IR 05000334/1993002
| ML20057D392 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/23/1993 |
| From: | Durr J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Sieber J DUQUESNE LIGHT CO. |
| References | |
| NUDOCS 9310040174 | |
| Download: ML20057D392 (2) | |
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SEP 2 31993 i
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Docket Nos. 50-334 l
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50-412 Mr. J. Senior Vice President Nuclear Power Division
Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077
Dear Mr. Sieber:
SUBJECT: INSPECTION 50-334/412/93-02 This reters to your May 21, 1993, correspondence, in response to our April 14, 1993, letter.
Thank you for informing us of the corrective and preventive actions docurraaed in your letter. These actions will be examined during a future inspection of your licensed prograin.
Your cooperation with us is appreciated.
Sincerely, M ietn:1 sinna 07:
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Jacqu P. Durr, Chief Engineering Branch Division of Reactor Safety (
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SEP 2 31993 Mr. J.' '
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G. S. Thomas, Vice President, Nuclear Services D. E. Spoerry, Vice President, Nuclear Operations L. R. Freeland, General Manager, Nuclear Operations Unit K. D. Grada, Manager, Quality Services Unit N. R. Tonet, Manager, Nuclear Safety Department H. R. Caldwell, General Superintendent, Nuclear Operations K. Abraham, PAO (10) SALP Reports and (2) All Inspection Reports Public Document Room (PDR)
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Local Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector Commonwealth of Pennsylvania State of Ohio bec w/ encl:
Region I Docket Room (with concurrences)
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W. Butler, NRR G. Edison, h7tR V. McCree, OEDO
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Beaver Valley Power Station I
Sh4ppingport. P A 15077-0004
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JOHN D. SIEBER (412) 393-5255 Senior Vice President and Fax (412) 643-8069 Chief Nuclear Officer Nuciese Power Division May 21, 1993 U.
S. Nuclear Regulatory Commission Attn:
Document Control Desk
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Washington, DC 20555 Subject:
itw2 aver Vaalley rower Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Inspection Report 50-334/93-02 - Notice of Violation In response to NRC correspondence dated April 14, 1993, and in accordance with 10 CFR 2.201, the attached reply addresses the Notice i
of Violation transmitted with the subject inspection report.
As requested, this reply addresses the actions taken to restore the river water piping to its original construction margins and to preclude similar deviations in the future.
If there are any questions concerning this response, please contact Mr. N. R. Tonet at (412) 393-5210.
Sincerely,
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. Attachment cc: Mr. L.
W. Rossbsch, Sr. Resident Inspector l
Mr. T.
T. Martin, NRC Region I Administrator (
Mr. J.
P. Duvr, Chief, Engineering Branch Divisiore of Reactor Safety, Region I Mr.
G.
E.
Edison, Project Manager
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Mr. M.
L. Bowling (VEPCO)
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_M Beaver Va ey Power Station Shippmgport. PA 15077-0004 JOHN D. SiELER 1412) 393-5255 Senior Vice President and Fax (412) 643-8069 Cheef Nuclear Officer Nuclear Power Division May 21, 1993
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U.
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Nuclear Regulatory Commission
Attn:
Document Control Desk Washington, DC 20555 j
i subject:
Beaver Valley rower station, Unit No. I Docket No. 50-334, License No. DPR-66 Inspection Report 50-334/93-02 - Notice of Violation
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In response to NRC correspondence dated April 14, 1993, and in accordance with 10 CFR 2.201, the attached reply addresses the Notice of Violation transmitted with the subject inspection report.
As
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requested, this reply addresses the actions taken to restore the river water piping to its original construction margins and to
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preclude similar deviations in the future.
l If there are any questions concerning this response, please contact Mr.
N. R. Tonet at (412) 393-5210.
Sincerely,
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. Attachment cc: Mr. L.
W.
Rossbach, Sr. Resident Inspector Mr. T.
T. Martin, NRC Region I Administrator k
Mr. J.
P. Durr, Chief, Engineering Branch Division of Reactor Safety, Region I Mr.
G.
E.
Edison, Project Manager Mr. M.
