IR 05000272/1999006
| ML18107A539 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/20/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18107A538 | List: |
| References | |
| 50-272-99-06, 50-311-99-06, NUDOCS 9909280367 | |
| Download: ML18107A539 (16) | |
Text
Docket No:
License N Report N Licensee:
Facility: *
Location:
Dates:
Inspectors:
Approved By:
9909280367 990920 PDR ADOCK 05000272 G
PDR U.S. NUCLEAR REGULATORY COMMISSION
REGION I
50-272 & 50-311 DPR-70, DPR-75 50-272/99-06, 50-311199-06 PSEG Nuclear LLC Salem Nuclear Generating Station, Units 1 & 2 P.O. Box236 Hancocks Bridge, NJ 08038 July 19 - August 6, 1999 August 9 - August 19, and September 1, 1999 (Region I - lnoffice)
L. Prividy, Senior Reactor Engineer, DRS P. Kaufman, Senior Reactor Engineer, DRS F. Arner, Reactor Engineer, DRS L. Scholl, Senior Reactor Engineer, DRS S. Pindale, Reactor Engineer, DRS H. Nieh, Resident Inspector, Salem 1 & 2, DRP Lawrence T. Doerflein, Chief Engineering Programs Branch Division of Reactor Safety
SUMMARY OF FINDINGS Salem Generating Station, Units 1 & 2 NRC Inspection Report 50-272/99-06, 50-311/99-06 The report includes the results of a team inspection of the auxiliary feedwater system by region based inspector Inspection findings were assessed according to potential risk significance, and were assigned colors of GREEN, WHITE, YELLOW, or RED. GREEN findings are indicative of issues that, while not necessarily desirable, represent little risk to safety. WHITE findings would indicate issues with some increased risk to safety, and which may require additional NRC inspection YELLOW findings would be indicative of more serious issues with higher potential risk to safe performance and would require the NRC to take additional actions. RED findings represent an unacceptable loss of margin to safety and would result in the NRC taking significant actions that could include ordering the plant to shut down. The findings, considered in total with other inspection findings and performance indicators, will be used to determine overall plant performanc Cornerstone: Mitigating Systems
Green.. The acceptance criteria in the auxiliary feedwater (AFW) pump surveillance test procedures was inadequate since it did not ensure that the pump performance was capable of meeting accident analysis performance or pump operability criteria found in the Unit 1 and 2 Technical Specifications. This issue was considered potentially significant due to the allowable degraded core cooling capability associated with the established AFW pump minimum performance acceptance criteria. The issue was considered green in the significance determination process since it did not have an immediate impact on AFW system operability as determined by an evaluation of recent pump performance. Also, the issue has been entered into PSEG Nuclear's corrective action program (CAP). The failure to establish AFW pump surveillance acceptance criteria, which ensured that each pump achieved its minimum design required performance, was a violation of NRC test control requirements~ (NCV 50-272/99-06-01; 50-311/99-06-01) The team noted that the non-conforming condition, whereby surveillance test acceptance criteria was inconsistent with the AFW flow model pump performance assumptions, was not promptly incorporated into PSEG Nuclear's CA This failure was a violation of NRC requirements concerning corrective action. (NCV 5Q-272/99-06-02; 50-311/99-0~-02) (Section 1 R21.1)
Green. The team found that the positioners for the AFW flow control valves had been maintained with parts that were not qualified for use in safety-related applications. This issue was considered significant since the functionality for valves in a risk significant system was being challenged. The qualification issue was considered green in the significance determination process because it did not have an immediate impact on AFW system operability based on an engineering evaluation and the issue was being addressed by PSEG Nuclear's CAP. This issue was considered to be a violation of NRC ii
- design control requirements. (NCV 50-272/99-06-03; 50-311/99-06-03) Although PSEG Nuclear entered the issue into the CAP, t.he team found that this action was not timely and was a violation of NRC requirements concerning corrective action. (NCV 50-272/99-06-02; 50-311199-06-02) (Section 1 R21.1)
Green. The team identified some instances where PSEG Nuclear did not effectively implement the CA.
A prior performance weakness resulted in a governor control oil system needle valve setting for the 23 AFW pump turbine that was not optimal for actual operating condition.
