ML051750004
| ML051750004 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/24/2005 |
| From: | Cobey E NRC Region 1 |
| To: | Bently D Bently Pressurized Bearing Co |
| Shared Package | |
| ML051460002 | List: |
| References | |
| CAL-1-05-001, ir-04-005, TIA-2004-006 | |
| Download: ML051750004 (15) | |
Text
June 24, 2005 Mr. Donald E. Bently, P.E.
Chairman and Chief Executive Officer Bently Pressurized Bearing Company 1711 Orbit Way Minden, NV 89423
SUBJECT:
HOPE CREEK NUCLEAR GENERATING STATION B REACTOR RECIRCULATION PUMP
Dear Mr. Bently:
In a letter dated May 18, 2005, to Mr. William Levis of Public Service Enterprise Group (PSEG) and copied to the United States Nuclear Regulatory Commission (NRC), you expressed concern that the Hope Creek Nuclear Generating Station B reactor recirculation (RR) pump shaft may be broken or severely cracked. You stated that the B RR pump should be stopped and checked for a shaft crack or other major malfunction.
During a telephone call on May 20, 2005, with Mr. Eugene Cobey, Chief, Projects Branch 3 and Mr. Raymond Lorson, Chief, Plant Support Branch 1 of the NRC Region I, you discussed the content of your letter. During this discussion you indicated that you had no direct knowledge of the specific condition of the B RR pump and did not have direct access to the reports and documents concerning the B RR pump. You also were not aware that PSEG had installed a Bently Nevada vibration monitoring system capable of monitoring the secondary harmonic parameters recommended by your letter. Based on the above telephone discussion we determined that your letter did not contain any new technical information that would indicate a need for us to take additional action regarding this pump.
In January 2005, the NRC completed a review of PSEGs assessment of the condition of this pump and their preparations for mitigating actions needed if this pump should further degrade.
The NRC concluded PSEGs enhanced vibration monitoring program provides reasonable assurance that a potential crack in the RR pump shaft can be detected in time for operators to take appropriate actions to reduce pump speed and remove the pump from service prior to a complete shaft failure. The NRC therefore concluded that operation of the pump for one more operating cycle (approximately 18 months) does not represent an unacceptable increase in the probability of a shaft failure leading to a small loss of coolant accident event. On January 11, 2005, the NRC issued Confirmatory Action Letter (CAL) 1-05-001 to confirm PSEGs commitments to implement the enhanced vibration monitoring program to continuously monitor the B RR pumps primary and secondary harmonic parameters during periods of plant operation until an inspection of the B RR pumps rotating assembly and replacement of the pump shaft have been completed.
For your convenience, we have enclosed our Confirmatory Action Letter (1-05-001) to PSEG and Enclosure 2 to NRC Inspection Report 05000354/2004005, which documents our technical review and conclusions regarding the B RR pump. In an Attachment to this letter we have provided a list of other publically available documents associated with our review of this pump.
Mr. Bently 2 We appreciate your interest in this matter. If you have any specific concerns regarding our technical conclusions and actions with regard to this pump, feel free to contact me.
Sincerely,
/RA/
Eugene W. Cobey, Chief Projects Branch 3 Division of Reactor Projects
Enclosures:
Confirmatory Action Letter (1-05-001), dated January 11, 2005 NRC Inspection Report 05000354/2004005 (Enclosure 2, only) Response to Task Interface Agreement - TIA 2004-006, dated January 12, 2005
Attachment:
List of Publically Available Documents
Mr. Bently 3 Distribution:
S. Collins, RA B. Holian, DRP E. Cobey, DRP B. Welling, DRP J. Wiebe, DRP M. Gray - NRC Resident Inspector K. Venuto, DRP - Resident OA Region I Docket Room (with concurrences)
DOCUMENT NAME: E:\Filenet\ML051750004.wpd SISP Review Complete: _____________ (Reviewers Initials)
After declaring this document An Official Agency Record it will / will not be released to the Public.
To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE RI/DRP RI/DRP NAME JWiebe ECobey DATE 06/22/05 06/23/05 OFFICIAL RECORD COPY
Attachment List of Publically Available Documents Sargent & Lundy, LLC, Independent Assessment of Hope Creek Reactor Recirculation System and Pump Vibration Issues, dated November 12, 2004. Available in the NRCs Agencywide Documents Access and Management System (ADAMS) on the NRCs website at http://www.nrc.gov/reading-rm/adams.html under accession number ML050030176.
