IR 05000334/1993006
| ML20044G269 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/05/1993 |
| From: | Carrasco J, Gray E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20044G264 | List: |
| References | |
| 50-334-93-06, 50-334-93-6, 50-412-93-06, 50-412-93-6, NUDOCS 9306020253 | |
| Download: ML20044G269 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION l
REGION I
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Report Nos.
50-334/93-06 and 50-412/93-06
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Docket Nos.
50-334 and 50-412 License Nos.
Duquesne Light Company
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Post Office Box 4
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Shionincoort. Pennsv1vania 15077 Facility Name:
Beaver Valley Units 1 and 2
Inspection At:
Shiopingoort. PA
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Inspection Conducted:
April 5-9.1993
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Inspector:
_J. E. Md, Reactor Engineer, Date
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Materials Section, EB, DRS Approved by:
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E. H.- Gray, Chief, hTaterials Section, Date EB,DRS
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gDR ADOCK 05000334 PDR
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Inspection Summary:
Areas Insoccted: A safety inspection was conducted to determine the licensee's actions to
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address the anomalies observed with the auxiliary feedwater (AFW) system for Beaver Valley 2 (BV-2) Power Station, including Licensee Event Report (LER) No.
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50-412/93-001-00, " Design Stress for the Auxiliary Feedwater System Exceeded Due to
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Water Hammer." In addition, the inspector reviewed the licensee's actions to address a Unit I condition outside the design basis due to MSIV closure not considered in the original design (LER No. 50-334/93 @ l-00).
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Results: Based on the review of assumptions, methodology and the results of the licensee's calculations, it was concluded that the operability analysis for line "C" of the AFW Unit 2 is
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adequate until the next refueling outage when the AFW piping loop "C" would be restored to the FSAR design basis. The AFW system walkdown and the Inservice Testing (IST) of the AFW check valves revealed that the licensee is taking the proper steps to monitor the lines to identify conditions that could cause a water hammer.
The licensee has followed the guidelines established in the generic Letter 91-18 to demonstrate that the Beaver Valley, Unit 1, Main Steam piping and pipe supports are operable until the next refueling outage.
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DETAILS 1.0 PURPOSE AND SCOPE The purpose of this inspection was to review the root causes and the licensee's corrective actions to address the anomalies observed with the auxiliary feedwater (AFW) system for Beaver Valley 2 (BV-2) Power Station, including Licensee Event Report (LER) No.
50-412/93-001-00, " Design Stress for the Auxiliary Feedwater System Exceeded Due to Water Hammer." In addition, the inspector reviewed the licensee's actions to address a
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Unit I condition outside the design basis due to MSIV closure not considered in the original design (LER No. 50-334/93-001-00).
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2.0 REVIEW OF CORRECTIVE ACTION FOR BV-2 AUXILIARY FEEDWATER SYSTEM ANOMA. LIES (37700)
At BV-2 several auxiliary feedwater events have occurred since the unit went operational in 1987. In their root cause analyses, the licensee identified two probable causes. These were water hammers due to column rejoining or steam bubble collapse and overheating of piping and penetration of the AFW system due to intra-system recirculation.
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2.1 Auxillary Feedwater System Piping Water Hammers
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The auxiliary feedwater piping and pipe supports are classified as QA Category I, Seismic, Safety Class II. The piping was designed to the requirements of the ASME Boiler and Pressure Vessel Code,Section III,1971 Edition, including addenda through Winter 1972.
The supports were designed to the guidelines established in the 7th edition of the AISC l
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Manual of Steel Construction,1970.
2.1.1 Review of Fluid Transient Forcing Function (Calc. No. 849-FA)
The inspector reviewed calculation 10080-DLC(B)-849-FA entitled waterhammer analysis of the auxiliary feedwater system inside containment resulting from AFW pump start-up with a voided line. The objective of this calculation was to develop fluid transient forcing functions
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on the BV-2 AFW piping inside the containment, specifically lines "B" and "C", for input to a dynamic pipe stress analyses. In accordance with their FSAR the licensee used the WATHAM computer program to predict the forcing functions acting on the AFW piping assuming a 1 millisecond stoppage of fluid flow. The inspector verified that the hydraulic model and the input parameters reflected the actual conditions.
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2.1.2 Review of the Evaluation of the Auxiliary Feedwater Piping for Postulated Water
IIammer (Cale. No. 848-XF)
On January 26,1993, the licensee's results of the Beaver Valley Power Station Unit 2 design -
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stress analysis for the AFW piping to the "C" steam generator, for the combined water hammer and seismic events, exceeded the design stress allowable. The licensee properly notified to the NRC through LER 412/93-001-00. The licensee hypothesized that the i
"C" AFW line has been subjected to water hammer due to steam pocket formation (voiding)
and steam bubble collapse (LER 50-412/93-001-00).