L. Bowling (VEPCO)
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DUQUESNE LIGHT COMPANY
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Nuclear Power Division
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Beaver Valley Power Station, Unit No. 1
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Ren1v to Notice of ViolatioD Combined Inspection Report 50-334/93-02 and 50-412/93-02 Letter dated April 14, 1993 VIOLATION (Severity Level IV; Supplement I)
Descriotion of Violation (50-334/93-02-01)
CFR 50. 55a (g) (4 ) requires that components which are classified as ASME Code Class
(or an equivalent system as determined by the original construction code, e.g.
ANSI B31.1 for the River Water
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System at Unit 1)
shall meet the requirements of Section XI of the ASME Boiler and Pressure Vessel Code.
Article IWA-4000 of Section XI of the ASME Code describes the Code repair procedures.
IWA 4340 requires the affected surfaces to be examined by nagnetic particle or liquid penetrant inspection to ensure that the indication has been reduced to an acceptable limit, or when repair welding is to be performed, the original defect shall be removed.
Contrary to the above during 1986 and 1989, the licensee repaired three areas in the River Water System because of pinhole leaks and wall thinning without assuring that all flaws and surface degradation were removed prior to executing the repairs.
Discussion of the Violation The violation cited paragraph IWA 4340 of ASME Section XI.
This paragraph is in the 1986 Edition of Section XI but does not exist in the 1983 Edition of ASME Section XI, which is the code edition that l
Beaver Valley Unit
is required to meet in accordance with 10 CFR 50. 55a(g) (4) (ii).
However, paragraph IWD-4 200 (b) (1)
of the 1983 Edition of Section XI contains requirements similar to those cited in
the violation.
I Reason for the Violation At the time of the repairs, the actions taken were considered to meet code requirements.
These actions included notification of the Authorized Nuclear Inservice Inspector, grinding of the pinhole areas, magnetic particle examination of the ground areas, repair welding of the affected areas utilizing qualified welders, procedures and materials, and the performance of post-repair pressure tests.
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Reply to Notica of Violation
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Corrective Action Taken The section of river water piping containing the repaired areas was replaced during the ninth refueling outage.
The replacement section was fabricated to the original design specifications and therefore restored that section of piping to its original construction margins.
Actions Taken to Prevent Recurrence The ASME Section XI Repair / Replacement Program (NPDAP 8.5) will be revised to provide guidance on the use of non-code repairs on ASME Code Class 1,
and 3 piping.
This revision will require that any proposed use of a temporary non-code repair be concurred with by the Nuclear Engineering Department and that the guidance provided by NRC Generic Letter 90-05 be met.
l Date When Full Compliance Will Be Achieved The station is in full compliance at this time.
The revision to NPDAP 8.5 will be issued by August 31, 1993.
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APR 141993 Docket Nos. 50-334 50-412 Mr. J. Senior Vice President e
Nuclear Power Division Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077
Dear Mr. Sieber:
SUBJECT: NRC INSPECTION REPORT NOS. 50-334/93-02 AND 50-412/93-02 This letter refers to the safety inspection of engineering and technical support activities
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related to the Duquesne Light Company Beaver Valley Power Station (BVPS) Unit Nos. I and 2, conducted by Mr. H. Kaplan, Mr. A. Imhmeier, and Mr. K. Battige of this office during the period January 11-15, 1993, at Shippingport, Pennsylvania. At the conclusion of the inspection, their findings were discussed widi you and members of the
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Beaver Valley engineering staff.
The scope of this inspection included review of your erosion / corrosion program, river water / service water systems, transient operating cycle monitoring, reactor vessel material surveillance, reactor coolant system iron pickup, and engineering performance monitoring.
Effective performance in these areas by your engineering and technical support personnel is important to the safe operation of the plant.