Hope Creek Generating Station, Questions Regarding B Recirculation Pump High Vibrations and High Pressure Coolant Injection System Exhaust Damage, dated December 13, 2004.
Available in the NRCs Agencywide Documents Access and Management System (ADAMS) on the NRCs website at http://www.nrc.gov/reading-rm/adams.html under accession number ML043480164.
Startup Readiness, Hope Creek Generating Station, Facility Operating License NPF-57, Docket No. 50-354, dated December 29, 2004. Available in the NRCs Agencywide Documents Access and Management System (ADAMS) on the NRCs website at http://www.nrc.gov/reading-rm/adams.html under accession number ML050050466.
Restart Readiness, Hope Creek Generating Station, Facility Operating License NPF-57, Docket No.60-354, dated January 4, 2005. Available in the NRCs Agencywide Documents Access and Management System (ADAMS) on the NRCs website at http://www.nrc.gov/reading-rm/adams.html under accession number ML050060214.
Flowserve Response to NRC Questions about Hope Creek Generating Station B Recirculation Pump, dated January 12, 2005. Available in the NRCs Agencywide Documents Access and Management System (ADAMS) on the NRCs website at http://www.nrc.gov/reading-rm/adams.html under accession number ML050120303.
Restart Readiness, Supplemental Information, Hope Creek Generating Station, Facility Operating License NPF-57, Docket No.60-354, dated January 7, 2005. Available in the NRCs Agencywide Documents Access and Management System (ADAMS) on the NRCs website at http://www.nrc.gov/reading-rm/adams.html under accession number ML050100286.
PSEG Actions in Response to NRC Concerns Regarding 'B' Reactor Recirculation Pump, Hope Creek Generating Station, Docket No. 50-354, dated January 9, 2005. Available in the NRCs Agencywide Documents Access and Management System (ADAMS) on the NRCs website at http://www.nrc.gov/reading-rm/adams.html under accession number ML050100288.
Hope Creek Nuclear Generating Station - Summary of Results of NRC Review of Technical Issues, dated January 10, 2005. Available in the NRCs Agencywide Documents Access and Management System (ADAMS) on the NRCs website at http://www.nrc.gov/reading-rm/adams.html under accession number ML050100194.
January 11, 2005 CAL No. 1-05-001 Mr. A. Christopher Bakken, III President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P. O. Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
CONFIRMATORY ACTION LETTER (1-05-001)
Dear Mr. Bakken:
On December 17, 2004, the NRC held a meeting at NRC Headquarters with Mr. M. Gallagher and other PSEG representatives to discuss questions that the NRC had regarding the vibration levels of the B reactor recirculation pump at Hope Creek. The vibration levels on this pump have been about twice the levels seen on the A recirculation pump. The vibration levels have been attributed to slight bowing of the pump shaft in the area below the seal package area which has led to frequent seal replacements. Also, industry operating experience indicates that some cracking is likely to be present in the pump shaft, which leads to questions about the expected remaining service life of the shaft. The recirculation pump forms part of the reactor coolant system boundary, and the NRC requires high reliability of that boundary during periods of plant operation.
During the December 17, 2004 meeting, you also discussed the findings of a review done for you by Sargent and Lundy (S&L) who independently assessed this vibration problem. The S&L assessment was summarized in the report, Independent Assessment of Hope Creek Reactor Recirculation System and Pump Vibration Issues, dated November 12, 2004. The NRC had reviewed the S&L report and developed a number of questions which were provided to PSEG during December 2004, and which were discussed during the aforementioned December 17, 2004, public meeting and several subsequent teleconferences. You also addressed those issues in subsequent submittals sent to the NRC on December 29, 2004, January 4, 2005, and January 7, 2005.
The S&L Report had concluded that there is no immediate need to replace the B pump rotor during the current refueling outage, and the pump could be returned to service for the next operating cycle given the current level of reactor recirculation pump and piping vibrations.
However, S&L recommended that both pumps be closely monitored for vibrations.