During operation of the 23 AFW pump, the team identified that the turbine outboard bearing temperature exhibited an increasing trend, which represented an operability question, and PSEG Nuclear was not aware of this potential problem. Consequently, this problem was not entered into the CA.
Other examples of ineffective implementation of the CAP were identified, including deficiencies not promptly entered into the CAP or not adequately correcte These issues were considered to be significant in that they indicated an overall CAP
. weakness since problems were noted in not identifying corrective action items and the lack of timeliness and effectiveness of corrective actions. The issues were considered green in the significance determination process because they did not have an immediate impact on AFW system operability and were ultimately included in PSEG Nuclear's CA The issues collectively represented the third example of a violation of NRC requirements concerning inadequate corrective action. (NCV 50-272/99-06-02; 50-311/99-06-02)
(Section 1 R21.2)
Performance Indicator Verification
The team reviewed the safety system unavailability for the high pressure safety injection, auxiliary feedwater, emergency AC power and residual heat removal system performance indicators (Pis) for the first six months of 1999 and found that the data was accurately reported for these systems and met the Pl reporting guidance. (Section 40A2)
iii
1.
REACTOR SAFETY Cornerstone: Mitigating Systems 1 R21 Safety System Design and Performance Capability
.1 Introduction The auxiliary feedwater (AFW) system at Salem Units 1 and 2 was reviewed using Inspection Procedure 71111, Attachment 21. AFW was selected since it is a risk significant mitigating system for responding to transients, such as station blackout, loss of offsite power, and loss of main feedwate Design - Mechanical. Electrical, and Instrumentation and Controls
. Inspection Scope The team reviewed the AFW design and licensing basis documents to determine the system functional requirements during abnormal and accident conditions. For the documents reviewed, which included calculations and analyses, the team verified that the assumptions were appropriate, that proper engineering methods and models were used and that there was an adequate technical basis to support the conclusions. Where possible, the team performed independent calculations to evaluate the document adequacy. The review was performed to determine that: (1) the design basis was in accordance with the licensing commitments and regulatory requirements; (2) the design output documents such as drawings and procurement specifications were correct; and, (3) the installed system and components were tested to verify the design bases were me The team reviewed the Updated Final Safety Analysis Report (UFSAR) to establish the design and licensing basis for the AFW and interfacing systems. The piping and instrumentation drawings, the configuration baseline documents and the installed configuration were also reviewed to assess the capability of the system to satisfy the design intent. The adequacy of surveillance testing to ensure that adequate flow would be supplied to the steam generators during worst case accident conditions was also reviewe The team reviewed changes made to the system to verify that the system met the design and licensing basis in the modified configuration and that the changes did not introduce any unreviewed safety question Observations and Findings The team identified two issues that could have resulted in the inoperability of important AFW equipment. The acceptance criteria in the AFW pump surveillance test procedures was inadequate since it did not ensure that the pump performance was capable of meeting accident analysis performance or pump operability criteria found in the Unit 1 and 2 Technical Specification (TS). Also, the team found that the positioners for the AFW flow control valves had been maintained with parts that were not qualified for use in safety-related applications, thus rendering them as potentially inoperable. These
- potential inoperability issues were considered significant due to (1) the degraded core cooling capability associated with less than required pump performance; and (2)
challenging the functionality of the control valves in a risk significant system. The issues were considered green in the significance determination process because they did not have an immediate impact on AFW system operability as determined by evaluations of recent pump performance and control valve operation. Also, the issues were being addressed by PSEG Nuclear's corrective action progra Pump Design Requirements Technical Specification 4.7.1.2.b.1 requires that each motor driven auxiliary feedwater pump (MDAFWP) be capable of developing a discharge pressure of 1275 psig on recirculation. Technical Specification 4. 