Based on the results of this inspection, your engineering and technical support staff was found to have effectively addressed these areas. However, there were indications of the need for some improvement in the identification and correction of the condition of service water system equipment. In addition, it was found that one of your activities was in violation of NRC requirements, as set forth in the Notice of Violation" enclosed herewith as Appendix A. The violation has been characterized by severity level in accordance with the i
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APR 141993
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Mr. J.,
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Region I Docket Room (with concurrences)
DRS/EB SALP Coordinator J. Linville, DRP W. I mrus, DRP L. Rossbach, SRI - Beaver Valley SRI, Perry, Region HI W. Butler, NRR G. Edison, NRR V. McCree, OEDO i
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APP i 41993 Docket Nos. 50-334 50-412 Mr. J. Senior Vice President Nuclear Power Division Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077
Dear Mr. Sieber:
J SUBJECT: NRC INSPECTION REPORT NOS. 50-334/93-02 AND 50-412/93-02 l
This letter refers to the safety inspection of engineering and technical support activities related to the Duquesne Light Company Beaver Valley Power Station (BVPS) Unit Nos. I and 2, conducted by Mr. H. Kaplan, Mr. A. Imhmeier, and Mr. K. Battige of this office during the period January 11-15,1993, at Shippingport, Pennsylvania. At the conclusion of the inspection, their findings were discussed with you and members of the l
Beaver Valley engineering staff.
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The scope of this inspection included review of your erosion / corrosion program, river water / service water systems, transient operating cycle monitoring, reactor vessel material surveillance, reactor coolant system iron pickup, and engineering performance monitoring.
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Effective performance in these areas by your engineering and technical support personnel is
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important to the safe operation of the plant.
Based on the results of this inspection, your engineering and technical support staff was found to have effectively addressed these areas. However, there were indications of the need for some improvement in the identification and correction of the condition of service water system equipment. In addition, it was found that one of your activities was in violation of NRC requirements, as set forth in the " Notice of Violation" enclosed herewith as Appendix A. The violation has been characterized by severity level in accordance with the P
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APR 141993
Mr. J. General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (Enforcement Policy). You are required to respond to the Notice of Violation.
In preparing your response, you should follow the instructions in Appendix A.
The practice of installing reinforcing patches to seismic category I power piping is considered to be beyond your licensing bases. The staff has provided guidance for the repair of degraded piping in Generic Letter 90-05, Guidance for Performing Temporary Non-Code Repair of ASME Code Class I,2, and 3 Piping. Although a specific violation has not been issued for this prac&e, we view this as a serious matter and expect appropriate actions to be taken to ensure the stmetural integrity of the piping is not degraded. In yout response to the
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Notice of Violation, please provide a discussion of your actions to restore the piping system to its original construction margins and any actions to preclude similar deviations from your licensing bases.
Sincerely, no MC Ja ue P. Durr, Chief Engineering Branch Division of Reactor Safety
Enclosure:
NRC Region I Inspection Report Nos. 50-334/9342 and 50-412/9342
REGION I==
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REPORT / DOCKET NOS. 50-334/93-02 50-412/93 42
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LICENSE NOS.
DPR-66 NPF-73
LICENSEE:
Duquesne Light Company i
FACILITY NAME:
Beaver Valley Power Station Unit Nos.1 & 2
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I INSPECTION AT:
Shippingport, PA INSPECTION DATES:
January Il-15,1993 INSPECTORS:
u _ yalb 2-13 q3 H.' KaplAn, Sr. Reactor Engineer, Date.
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Materials Section, EB, DRS
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A. IAhmeier, Sr. Reactor Engineer, D' ate Materials Section, EB, DRS
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APPROVED BY:
E. Harold Gray, Chief, Materials Section, Date -
Engineering Branch, DRS
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1 Areas Insoected: The scope of this inspection included review of your erosion / corrosion program, river water / service water systems, transient operating cycle monitoring, reactor vessel material surveillance, reactor coolant system Iron pickup, and engineering performance monitoring to assess the effectiveness of engineering and technical support personnel in providing the corrective actions important to the safe operation of the plant.
Emults: The licensee's engineering and technical support staff effectively addressed issues and provided effectively engineered corrective action plans. However, there were indications of a need for improvement in identification and corrective action in maintenance of service water system equipment. It was also found that pipe repairs were in violation of NRC requirements.