During various telephone conversations with the NRC staff in late December 2004, and early January 2005, your staff committed to a number of actions that would be taken to ensure acceptable operation of the B recirculation pump. These commitments are described in a subsequent letter you sent to the NRC on January 9, 2005. As a result of another telephone conversation that I had with you on January 10, 2005, it is our understanding that you have
Mr. A. C. Bakken 6 taken (or will take) all of the actions set forth in your January 9, 2005 letter, consistent with the schedule set forth therein. These commitments included: implementing a vibration-monitoring program to continuously monitor the B reactor recirculation pumps primary and secondary harmonic parameters (total amplitude, 1X and 2X amplitude, and 1X and 2X phase angle) during future periods of plant operation; establishing objective criteria that demonstrate that monitored parameters are within an acceptable range; developing procedures which specify the actions to be taken if the monitored parameters are outside of the specified range of the acceptance criteria [the procedures HC.OP-AB.RPV-0003(Q), HC.OP-AR.ZZ-0008(Q)
Attachment E-4, and HC.ER-AP.BB-0001(Z) Rev. 0 were provided by PSEG in a letter dated January 4, 2005]; continuing this program until an inspection of the B reactor recirculation pumps rotating assembly and replacement of the pump shaft have been completed; notifying the NRC prior to implementing any change to this vibration monitoring and operating procedures cited above, to provide sufficient time for the NRC to complete a review of the proposed changes; and replacing the B reactor recirculation pump shaft and inspecting the pumps rotating assembly and pressure boundary components (such as the pump casing, etc.)
at the earlier of the next refueling outage (RFO13) or during an outage of sufficient duration to accomplish pump replacement.
Pursuant to Section 182 of the Atomic Energy Act, 42 U.S.C. 2232, you are required to:
- 1) Notify me immediately if your understanding differs from that set forth above;
- 2) Notify me if for any reason you cannot complete the actions within the specified schedule and advise me in writing of your modified schedule in advance of any change; and
- 3) Notify me in writing when you have completed all of the actions addressed in this Confirmatory Action Letter.
Issuance of this Confirmatory Action Letter does not preclude issuance of an order formalizing the above commitments or requiring other actions on the part of the licensee; nor does it preclude the NRC from taking enforcement action for violations of NRC requirements that may have prompted the issuance of this letter. In addition, failure to take the actions addressed in this Confirmatory Action Letter may result in enforcement action.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, your submittals referenced herein (including your January 9, 2005, letter), and your response, will be made available electronically for public inspection in the NRC Public Document Room or from the NRCs document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction. If personal privacy or proprietary Information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for withholding confidential commercial or financial
Mr. A. C. Bakken 7 information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
Sincerely,
/RA/
Samuel J. Collins Regional Administrator Docket No. 50-354 License No. NPF-57 cc w/encl:
M. Brothers, Vice President - Site Operations J. T. Carlin, Vice President - Nuclear Assessment M. Gallagher, Vice President - Engineering and Technical Support W. F. Sperry, Director - Business Support C. Perino, Director - Nuclear Safety and Licensing J. A. Hutton, Hope Creek Plant Manager R. Kankus, Joint Owner Affairs J. J. Keenan, Esquire M. Wetterhahn, Esquire Consumer Advocate, Office of Consumer Advocate F. Pompper, Chief of Police and Emergency Management Coordinator J. Lipoti Ph.D., Assistant Director of Radiation Programs, State of New Jersey K. Tosch - Chief, Bureau of Nuclear Engineering, NJ Dept. of Environmental Protection H. Otto, Ph.D., DNREC Division of Water Resources, State of Delaware N. Cohen, Coordinator - Unplug Salem Campaign W. Costanzo, Technical Advisor - Jersey Shore Nuclear Watch E. Zobian, Coordinator - Jersey Shore Anti Nuclear Alliance
Enclosure 2 January 12, 2005 MEMORANDUM TO: Wayne D. Lanning, Director Division of Reactor Safety Region I FROM: James E. Lyons, Deputy Director /RA/
Division of Licensing Project Management Office of Nuclear Reactor Regulation
SUBJECT:
RESPONSE TO TASK INTERFACE AGREEMENT - TIA 2004-006, REQUEST FOR EVALUATION OF SARGENT AND LUNDY REPORT ON HOPE CREEK GENERATING STATION B REACTOR RECIRCULATION PUMP AND PSEG NUCLEAR, LLC EVALUATION OF HIGH PRESSURE COOLANT INJECTION SYSTEM EXHAUST SNUBBERS (TAC NO. MC5111)
By letter dated December 13, 2004, Region I submitted TIA 2004-006 requesting assistance from the Office of Nuclear Reactor Regulation (NRR) in reviewing PSEG Nuclear, LLCs (PSEG or the licensee) resolution of technical concerns related to the reactor recirculation pump and the high pressure coolant injection (HPCI) system exhaust line. Region I requested NRR review of three specific items.