Findings
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Based on pressure monitoring of loops A, B, and C of the AFW piping, only loop C has experienced a depressudzation due to check valve leakage and is subjected to water hammer
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loads. Therefore, the licensee performed calculation 10080-DLC(P)-848-XF to provide justification for continued operation of the AFW system until the long-term corrective actions are implemented during the next refuleing outage. In this calculation, the licensee evaluated the piping, pipe supports, pipe supports attachments, pipe nozzle connections, piping penetrations, for the maximum fluid transient loads and stresses for the loop C piping. Due l
to similarity between the loop A and B piping configuration, Loop B was evaluated in detail
for fluid transient event.
The time history evaluation of the loops B and C was performed using NUPIPE-SW computer program, modified to include the time history evaluation obtained in calculation
No.10080-DLC(B)-849-FA.
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In this calculation, the stress values from the time history load case were added to equation
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(9a and 9b normal / upset and faulted) of the ASME Edition 1971 Winter 1972 Addenda for comparison to the Code allowables.
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For the resultant stresses, according to Generic Letter 91-18, the licensee had the option to
use the criteda in Appendix "F" of the Section III of the ASME Code, which for the AFW
piping material SA106, Grade B, and penetration material SA333, Grade 6 stress l
corresponding to 3.0 Sm. However, the licensee was able to demonstrate operability by the use of the faulted condition stress allowable limit of 36,000 psi. The inspector verified that the use of this allowable in accordance with the ASME Code. The inspector had no further questions in this regard.
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Conclusion
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Based on the review of assumptions, methodology and the results of calculation 10080-
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DLC(P)-848-XF, the inspector concluded that the licensee's analysis demonstrates that the system is operable until the next refueling outage when the AFW piping loop "C" will be
restored to the FSAR design basis, i
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2.2 Instra-System Recirculation Intra-system recirculation is an indication ofleakage of several valves in series on the same flow path. The licensee explained that for this scenario to take place, concurrent check valve back leakage of the 2FWE 43A or B, and 2FWE-100 (noted that system back leakage is internal to the system). Intra-system flow through the AFW exists due to a difference in pressure between the "B" and "C" main feedwater lines to the steam generators. In this particular case, the pressure in the "B" line had exceeded that in the "C" line. Evidence of this condition was given in Incident Report 2-89-089 describing elevated temperature in the
"B" to "C" loop flow path outside containment.
Intra-system recirculation in and of itself would not create a water hammer transient, but since the AFW piping was designed as cold piping system, the elavated temperature could create a condition outside the design bases.
Finding To address the Incident Report 2-89-089, the licensee performed calculation 10080-NP(F)-
825. This calculation was performed only to reflect the conditions described in the incident report. Therefore, seismic loads were not included. The inspector verified that the maximum calculated stresses for the piping using the ASME,Section III, Equation (10)
expansion stress, did not exceed the allowable stress. The supports were also found adequate, since the design load was larger than the applied load.
Conclusion The inspector verified that upon analytical simulation of the thermal conditions stated in the incident report, that the piping, pipe supports and penetrations met the ASME,Section III allowables for Equation (10).
2.3 Walkdown of the Auxiliary Feedwater Piping and Inservice Testing of Check Valves The inspector and the licensee's ISI engineer walked-down the loops A, B and C outside containment. The highlights of the walkdown are summarized as follows:
Auxiliary Feedwater check valve 2FWE-43B has a temporary clamp seal. The clamp seal was installed to stop body to bonnet leakage. Licensee stated that the bonnet leakage was significant enough to contribute to depressurization of the "B" AFW line.
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S Thermocouples are installed on the "A," "B" and "C" AFW lines at the containment penetration. Temperature indication is provided outside of the recirculation spray cubicles for the operators. The accessibility of the temperature indicators allows the operator to check AFW piping temperature once/ shift. Also, since the thermocouples are mounted on piping near the containment penetrations, the operators would see changes in piping temperature rather quickly.
O Pressure gauges (3) are installed between the 40 series check valves (2 FWE 42 A and B,2 FWE 43 A and B, and 2 FWE 44 A and B) and 2FWE-99,100 and 101.
These gauges showed that the "A" and "B" AFW lines are pressurized and the "C" ARV line is depressurized (refer to Figure 1).
Pressure gauges (2) are installed upstream of the 42A & B,43 A & B, and 44 A & B check valves in the "A" and "B" AFW header. These gauges along with the gauges downstream of the 42A & B,43 A & B, and 44 A & B check valves verify the pressure changes across these check valves.
