Containment 4/7/80 All power reactors with Liner Penetration Welds a CP and/or OL no later than April 7, 1980 80-07 BWR Jet Pump Assembly 4/4/80 All GE BWR-3 and Failure BWR-4 facilities with an OL 80-06 Engineered Safety Feature 3/13/30 All power reactor (ESF) Reset Controls facilities with an.0L 80-05 Vacuum Condition Resulting 3/10/80 All PWR power reactor In Damage To Chemical Volume facilities holding Control System (CVCS) Holdup OLs and to those with Tanks a CP 79-01B Environmental Qualification 2/29/80 All power reactor of Class IE Equipment facilities with an OL
)
80-04 Analysis of a PWR Main 2/8/80 All PWR reactor facilities Steam Line Break With holding OLs and to those Continued Feedwater nearing licensing Addition 80-03 Loss of Charcoal From 2/6/80 All holders of Power Standard Type II, 2 Inch, Reactor OLs and cps Tray Adsorber Cells
_