The first item was to review the Sargent and Lundy (S&L) report, Independent Assessment of Hope Creek Reactor Recirculation System and Pump Vibration Issues, and determine whether operation of Hope Creek Generating Station (Hope Creek) over the next operating cycle represents an unacceptable increase in the probability of a recirculation pump shaft failure or a small break (i.e. seal) loss-of-coolant accident (LOCA) event. The staffs review of the S&L report found that it did not provide sufficient information to completely address the concern.
The licensee subsequently provided additional information to the Nuclear Regulatory Commission (NRC) staff; however, based on its review of the technical information provided by the licensee, the NRC staff concludes that the probability of a pump shaft failure of RR pump B during the next cycle of operation is indeterminate. The licensee proposed enhanced vibration monitoring of the reactor recirculation pumps. The NRR staff found that there is reasonable assurance that the licensees enhanced vibration monitoring program can detect a potential crack in the reactor recirculation pump shaft in time to take appropriate actions to reduce pump speed and remove the pump from service prior to a complete shaft failure. Thus, the NRR staff considers that operation of the recirculation pump for one more cycle does not represent an unacceptable increase in the probability of a shaft failure leading to a small LOCA event. The details of NRRs assessment are contained in Attachment 1.
The second item was to review the S&L report and determine whether PSEGs decision to not perform the recirculation pump shaft inspections for potential shaft cracking as described in General Electric (GE) Service Information Letter (SIL) 459 represents an unacceptable increase in the probability of a recirculation pump shaft failure or small break (i.e. seal) LOCA event.
The licensees survey of the industry indicates that a number of recirculation pumps have
successfully operated well past the inspection interval proposed in SIL 459. The purpose of the inspection recommended in SIL 459 was to detect a potential crack in the recirculation pump shaft. The NRR staff found that there is reasonable assurance that the licensees enhanced monitoring program can detect a potential crack in the reactor recirculation pump shaft in time to take appropriate actions to reduce pump speed and remove the pump from service prior to a complete shaft failure. Thus, the NRR staff concludes that PSEGs decision not to perform the pump shaft inspection as recommended in GE SIL 459 does not represent an unacceptable increase in the probability of a shaft failure leading to a small LOCA event. The details of NRRs assessment are contained in Attachment 1.
The third item was to provide a technical assessment of PSEGs engineering evaluation for the failed HPCI system steam exhaust line snubbers and determine whether it provides an adequate basis for the operability of the HPCI system per GL 91-18. The NRR staff found that the licensees evaluation provides an adequate basis for the operability of the HPCI system per GL 91-18. The details of NRRs assessment are contained in Attachment 2.
Principal Contributors: J. Fair W. Poertner S. Unikewicz Docket No. 50-354 cc w/ attachments: C. Casto C. Pederson D. Chamberlain
Reactor Recirculation Pump Vibration Review
Background
The B Hope Creek Generating Station (Hope Creek) reactor recirculation (RR) pump has had a historical problem involving high vibration levelsabout double those on the A RR pump.
Past PSEG Nuclear, LLC (PSEG or the licensee) actions to reduce the vibration levels have not been effective. The high vibrations have been attributed, in part, to a slight bowing of the shaft in the area below the seal package area. The vibrations have led to frequent seal replacements (1.5-year intervals versus the expected 6-year intervals).
In addition to the bowing, the A and B RR pump shafts are expected to have some degree of thermally induced stress cracking based on industry operating experience described in General Electric (GE) Service Information Letter (SIL) 459. GE SIL 459 recommends three actions to address this problem: vibration monitoring, shaft inspections after about 80,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation and action to mitigate the thermal stress initiators. Hope Creeks RR pumps have over 130,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation, and PSEG has not performed the recommended inspections.