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.l.0 SCOPE OF INSPECTION The scope of this inspection included review of the licensee's erosion / corrosion program, river water / service water systems, transient operating cycle monitoring, reactor vessel material surveillance, reactor coolant system Iron pickup, and engineering performance monitoring to assess the effectiveness of engineering and technical support personnel in providing the corrective action important to the safe operation of the plant.
2.0 EROSION / CORROSION PROGRAM (Inspection Procedure (IP) 49001)
The inspector reviewed the results of the erosion / corrosion (E/C) program initiated by the licensee in January 1987.
2.1 Hndings The licensee provided the inspector with records which inficated that ultrasonic thickness measurements were performed on 173 components in Unit 1, and 139 components in Unit 2 between 1987 and the present. A review of training records indicated that the Ixvel II
nondestmetive examination technicians who performed the measurements were properly qualified at the time. The data indicated that except for two components in Unit 2, no evidence of significant wear had been found. The first case in which wear was found was a
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10 inch,90 degree elbow in the feedwater system downstream of a multistage orifice that i
averaged 85 mils loss between two outages. Although the licensee is of the opinion that the wall loss is due to flashing rather than E/C, they have dec ded to replace this elbow during the next outage. The second case involved a twenty inch elbow in the condensate system that showed wear, but after being weld repaired in refuel outage 1 (RF1), it has not suffered any significant wall loss. Additionally, eight elbows in Unit 2 were suspected of_ experiencing E/C. After destructively examining two of these elbows, it was determined that the apparent loss in wall thickness was not due to E/C, but was a condition that was present in the original material since the ID surface still showed evidence of substantial surface decarburization and scale that would not have been present if FJC had occurred. The inspector reviewed the metallurgical report that showed this condition and concurred with the conclusion. 'Ihe licensee attributes the lack of significant E/C to their secondary water control that features high pH attendant with the use of morpholine.
The requirements for the E/C program are detailed in the Engineering Standard ES-M-009, dated 10/16/87. The inspector's review of this standard revealed two weaknesses which the licensee acknowledged, and agreed to correct. These were (1) inadequate expansion criteria, and (2) base line data requirements for new materials.
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The inspector obtained the following information regarding the J tubes in the Unit I steam generators (SGs) which showed some wall thinning in RF3/RF4. SG-A, which appears to have the most degraded J tubes of the three SG's, will be inspected during the next outage (RF9). Those tubes showing excessive wear will be replaced with Inconel tubes on a case-by-case basis. Unit 2 steam generator J tubes were replaced with Inconel at the site prior to initial startup.
The licensee noted that, although the EPRI CHEC computer code had been used to a limited extent, its use was to be increased in future inspections. In addition, trending of the data covering at least three outages for each plant would be initiated.
2.2 Conclusion The licensee is implementing a comprehensive E/C monitoring program.
3.0 RIVER WATER (UNIT 1) AND SERVICE WATER (UNIT 2) SYSTEMS (IP 37702)
The licensee prcvided the inspector with a comprehensive historical summary report of the service / river water systems in Units 1 and 2. The inspector's assessment of the report is as follows. Also included are the observations reade by the inspectors during a walkdown of the river / service water system and review of the licensee's clam contml system.
3.1 Mndings Since the spring of 1991, the licensee expended considerable effort to assure system reliability. 'Ite major actions, some of which were recommended by Generic letter 89-13 and others due to natural causes (e.g. clam blockage), included installation of surveillance flow test instrumentation, chemical cleaning and hydrolazing, biocide treatments, eddy current inspection of heat exchanger tubes, and replacement of piping, valves, pump impellers and tubing as required.
In reviewing the leakage and repair records of the Unit I service water system, the inspector discussed the Code acceptability of three repairs in component cooling heat exchanger return lines identified as WR-14, WR-15, and WR-16. The return lines were fabricated from 18 inch diameter x.375 inch wall, A106 carbon steel piping. In regard to the details of the repairs, which occurred in 1986 and 1989, the licensee provided the following information in a letter, dated January 25,1993:
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The repairs, one area with wall thinning (WR-14) and in two areas (WR-15-16), which contained pin hole leaks, consisted of: (1) locally removing the defects by grinding from the O.D., followed by magnetic particle testing (MT), (2) welding the ground out areas on WR15-16, (3) ultrasonic testing the affected areas, (4) fillet welding a patch plate over three areas, (5) weld overlaying areas outside the patch plate which were below minimum wall followed by MT, and (6) performm' g a pressure test.
