ML19253A500
| ML19253A500 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/19/1979 |
| From: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| NUDOCS 7909100167 | |
| Download: ML19253A500 (9) | |
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U.S. Nuclear Regulatory Commission Region V Suite 202, Walnut Creek Plaza 1990 N. California Blvd.
Walnut Creek, CA 94596
Dear Sir:
Attached are our responses to concerns raised in IE Bulletin 79-13.
This informatio involves recent nondestructive testing performed to guarantee that no corrosion-fatigue or tharmal-fatigue cracking exists in the Trojan feedwater piping.
Supplemental information concerning this subject has previously been submitted to Mr. Victor Stello, Director of the Division of Cperating Reactors, by letters dated June 18, 1979 and July 19, 1979.
Also attached are copies of these letters for your information and use.
If you have any questions regarding this submittal, please contact me.
Sincerely,
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C. Gaodwin, Jr.
Assistant Vice President Thertaal Plant Operation and Operation CC/SML/4kk8A15 Attachments c:
'tr. Lynn Frank, Director w/ attach State of Oregon Department of Energy S350'38 7gog1oo/g n ns m.:
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ATTACHMENT 1 RESPONSES TO IC BULLETIN 79-13 CONCERNS NRC Request 1 Facilities which have steam generators fabricated by Westinghouse or Combustion Engineering that have not conducted volumetric examination of feedwater nozzles since May 1979 shall complete the inspection program described 'celow at the earliest practical time but no later than 90 daya after the date of this Bulletin.
a.
Perform radiographic examination, supplemented by ultra-sonic examination as necessary to evaluate, indications, of all feedwater nozzle-to piping welds and of adjacent pipe and nozzle areas (a distance equal to at leat. two wall thicknesses).
Evaluation shall be in accordance with ASME Section III, Subsection NC, Article NC-5000.
Radiography shall be perforned to the 2T penetraneter sensitivity level, in lieu of Table NC-5111-1, with systems void of water.
b.
If cracking is identified during examination of the nozzle-to piping weld, all feedwater line welds up to the first piping support or snubber and high stress points in contain-ment shall be volumetrically examined in accordance with 1.a. above. All unacceptable code discontinuities, other than cracking, shall be subject to repair unless justifica-tion for continued operation is provided.
c.
Perform a visual inspection of feedwater system piping supports and snubbers in Containment to verify operability and conformance to design.
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PCE Response Radiographic and ultrasonic examination of all feedwater a.
nozzle-to-reducer welds and adjacent pipe-to-reducer welds has been perforned. Radiography was performed to the 2T penetrameter sensitivity level with the system dry.
Evaluation of the results is being performed in accordance with ASME Section III, Subsection NC, Article NC-5000.
A written report of the results will be available by August 10, 1979.
b.
No cracking was found during the examination.
Although PCE had performed volumetric examination of all c.
four feedwater nozzles "...since May 1979..." when the Bulletin was issued (directly as a result of the May 25, 1979 NRC letter requesting information on the same matter),
PGE decided to perform all inspections identified in the Bulletin.
Accordingly, a visual examination of all feed-water piping supports in Containment was performed on June 27, 1979.
The inspection was performed by a PGE mechanical engineer and stress analyst and involved the following specific visual checks:
a.
Crcut and anchor bolt integrity b.
Welds c.
Spring support readings d.
Nut and bolt tightness c.
Snubber cold pet f.
Pins and cotters g.
Clamp location (e.g., has any " walking" occurred) h.
Any signs of structural distress (e.g., bent steel) 1.
Interference with adjacent structures or components j.
Presence of unusual cold load.
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In addition, all snubbers were disconnected at one end and manually stroked to verify that lockup had not occurred. While the inspec-tion was performed, the piping was full of water and was at anbi-eut tenperature (i.e., <70 F).
The Plant was in a cold shutdown condition and had been for about 2 conths.
Findings Generally, the supports appeared in good condition.
All springs were at the cold load position and all mechanical snubbers (no hydraulic snubbers are used) were successfully stroked by hand. -
Two (2) restraints had the following conditions recorded:
a'.
EBB-3-1-SR-4 (Loop A) - This support is a rigid seismic strut.
Cracking of the grout was discovered as well as nInor uplift (less than 1/64 in.) of the baseplates.
The anchor bolts were tight and in good condition. The strut was apparently under load (in the cold condition).
b.
EBB-3-1-SR-8 (Loop B) - This support is the mirror image of the SR-4 support.
Similar conditions were observed for SR-8, incli ding a loaded strut and cracked grout.
All feedwater system piping supports and snubbers
- Contain-ment were verified to be operable and in conformance with intended design. Cracks observed in restraint baseplate grouting described above was determined to be minor and of no structural significance.
Anchor bolts were secure and base concrete was sound.
Loads on the two restraints were released by loosening turnbuckles.
Combined stresses from thermal loads and neasured displacements do not exceed code allowable values.
Although the cause of cold loading could not be determined, the two supports will be inspected during the next-scheduled Plant shutdown in the spring of 1980.