In addition to the pump vibrations, there are vibrations on the associated RR and residual heat removal system piping which have resulted in damage to system sub-components (motor operated valve handwheel and limit switches). To date, none of the vibration-induced component problems have rendered any safety-related system inoperable.
Sargent and Lundy (S&L) performed an independent assessment for PSEG which concluded that return of Hope Creek to service for the next operating cycle was acceptable given the current level of RR pump and piping vibrations. S&Ls conclusion was based upon data which indicated that the vibration level for Hope Creeks B RR pump was consistent with RR pumps at other facilities and also based on an assumption that operators would be able to respond to an increasing vibration trend and take action to remove the pump from service prior to shaft failure.
The S&L assessment is summarized in the report, Independent Assessment of Hope Creek Reactor Recirculation System and Pump Vibration Issues, dated November 12, 2004. The NRC staff reviewed the S&L report and developed a number of questions which were provided to the licensee on December 1, 2004. PSEG responded to the questions during a December 17, 2004, public meeting with the Nuclear Regulatory Commission (NRC). PSEG provided additional responses to the NRC staffs questions in letters dated December 29, 2004, January 4, 2005, January 7, 2005, and January 9, 2005. In addition, numerous teleconferences were held between PSEG and the NRC in December 2004 and January 2005 to discuss the B RR pump vibration issue.
The S&L report concluded that there is no immediate need to replace the B pump rotor during the current refueling outage. S&L recommended that both pumps be monitored for vibrations and that a rapid rise in vibrations would be a sufficient reason to shut the pump down immediately for an internal inspection and shaft replacement, as the window between the rise in vibration and potential shaft failure is expected to be small.
PSEG also provided additional background information in Report H-1-BB-MEE-1878, Hope Attachment 1
Creek B Recirculation Pump Vibration Analysis, Revision 1, dated December 16, 2004. The report concluded that, while the B RR pump has elevated vibrations when compared to the industry average, these vibration levels are not detrimental to the operation or reliability of the pump. The report also indicated that, although the risk of an RR pump shaft cracking event during any given cycle cannot be quantified, the operating experience of 29 RR pumps in operation longer than the Hope Creek B RR pump provides sufficient data to conclude that the risk of a shaft cracking event during the next cycle is minimal.
NRC Staff Review The NRC staffs review focused on the following key issues regarding the RR pump operation:
(1) Does PSEG have a technical evaluation which shows that the RR pumps can be operated for another cycle without failure of the shafts considering the identification of shaft cracks that have been observed at other facilities with the same design RR pumps?
(2) Can PSEG provide data which demonstrates that shaft cracks have been detected at other facilities with the same design RR pumps using vibration monitoring? Can the cracks be detected in time for the operators to take appropriate actions?
(3) What are the consequences of an RR pump failure during plant operations?
GE SIL 459 indicates that all Byron Jackson RR pump shafts inspected have shown some degree of thermally-induced cracking. The cracking occurs near the pump thermal barrier where mixing of cold seal purge system water and the hot reactor coolant water occur. The cracks initiate as axial cracks in the pump shaft. The licensee indicated that, if the cracks remain axial, the cracks will grow slowly and not affect the operation of the pump. However, the licensee also indicated that given sufficient mechanical loads, the cracks can become circumferential. The circumferential cracks can propagate to shaft failure under mechanical loading. The time it takes to transition from slow growing axial cracks to more rapidly growing circumferential cracks depends on the magnitude of the mechanical loads acting on the pump shaft. Since the licensee does not know the magnitude of the mechanical loads, it is difficult to predict the shaft life based on the magnitude of the operational loads.
The licensee cited operating experience of other boiling water reactors (BWRs) with similar Byron Jackson RR pumps. The licensee indicates that the age of the Hope Creek RR pumps is about average for the pumps of similar design at other BWRs. The NRC staff notes that a number of the older pumps included in the licensees comparison are much smaller than the Hope Creek pumps. While the operating experience provides some confidence that the pumps can be safely operated beyond the time interval recommended in GE SIL 459, the crack growth analyses provided by the licensee indicates that the time is highly dependent on the magnitude of the mechanical loads, which is not well known.