In reviewing the licensee's submittal, the inspector r.oted that two aspects of the repairs were in conflict with Code requirements. These are: (1) failure to inspect the internal surfaces to assure removal of all defects surrounding the pin holes, as required by ASME Section XI, paragraph IWA-4340 as well as not inspecting the internal surfaces of those areas outside the area covered by the patch plates which were overlayed to compensate for wall loss, and (2)
even though the licensee chose to view the patch plates as welded attachments, the use of welded external plates is not considered an acceptable ANSI B31.1 design for a primary load carrying member in the event the thinning mechanism propagates a defect through the wall.
Paragraph 127.4.2 of ANSI B31.1 requires that girth welds be complete penetration welds.
Additionally, fillet welds are not desirable because they preclude inspection of the flawed area in the event that new defects appear. The cause of the wall thinning is believed to be due to cavitation caused by velocity changes resulting from throttling the adjoining butterfly valves. Inadequate inspection during the repairs, and the use of fillet welds in place of full penetration weids are violations of the codes. '(50-334/93-03-01). The NRC would not license a facility that used this technique during orig ~ mal construction and does not consider these repairs within the scope of the original code. The staff considers these changes to be beyond the licensing bases and that they require NRC approval as described in Generic I2tter 90-05, Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, 3 Piping.
' Die inspector reviewed procedures used by BVPS to limit clam growth and clogging of the service / river water systems. The licensee's primary method to insure' that fresh water clams do not clog the service / river water systems, especially the heat exchangers in these systems, is through the use of Clamtrol clamicide. Injected into the headers downstream of the intake structure and in the cooling tower basin, the biannual Clamtrol treatment is usually..
performed in the spring and fall. This coincides with the spawning season for maximum effectiveness. A neutralizing agent is added before water is returned to the river so as to minimize the ecological impact. Zebra mussels, a significant problem at other power plants, have not yet been detected in the section of the Ohio River used by BVPS for cooling.
Dudng a walkdown of the service /dver water system, the inspectors observed the condition of operating equipment and the weld repair. It was noted that the fillet welds on the pump line repair appeared to sufficiently cover the patch plate thickness. In obsening the
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condition of operating equipment, the following was noted:
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MOV RW 106A - A leaking packing gland
Unit 1 River Water Pump WR-P-1A - Oil leakage on floor
Unit 2 Pump 2SWSP21 A - Excessive water seal leakage
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These items should have been identified by plant staff during walkdowns or surveillances.
This repmsents a potential weakness in the visual appraisal of plent equipment by the plant staff. The licensee stated that the maintenance procedures in this area would be reviewed for appropriate corrective action.
3.3 Conclusions The licensee has generally expended considerable effort to maintain rehable service / river water systems by monitoring, testing, and replacement of degraded materials in the piping service / river water systems. Service / river water system clamicide treatment appears to effectively preclude system clogging. Repairs to three component cooling river water lines in Unit I appear to have been executed in violation of Code requirements. Several areas of equipment leakage have indicated the need for improved system walkdowns and equipment maintenance.
4.0 TRANSIENT OPERATING CYCLE MONITORING (IP 37700)
4.1 Background The primary system components are designed to meet the requirements of Section III of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels. The Code requires a " design by analysis" approach to evaluating not only whether the components can sustain the prescribed steady ' tate pressure and thermal loadings, but also the cyclic application of these loadings in view of the fatigue strength of the component materials.
The utility (owners of the components) specifies the types and circumstances ofloadings which are anticipated during the plant lifetime. Components are designed in accordance with these specifications. Therefore, in the case of cyclic loading, the specification will state the numbers and types of transient operation which can be anticipated throughout the plant life.
These transients are described in the Updated Final Safety Analysis Report (UFSAR) for the nuclear power plant.
