ML17053B469
| ML17053B469 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 02/29/1980 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Rhode G NIAGARA MOHAWK POWER CORP. |
| References | |
| NUDOCS 8003120532 | |
| Download: ML17053B469 (8) | |
Text
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+0 'o" UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 631 PARK AVENUE KING OF PRUSSIA, PENNSYLVANIA19406 Docket No. 50;410 Niagara Mohawk Power Corporation ATTN:
Mr.
G.
K.
Rhode Vice President System Project Management 300 Erie Boulevard, Mest Syracuse; New York 13202
~~B 2 8 1980 Gentlemen:
The enclosed IE Information Notice No. 80-07, Pump Shaft Fatigue Cracking, is forwarded to you for information.
No written response is required, Should you have any questions regarding this matter, please contact this office.
Sincerely, Boyce H. Grier Director
Enclosures:
l.
IE Information Notice No. 80-07 2.
List of Recently Issued IE Information Notices CONTACT:
L.
E. Tripp (215-337-5282) cc w/encl s:
Eugene B. Thomas, Jr.,
Esquire
ENCLOSURE 1 SSINS No.:6870 Accession No.:
7912190675 PUMP SHAFT FATIGUE CRACKING UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 IE Information Notice No. 80-07 Date:
February 29, 1980 Page 1 of 2 On 'October 9, 1979, the Tennessee Valley Authority (TVA) reported to NRC that low-stress high-cycle fatigue cracks had been found in two non-safety-related feedwater pumps at their Browns Ferry BWR facility.
One pump had been operated for approximately 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and the other for approximately 28,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
TVA indicated that excessive vibration had been experienced prior to detection of the cracks by ultrasonic and visual inspection.
TVA also reported that catastrophic failure had been experienced with a boiler feed pump at their Paradise fossil unit.
The TVA failed pump shafts were of type 414 stainless steel.
TVA plans to replace the shafts with a type 17-4ph stainless steel.
A number of centrifugal charging/safety injection pump shaft failures have been reported since 1977.
The pumps of interest are the dual purpose centri-fugal pumps which are utilized for a normal charging function and in the safeguard mode are aligned for high pressure borated coolant injection.
These pumps were procured by Westinghouse from the Pacific Pump Division of Dresser Industries.
The pump shaft material was supplied to Pacific Pump by the Carpenter Technology Corporation.
I Westinghouse and Pacific Pump have been conducting an on-going investigative program since 1977 to resolve the shaft failure problem.
Results from the program suggested that the causes of the shaft failures could be related to shaft material deficiencies, design and/or abnormal operation of the centri-fugal pumps.
Forty centrifugal charging/safety injection pump shafts fabricated from four heats of A276 type 414 stainless steel bar were air cooled and tempered at 1000 F.
The above stated heat treating practice resulted in shafts with low toughness (Charpy "V" notch energies
- Longitudinal 6-10 ft-lbs, transverse 2-3 ft-lbs)
Corrective actions taken to reduce the incidence of charging pump/safety injection pump shaft failure have included:
Replacement of the type 414 low toughness stainless steel material.
Replacement material for the char ging pump shafts is type 414 stainless steel, oil quenched and tempered at 1150-1200 F.
Some of the air cooled shafts have been re-tempered at 1150 F.
Enclosure 1
IE Information Notice No. 80-07 Date:
February 29, 1980 Page 2 of 2 2.
Design modifications to the shaft to reduce stress raisers.
The modif-ications included increasing the fillet radius in the split ring groove, increasing the thread root in the locknut section of the shaft and the use of formed cutting tools during fabricati'on.
3.
A review of the operating history of the dual purpose centrifugal pumps indicate abnormal operation to be a potentially significant 'contributor to the early failure of the pump shafts.
Abnormal operation includes any condition that results in partial or complete loss of fluid in the pump or continued operation under high vibration conditions caused by misalign-ment or other installation problems.
Mestinghouse Nuclear Service Division s Technical Bulletins 77-9, 78-1 Rev; 1 and 79-6 provide guidance for monitoring of and limits for pump vibration as well as guidance for pump operation and maintenance.
Ten additional Safety Injection Pumps manufactured by Pacific Pump were identified as having shafts with the same material and heat treatment as the failed charging/safety injection pump shafts, These ten additional pumps are of a different design and provide for medium pressure boron injection in the safeguard mode but do not function in normal operation.
Licensee's are encouraged to review the materials utilized for and the design of shafts in safety-related pumps and non-safety-related pumps which could affect safety-related system performance.
Particular attention should be paid to abnormal operating conditions which could induce high stress in the pump shafts, susceptibility of materials to crack initiation and growth, and the presence of stress raisers.
This Information Notice is provided as an early notification of a possibly significant matter that is still under review by the NRC staff. It is expected that recipients will review the information for possible applicability to their facility.
No specific ac'tion or response is requested at this time. If you have any questions regarding this matter, please contact the Director of the appropriate NRC Regional Office.
0 ENCLOSURE 2
IE Informat'ion Notice No. 80-07 Date:
February 29, 1980 Page 1 of 1 RECENTLY ISSUED IE INFORMATION NOTICES Information Notice No. 79-31 Subject Use of Incorrect Amplified
Response
Spectra (ARS)
Date Issued 12/13/79 Issued to All Power Reactor Faci-lities with an Operating License (OL) or Construc-tion Permit (CP) 79-32 79-33 Separation of Electrical Cables for HPCI and ADS Improper Closure of'rimary Containment Access Hatches 12/21/79 All Power Reactor Faci-lities with an OL or CP 12/21/79 All Power Reactor Faci-lities with an OL or CP 79"34 Inadequate Design of Safety-
.12/31/79 All Power Reactor Faci-Related Heat Exchangers lities with an OL or CP 79-35 Control of Maintenance and Essential Equipment 12/31/79 All Power Reactor Faci-lities with an OL or CP 79-36 79"37 80-01 80-02 80-03 80-04 80-05 80-06 Computer Code Defect in Stress Analysis of Piping Elbow Cracking in Low Presssure Turbine Discs Fuel Handling Events 8Ã8R Mater Rod Lower End Plug Mear Main Turbine Electro-hydraulic Control System BMR Fuel Exposure in Excess of Limits Chloride Contamination of Safety Related Piping and Components Notification of Signif-icant Events 12/31/79 12/31/79 1/4/80 1/25/80 1/31/80 2/4/80 2/8/80 2/27/80 All Power Reactor Faci-lities with an OL or CP All Power Reactor Faci-lities with an OL or CP All Power Reactor Faci-lities with an OL or CP All BMR Facilities with an OL or CP All Power Reactor Faci-lities with an OL or CP All BWR Facilities with an OL or CP All Power Reactor Faci-lities with an OL or CP and applicants for a CP All Power Reactor Faci-lities with an OL and applicant for OL
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