The licensee also provided the level of vibration recorded at other BWRs with similar Byron Jackson RR pumps. The licensee concluded that measured vibration levels of the Hope Creek Attachment 1
RR pumps are within the range of the vibration levels measured at other BWRs. However, the level of vibration of the B pump is toward the high end of the range of vibration levels measured at other BWRs. Therefore, the B pump is experiencing higher vibratory loadings than most of the pumps in the licensees survey. In addition, the licensee cited a history of problems in its attempt to balance and align the pump shaft. These problems caused additional mechanical loadings on the pump shaft which could increase the potential for circumferential cracks to have developed in the shaft. On the basis of the above discussion, the NRC staff concludes that the probability of a pump shaft failure of RR pump B during the next cycle of operation is indeterminate based on PSEGs evaluation of the potential thermal and mechanical loads on the pump shaft.
The licensee relies on vibration monitoring to detect circumferential cracking of the RR pump shaft with sufficient lead time for operators to secure the pump prior to complete shaft failure.
The licensee developed a plan for monitoring the vibration levels of the RR pumps. The key elements of the plan involve continuous basic monitoring of the overall level of vibration and continuous monitoring of the vibration harmonics for enhanced detection capability of potential shaft cracking.
The licensees continuous basic vibration level monitoring by the operations department consists of a pump vibration alarm and pump speed reduction if the B pump vibration level reaches 11 mils (0.011 inch), and removal from service if the pump vibration level reaches 16 mils (0.016 inch). The continuous monitoring of the vibration harmonics consists of pump vibration alarms and pump speed reduction if the synchronous speed (1X) vibration amplitude, two times synchronous speed (2X) vibration amplitude, 1X phase angle, or 2X phase angle exceed defined allowable limits. If the monitored values do not fall within their allowable limits at the reduced pump speed, the licensee will remove the RR pump from service. The allowable limits are established using the Operations and Maintenance Committee of the American Society of Mechanical Engineers standard, Reactor Coolant and Recirculation Pump Condition Monitoring. The licensee will record baseline data to establish these allowable limits during plant startup. The licensee provided two technical papers in support of the proposed vibration monitoring criteria.
The first technical paper is entitled, Case History Reactor Recirculation Pump Shaft Crack, Machinery Messages, December 1990. The paper discusses the RR pump shaft cracking experience at the Grand Gulf nuclear power plant. The paper indicates that the vibration level increased rapidly over a three-hour period before the pump was secured at slow speed.
Although the shaft did not experience a complete failure, subsequent inspection revealed the shaft was cracked approximately 320 degrees around the circumference. The paper indicates that it is necessary to monitor the 1X and 2X steady state vectors (1X and 2X amplitudes and phase angles) on a continuous basis and to compare these monitored values to an acceptance criteria. The paper also indicates that alarms are necessary to alert the user to amplitude and phase deviations that are outside the acceptance criteria.
The second paper is a technical bulletin from Bently, Nevada, Early Shaft Crack Detection on Rotating Machinery Using Vibration Monitoring and Diagnostics. The technical bulletin indicates that shaft cracking can be detected by monitoring the 1X and 2X vectors.
Attachment 1
The technical bulletin also recommends continuous monitoring of machines that are susceptible to shaft cracking.
These papers recommend using continuous monitoring of the 1X and 2X vectors as a predictive method to detect significant shaft cracking. The NRC staff requested that the licensee provide some evidence that vibration monitoring was effective for detecting shaft cracks in RR pumps similar to the Hope Creek RR pumps. The licensee cited the experience at Grand Gulf discussed above. The Grand Gulf RR pump shafts are hollow shafts as opposed to the solid shafts used in the Hope Creek RR pumps. Therefore, the Grand Gulf experience may not be directly applicable to Hope Creek. The licensee provided additional information which indicates that cracks in reactor coolant pump shafts were identified at Sequoyah (technical presentation to non-destrictive examination Steering Committee by G. Wade, July 12, 2002) and Palo Verde Unit 1 (Palo Verde Nuclear Generating Station Cracked Reactor Coolant Pump Shaft Event, H. Maxwell, 1996) using vibration monitoring. Although these plants are pressurized water reactors (PWRs), the reactor coolant pumps have solid shafts. The licensee indicated that these pumps had operated for a significant period of time after the first indication of shaft cracks by vibration monitoring. The NRC staffs review of related pump shaft vibration concerns also identified that vibration monitoring successfully identified a reactor coolant pump shaft cracking at St. Lucie Unit 2 (licensee event report (LER) Number: 1993-005). The PWR reactor coolant pump experience provides some indication that a solid pump shaft will provide better early crack detection capability than the hollow pump shafts, such as those used at Grand Gulf. PSEG has provided data which demonstrates that shaft cracks in pump shafts similar to those used at Hope Creek have been detected at other facilities, and that these cracks were detected in time for operators to take appropriate actions.