Since primary system components are designed to sustain limited numbers of transients, the plant Technical Specifications (TS) require that records and documents relating to the cyclic operation of the plant must be maintained through the plant lifetime, so that critical areas of the components subject to the operating transients may be monitored to determine whether the design life of the component has been exhausted.
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The criteria for exhaustion of fatigue life are reflected in a cumulative usage factor (CUF),
which is an integrated summation of the ratio of expected numbers of cycles at the applied strain range to the cycles at that strain range necessary to cause fatigue failure. An appropriate factor of safety in terms of stnun level or cycles is utilized in the same sense as a factor of safety for stress level in relation to fracture stress.
4.2 Updated Final Safety Analysis Report and Technical Specification Requirements The Updated Final Safety Analysis Report for each unit at BVPS indicates the number and type of reactor coolant system (RCS) transients which each component is designed to sustain over the 40 year design life. These values are listed in " Summary of Reactor Coolant System Design Transients," UFSAR Table 4.1-10 for BVPS Unit I and UFSAR Table 3.9N-I for BVPS Unit 2c The two tables are similar, although limits for refueling (80) and RCS cold overpressurization (10) are expressed for Unit 2 only. TS Section 6.10.2f for each unit requires the licensee to retain, for the duration of the Facility Operating License, " Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles."
4.3 Documenting Transient Operation The Shift Technical Advisor (STA), who functions for both units, demonstrated to the inspectors the process of documenting RCS transients. In accordance with " Operations Assessment Procedure OAG 1.0," Revision 14, dated 4/11/91, the Operations Assessment Group (STA Group) is directed to maintain the Control Room Transient Iog, contact the Nuclear Engineering Dmrfment (NED) when safety injection actuations occur, and update monthly the " Record of Lindtin, Number of Occurrences of System Design Transients" for r
any that occurred during the month. The January Control Room Transient Iag was examined for both units. For transients listed in the UFSAR, the number experienced by each unit was consistent with unit age and below the limits set in the UFSAR.
One category had been added to the transient list beyond UFSAR requirements, " Safety Injection System (SIS)."Section VI, Step 7, of OAG 1.0 requires that the Nuclear Engineering Department (NED) be contacted for safety injection system trending. This activity is required by the Offsite Review Committee (ORC) letter ND10RC:0143, dated June 28,1985. The letter also states that safety injections should be trended with data analysis support from Nuclear Engineering.
Within the SIS category, for the "High Head Safety Injection" (HHSI) transient, Unit I had experienced 20 occurrences to date. The stated limiting occurrences is 25. NED personnel informed the inspectors that this value was derived from an engineering evaluation done by Westinghouse. While this number may seem high for a plant with more than halfits license period to go, the inspectors were informed that a large number of these occurred in the first cycle of Unit 1, a four and one half year cycle which was plagued by numerous trips and HHSI events. After this initial shakedown period, the number of HHSI transients decreased
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significantly. At the present time, the NED does not expect to exceed the HHSI limit before the end of the license. In an NRC letter to the licensee, dated December 18,1979, the Division of Operating Reactors' Safety Evaluation Repon on " Inadvertent Safety Injection During Cooldown," suggested that a plant specific analysis be performed if a Pressurized Water Reactor facility experienced more than 25 inadvertent safety ipjections. The licensee has demonstrated his awareness of the potential problem of exceeding the intended numbers of cycles and has provided for a corrective action plan for reevaluation of the fatigue life usage in such a case.
The " Record of Limiting Number of Occurmnces of System Design Transients", located in the Records Vault, showed UFSAR listed transients from the beginning of operation for each unit. The daylight shift STA updates this log on the first working day of each month with any transients experienced during that month. These actions appear to f,ully satisfy license requirements for transient logging.
4.4 Engineering Review of Transients The inspector noted that, although NED had no formal procedure for their review of plant transients, NED personnel had a very good understanding of the effect of cyclic transient operation and its relationship to fatigue of primary system components. The licensee has taken an active role in evaluating previously unevaluated operational phenomena such as pressurizer surge line thermal stratification.