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NRC Request 2 All pressurized water reactor facilities shall perform the inspection program described below at the next outage of sufficient duration or at the next refueling outage after the inspection required by item 1.
a.
For steam generator designs having a common nozzle for both main and auxiliary (emergency) feedwater systems, perform volumetric examination of all feedwater nozzle-to-pipe weld areas and all feedwa cer nipe weld areas inside containment in accordance with item 1 above.
In addition, conduct an examination of welds connecting auxiliary feedwater piping to the main feedwater line outside contain-ment.
This examination should include an area ef at least one pipe diameter on the main feedwater line downstream of the connection.
b.
For steam generator designs with separate nozzles for main feedwater and auxiliary feedwa ter, perform volumetric examination (in accordance with item 1 above) of all welds inside containment and upstream of the external ring header or vessel nozzle for each steam generator.
If an external ring header is employed, also inspect all welds of one inlet riser on each feed ring of each steam generator.
c.
Perform a visual inspection of all feedwater system piping supports and snubbers in containment to verify operability and conformance to design.
PGE Response a.
PGE will perform the required feedwater and auxiliary feed-water piping examination during the next outage of sufficient duration or the next refueling o, age.
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b.
Not applicable.
c.
At the next outage of sufficient duration, PCE will per-form a visual inspection of all feedwater system piping supports and snubbers in Containment similar to the one performed June 27, 1979 (as discussed in Response 1.c).
NRC Request 3 Identification of cracking indications in feedvater nozzle or piping weld areas in one unit of a multi-unit facility shall require shutdown and inspection of other similar units which have not been inspected since May 1979, unless justification for continued operation is provided.
PCE Response Not applicable.
NRC Request 4 Any cracking or other unacceptable code discontinuities identified 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.
PGE Response Not applicable.
NRC Request S Provide a written report to the Director of the appropriate NRC Regional Of fice within 20 days of the date of this Bulletin addressing the following:
a.
Your schedule for inspection if required by item 1.
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The adeqtacy of y )ur operating and emergency procedures to recognize and respond to a feedwater line break accident.
c.
The methods and sensitivity of detection of feedwater leaks in Containment.
PGE Response a.
As stated in PCE Response 1, the inspection has already been ne rfe rmed.
b.
The operating and emergency procedures for a feedwater break accident have been reviewed and.are considered to provide adequate instru.tions to recognize this type of accident, to differentiate between it aci other accidents having similar indications, and to ensure the safety of the Plant following this accident.
c.
For a minor feedwater system leak, the following methods of detection would be available:
a.
Containment llumidity Detector System - The Containment llumidity Detector System provides a means of measuring overall Icakage from all water and steam systems within the Containment.
The detector is a psychrometric instrument consisting of a water box and blower assembly, two thermometers, one of which is a wet bulb and one a dry bulb, and one pint flask of water used to keep the wet bulb cloth wick moist. The blower provides uufficient air flow past the two thermometers to ensure that the wet and dry bulb temperatures are representative of the Containment atmos-phere. The temperature values are displayed in the control nnr n. 4 u tJ u u c; m,.W"M ' :
room.
By comparing wet and dry bulb temperatures on a pyschrometric chart, values for specific humidity within the Containment are obtained which, when monitored over a period of time, provide a means of ueasuring overall leakage within the Containment.
The sensitivity of this system has been verified by its demonstrated ability to respond to small amounts of steam leakage inside the Containment.
b.
Containment Condensate Measuring System - The Containment Condensate Measuring System is a leakage detecrion system which collects and measures-moisture condensed from the Containment atmosphere by the cooling coils of the four Conta. ment air coolers.
The condensation from each of the four cooler units is collected and directed to a common header which drains to a 5 gal capacity collection vessel.
The collection vessel has a level switch which actuates a drain line isolation valve and indicates in the control room whenever the 5 gal vessel capacity is reached. The condensation is then drained by gravity to.the Containment Building sump "B".
This leakage detection system relies on the principle that all leakages up to sizes permissible with continued Plant operation will be evaporated into the Containment atmosphere. This system provides a dependable and accurate means of measuring integrated total leakage.
A major failure of a feedwater piping system could be recognized by combinations of the following systems: c o m a v.
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- 1) Pressurizer low pressure
- 11) Pressurizer low level lii) Increasing Containment pressure iv) High Containment humidity v) Steam flow / feed flow mismatch vi) Rapidly decreasing reactor coolant average temperature vii) High steam line differential pressure viii) Steam line high flow ix) Low steam generator water level.
NRC Request 6 A written report of the results of examinations, in accorda..ce with requests by Regional Offices preceding this Bulletin und with Bulletin iten 1 and 2 including any corrective measures taken shall be submitted within 30 days of the date of this Bulletin or within 30 days of com-pletion of the examination, whichever is later, to the Director of the appropriate NRC Regional Office with a copy to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D. C. 20555.
PCE Response As stated in PGE Response 1, a written report discussing the examination results will be submitted by August 10, 1979.
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