On the basis of the available operating experience, the NRC staff concludes that continuous monitoring of the 1X and 2X amplitudes and phase angles provides reasonable assurance that circumferential shaft cracking can be detected with sufficient time for the plant operators to take appropriate actions. The licensee will either reduce the RR pump speed or remove the pump from service if the monitoring system detects vibration levels that exceed the limits specified in the vibration monitoring plan.
The NRC staff also reviewed the licensees assessment of the potential consequences of an RR pump shaft failure. The RR pump shaft axial cracking that has been reported occurred below the seal area and above the pump hydrostatic bearing. This is the region where a potential RR pump shaft failure would be expected to occur. The pump impeller would be expected to settle at the bottom of the pump casing, which could potentially result in some damage to the pump casing. The unsupported end of the upper part of a broken shaft may damage the shaft seal. A seal failure would result in leakage of reactor coolant through clearances around the upper half of the broken pump shaft. This leakage would be bounded by the design basis small loss-of-coolant event. If such an event were to occur, the licensee would be able to isolate the pump using the RR loop isolation valves, thereby terminating any reactor coolant system leakage.
Attachment 1
Conclusion The NRC staff concludes that the licensees continuous monitoring program for the Hope Creek RR pumps, as discussed above, provides reasonable assurance that a potential crack in the RR pump shaft can be detected in time for operators to take appropriate actions to reduce the pump speed or remove the RR pump from service prior to a complete shaft failure.
Attachment 1
High Pressure Coolant Injection (HPCI) Exhaust Line Review
Background
On November 1, 2004, with Hope Creek Generating Station in Mode 5 for refueling outage 12, tandem snubbers from the HPCI turbine exhaust piping failed during dynamic testing. A followup inspection of the HPCI piping resulted in the observation of a damaged pipe support and a snubber anomaly that could have been the result of a water hammer event in the HPCI turbine exhaust line. A subsequent evaluation by PSEG Nuclear, LLC (the licensee) of the reported observations found that there was no conclusive evidence that a water hammer had occurred in the HPCI turbine exhaust line.
Nuclear Regulatory Commission (NRC) Staff Review The licensee provided an assessment of the tandem snubber failures performed by the snubber manufacturer, Lisega. The snubber failures occurred in the fluid reservoirs. Lisega indicated that the fluid reservoir failures were caused by stuck poppet valves that allowed fluid to leak into the reservoir during testing. Lisega concluded that repeated testing of the HPCI snubbers in compression resulted in over-pressurization of the reservoirs. Lisega also indicated that the snubbers would have functioned in response to a seismic event. The licensees assessments of the other observations, identified during the initial inspection of the HPCI exhaust line, provided reasonable dispositions of the observed conditions.
A licensee inspection of the accessible portions of the HPCI exhaust line in the turbine room and the torus room found no evidence of large pipe distortion or excessive pipe movement at support locations which likely would have been present if a water hammer had occurred. This was confirmed by the NRC inspectors. The licensee also performed non-destructive examination (NDE) of all field welds on the 20-inch HPCI exhaust line. All welds were found to be satisfactory. The inspections and weld examinations performed by the licensee are the type of actions the NRC staff would require after a water hammer event.
Conclusion The licensee provided plausible explanations for the snubber failures that occurred during snubber testing and for the identified support damage and snubber anomaly identified during the followup HPCI inspection. In addition, the licensee performed the type of inspections and NDE examinations that the NRC would require after a water hammer event and found no adverse results. Therefore, the NRC staff concluded that there was reasonable assurance that the integrity of the HPCI exhaust line had not been challenged by a water hammer event.
Attachment 2