4.5 Component Stress Report Review The inspectors briefly reviewed several stress analysis report for BVPS components, including " Analytical Report for Duquesne Light Company, BVPS Unit No.1 Reactor Vessel," Combustion Engineering Report No. CENC-1183, dated July 1972 and the Westinghouse Electric Company Steam Generator Stress Report.
In the Updated Final Safety Analysis Report, Section 4.1, the design parameters are given in Table 4.1-10 a " Summary of Reactor Coolant System Design Transients," which describes the numbers of occurrences of these transients for the normal, upset, faulted and test conditions. In monitoring these transients, the licensee noted that none of these transient cycles had been exceeded. Since the acceptance criteria limit for the CUF is 1.0 at the end of design life, the components have not been cycled such that the cumulative usage factor exceeds the design limitation.
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In review of the stress reports for both reactor and steam generators, the inspectors noted l
that the calculated lifetime usage factors for reactor components were as follows:
Comoonent
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Head Flange
.019 Vessel Flange
.004 Closum Studs (Position 7)
.635
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Primary Nozzles
.377
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Core Support Pad
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.1 CRDM Housing (IAcation 10)
.569 Further review of steam generator stress reports indicated the significant lifetime cumulative j
usage factors to be as follows:
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Comoonent-CUE
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Feedwater Nozzle
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Mist Extractor Skirt
.87 l
i The structural and fatigue analysis showed the Cumulative Usage Factor (CUF) value for
most structural components to be very low. _ The significantly high design CUF determined
were in the reactor bolting closure studs (.635), primary nozzles (.377), CRDM Housing
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(.569), Feedwater Nozzle (.917), and Mist Extractor Skirt (.87). However, for' a number of components, including the vessel closure studs and the control rod drive ' mechanism -
(CRDM), the CUF may vary widely with position. For example, the_CUF at location 9 on i
the CRDM housings is.004, three orders of magnitude below the CUF at position 10. The _
most restrictive value is used for the stress and fatigue analysis. These are lifetime
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predictions of CUF of which the licensee is aware and will be alert'to any changes in operation that will deviate from the design levels. Should the cyclic format change from that
.j assumed in design of the components, the licensee can reevaluate the components by
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reanalysis using less conservative assumptior.s as to severity of the cycles.
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4.6 Ca=eh>=tanc
The inspector concluded that the licensee was monitoring cyclic operation in accordance with
the technical specification requirements for retention of records, and in keeping with good engineering practice to be alert to changes in operational format.. The NED takes a proactive
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approach to evaluation of plant conditions, as evidenced by a paper on thermal stratification presented by a NED engineer in coopation with Stone & Webster Engineering Corporation.
"New Insights in Thermal Stratification of Feedwater Piping in PWR Plants," in PVP-Vol. 235, Design and Analysis of Pressure Vessels, Piping, and Components, ASME 1992.
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5.0 GENERIC LETIEK 92-01 " REACTOR VESSEL EMBRITTLEMENT"
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.1 The inspectors reviewed the licensee's response to Generic letter (GL) 92-01, Revision 1, in a letter of July 8,1992, to the NRC covering Beaver Valley Urdts 1 & 2. The licensee provided for a direct response to each concern (question) of G192-01. The inspector checked the responses to the letter and found the responses to be accurate for those areas
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inspected. The GI 92-01 response is under review by NRR.
6.0 REACTOR COOLANT SYSTEM - IRON PICKUP The inspector was informed that the monthly composite discharge sample for the second quarter of 1992 showed the Iron-55/ Cobalt-60 activity ratio to be ten times higher than
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normal. It was suspected that a cladding breach might be the cause of the high Iron content.
There is one known pre-existing area with a cladding flaw on the Unit 2 vessel interior.
This area is believed to have been caused by an are strike during manufacture. The licensee I
produced photographs of the area showing the results of an attempted repair. A Westinghouse analysis indicated there were no significant concerns for vessel corrosion. Due to the location of this area, it can only be inspected during the next 10 year in-service inspection scheduled four years from now.