In addition, the inspector reviewed and verified the Inservice testing (IST) program for the pertinent AFW check valves 2RVE*42A&B of loop "A"; 2FWE*43A&B of loop "B" and 2RVE*44A&B of loop "C," including 2FWE-99 loop "A," 2FWE-100 of loop "B" and 2FWE-101 of loop "C." For this IST, the ASME Code,Section XI, requires quarterly full stroke testing of the check valves. IWV-3412(a) states, valves that cannot be exercised during plant operation shall be identified and full-stroke exercised at cold shutdowns. This relief request was formalized under cold shutdown justifications 40 and 41 for 2FWE*42A,B,43A,B44 A,B,99,100,101. In addition,2FWE*99,100,101 are full stroked in the reverse direction to ensure the proper closing of the check valve by a leak test at refueling as approved by the NRC via Relief Request 19. At present, operators also monitor
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the three penetrations every shift for back leakage by the use of thermocouples. The inspector had no further questions in this regard.
Conclusion Based on the AFW system walkdown and the IST of the check valves, the inspector concluded that the licensee is taking the proper steps to monitor the AFW lines to identify conditions that could cause a water hammer. The instrumentation would be also used as a tool in verifying the valve upgrades and/or replacements have been effective.
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i 3.0 REVIEW OF BV-2 OPERABILITY ANALYSIS FOR TIIE MAIN STEAM
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PIPING i
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Unit I was performing an engineering analysis to support the possibility of increased steam generato1 tube plugging during the upcoming refueling outage. During the analysis, it was i
found that an inadvertent main steam isolation valve (MSIV) closure would generate a pressure transient which would result in increased loading on the main steam system piping j
and supports. This transient was not previously considered as a design condition for the plant. The licensee performed an assessment of the increased loading and this assessment
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has shown that the established design stress limits were exceeded. The licensee has performed an operability assessment discussed below (LER 50-334/93-001-00).
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Findings
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For this steam hammer analysis, time history forcing functions were developed for the main steam piping due to a new postulated transient scenario on MSIV closure. The licensee calculation No. 46 also revised stop valve closure forcing functions that were determined using more realistic steady state initial conditions. These time history stresses due to the MSIV closure were added to the maximum existing sustained and seismic stresses from i
design calculations. These were compared to the allowables stress values established in accordance with the requirements of the appendix B of the FSAR.
The inspector reviewed the results of the licensee's pipe stress analysis (calculation 8700-
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DMC-2844), which shows that the combined MSIV closure stresses and seismic stresses for l
the main steam piping are within allowable stress level for material A155 CL1, GR CMS 75 i
at 600*F (ANSI B31.1,1967).
It was demonstrated throughout this calculation that the main steam piping system's structural
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integrity and operability have not been compromised as a result of increased loads and stresses.
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In terms of the main steam pipe supports, three supports were determined to be functional until the next refueling outage since the stress intensity considering the increased loads and minimum normal stress values is less than the faulted allowable of 2.4 Sh. The remaining
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three supports were determined to be functional based on ASME Section III, Appendix "F" allowable stresses. During the course of the inspection and at the time of the exit meeting, it was noted that in accordance with the Generic 12tter 91-18 (operability criteria), these supports would be upgraded to bring them all within code stress allowables at the end of next refueling outage.
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Conclusion
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The inspector concluded that the licensee has followed the guidelines established in the i
Generic Letter 91-18 to demonstrate that the Beaver Valley, Unit 1, Main Steam piping and pipe supports are operable until the next refueling outage. The inspector has no further
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questions in this regard.
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4.0 MANAGEMENT MEETLNGS
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Licensee management was informed of the scope and purpose of the inspection at the beginning of the inspection. The findings of the inspection were discussed with the licensee management at the April 9,1993, exit meeting. See Attachment I for attendance.
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ATTACHMENT 1
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PERSONS CONTACTED Ducuesne Light C7mpany
- G. S. Thomas, Vice-president Nuclear Services
- D. Spoerry, Vice-president Nuclear Operations
- M. Siegel, Manager Nuclear Engineering Division (NED)
- N. R. Tonet, Manager Nuclear Safety Division (NSD)
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- K. E. Halliday, Director, Electrical Engineering R. L. Hansen, Director NED
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- M. F. Testa, Senior Engineer B. Zini, Supervisor NED
- R. Garver, System Engineer B. F. Sepelak, Licensing Engineer J. Baumler, Quality Services, Director A. Mizia, Quality Services F. J. Lipchick, Senior Licensing Engineer U.S. Nuclear Reculatory Commission
- L. Rossbach, SRI, BV-1
- R. Rasmussen, Reactor Engineer
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- P. Sena, Resident Inspector
- S. Greenlee, Resident Inspector B. Lazarus, Section Chief, DRP Asterisk (*) denotes those present at the exit meeting.
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