Discussions with the BVPS chemical department engineers indicated the source to be the
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radiological waste line. As such, the Iron-55 may have come from any number of sources other than the than the reactor vessel. They also indicated that Manganese-54, a strong gamma emitter when activated, would be present if the reactor vessel cormsion products were present (SA 533, Grade B, Class I steel, with a Manganese content of 1.07% to 1.62%, is used for the shell and head plates). The Cobalt-60/ Manganese-54 ratio has been constant for two and one-half years, indicating no elevated Mangenese-54 levels.
Licensee engineering demonstrated the ability to evaluate the problem of Iron-55 deviation from normal observations.
7.0 ENGINEERING PERFORMANCE MONITORING The inspectors reviewed the monthly performance indicators for Unit 1 and Unit 2. For Unit 1, both the engineering memoranda backlog and corrective maintenance rework (based on engineering error) are within objectives. For Unit 2, temporary modifications are slightly above goal. NRC Violations for both units are below the goal set for the year, and the corrective maintenance backlog is within the established goal. The licensee is monitoring its engineering performance relative to goals set for the division in a comprehensive manner.
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I 8.0 SUMMARY OF CONCLUSIONS ne licensee is implementing a comprehensive FJC monitoring program.
f The licensee expended considerable effort to maintain reliable service / river water systems in terms of monitoring, testing and replacement of degraded materials in the piping service / river
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water systems. Service / river water system clamicide treatment appears to effectively pre
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system cloggmg.
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Repairs to three component cooling river water lines appear to have been executed in
l violation of code requirements.
Several areas of equipment leakage have indicated need for improving system walkdowns and:
equipment maintenance.
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l The inspector observed that the licensee is monitoring cyclic operation in accordance with thl j
technical specification requirements for retention of records and in keeping with good engineering practice to be alert to changes in operational format. The NED takes a pro approach to evaluation of plant conditions.
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Ucensee engineering., demonstrated the ability to evaluate the problem of Fe-55 di
from normal observanons.
The licensee is monitoring its engineering performance relative to goals set for the division in
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i a comprehensive manner.
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l 9.0 MANAGEMENT MEETINGS
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Entrance meetings were held on January 11 and 12,1993, with Duquesne Light Company
personnel. An exit meeting to discuss inspection findings with licensee personnel was j
on January 15, 1993. Attendees are listed in Attachment A.
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t ATTACHMENT I
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Duouesne Licht and Power Company t
G. Albeni, ISI Coordinator
i J. Baumler, Director - Quality Services
G. Brown, Sr. Chemist
A. Dulick, Chemistry Manager J. Finke, Engineer - Inservice Inspection
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L. Freeland, General Manager - Nuclear Operations
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K. Halliday, Director - Electrical Engineering
R. Hansen, Director - General Engineering Department
T. Huminski, Sr. Engineer G. Kammerdeiner, Director - Materials Engineering
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V. Linnenbom, Corporate Nuclear Chemist F. Lipchick, Senior - Licensing Supervisor
A. Iennett, Sr. H.P. Specialist C. McFeaters, MEA - Plant Performance Supervisor G. McLain, Manager - Maintenance Engineering and Assessment
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S. Naas, Director - Nuclear Engineering
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T. Noonan, General Manager - Nuclear Engrg. and Safety Unit l
K. Ostrowski, Unit 1 Operations Manager.
M. Pavlick, Manager - Maintenance Planning and Administration
F. Schuster, Operations i
B. Sepelak, Licensing Engineer
G. Shildt, Supervisor - System Engineering
J. Sieber, Senior Vice President - Duquesne Light Company
-l H. Siegel, Director - Nuclear Engineering Department
W. Sikorski, Director - Inservice Inspection
P. Slifkin, System Engineer (Service Water)
D. Spoerry, Division Vice President - Nuclear Operations
M. Testa, Sr. Engineer G. Thomas, Vice President - Nuclear Services
N. Tonet, Manager, NSD
B. Zini, Engineering Supervisor U.S. Nuclear Reculatory Commission
- K. Battige, Reactor Engineer L. Rossbach, Senior Resident Inspector
- P. Sena, Resident Inspector Attended Exit Meeting on January 15, 1993