7.1.2.b.2 requires that the turbine driven auxiliary feeedwater pump (TDAFWP) develop a discharge pressure greater than or equal to 1500 psig on recirculation flow when secondary steam generator pressure is greater than 680 psig. PSEG Nuclear had verified these requirements during testing in accordance with Section Xi, "lnservice Testing," of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The team noted that in Information Notice 97-90, "Use of Nonconservative Acceptance Criteria in Safety-related Pump Surveillance Tests," the NRG alerted licensees that if minimum acceptable design performance was more stringent th~n the ASME Code acceptance criteria, then the test acceptance criteria must be adjusted to avoid the actual pump performance from being allowed to degrade below the minimum acceptable design performanc The AFW system design for the motor driven pumps incorporates a run-out protection feature to ensure cavitation is minimized under certain accident scenarios, such as faulted steam generators and reduced steam generator back pressure conditions. The system resistance is varied by throttling a flow control valve as a function of the pump discharge pressure. For the 11, 12 and 21 MDAFWPs, a range of 1150 to 1350 psig will throttle the AFW flow control valve from full closed to full open. Since the capability of the 22 MDAFWP is about 150 psi lower at the design point flow of 440 gpm than the other pumps, its control range is from 1000 psig to 1200 psig. The team noted that the allowable 7.5% to 10% reduction in pump performance permitted by the surveillance test procedures would result in the runout protection logic inadvertently sensing the need for flow control valve throttling, when, in fact, the discharge pressure reduction would be due to pump degradation. The effect would be throttling of the flow control valve to the steam generator, resulting in a higher system frictional loss and less flow delivered by the AFW system. In summary, the existing system runout protection results in an additional reduction to system flowrate capability under degraded pump performance condition The team concluded that any allowable pump degradation should be modeled to ensure that minimum design flowrates can be maintained.
..
Pump Performance Testing Review The design output of the AFW flow model contained in calculation 001.6-868 consisted, in part, of the minimum flow requirements for various accident scenarios such as loss of main feedwater and steam line break outside containment transients. The flow model simulated the pumps, piping, valves and other components in the system. The design pump curve (head versus flow) provided by the manufacturer was used to predict pump performanc The team reviewed the AFW pump test results and found the performance of the 22 MOAFWP had degraded about 2.2% from the pump manufacturer's performance curv Independent calculations performed by the team indicated that the 22 MOAFWP might not develop the minimum flows included in the calculation of record. PSEG Nuclear reviewed the data and determined that the AFW system minimum flow performance documented in table 2.0 of calculation 001.6-868 would not be achieved by the 22 MOAFWP at the assumed steam generator backpressures of 1133 psia or the 1150 psig stated in the AFW bases section of the plant TS. PSEG Nuclear performed an operability determination which utilized best-estimate assumptions for steam generator back pressures along with performance data from the last five full flow pump tests and concluded that the 22 MOAFWP would achieve the minimum required flow of 440 gallons per minute (gpm). Similar test reviews indicated a 1. 7% reduction in the 12 MOAFWP performance at full flow conditions. The team had no operability concern regarding the 12 MOAFWP since it had a higher developed head at the pump design point of 440 gp PSEG Nuclear performed additional calculations to determine an acceptable level of pump degradation which would maintain minimum licensing bases flow assumption For the 22 MOAFWP, PSEG Nuclear concluded that a 3:8% allowable level of pump degradation would result in maintaining minimum accident analysis flow assumptions and would replace the existing allowable pump degradation (a nominal 9%) in the surveillance procedure. PSEG Nuclear indicated that similar calculations would be performed for the other MOAFWPs to determine acceptable levels of pump degradatio Concerning the TOAFWPs, the team noted that current testing performed to satisfy TS requirements, permits increasing the speed of the turbine until the 1500 psig discharge pressure TS criterion is satisfied. Thus, an indeterminate amount of degradation would be allowed since it could be compensated for by increasing pump speed until the required 1500 psig pressure was achieved at the recirculation flowrate of 400 gpm. The team reviewed the latest full flow test results for the 13 and 23 TOAFWPs and noted that the 23 pump was about 5.0% degraded from the performance curve developed in calculation S-C-AF-MOC-0430, Revision 0. The team determined that the most recent test data from a May 21, 1999 full flow test (S2. OP-AF-0007) supported the capability of the 23 pump to supply 880 gpm at the best estimate steam generator backpressures included in the 22 MOAFWP operability review noted above. However, the allowable pump degradations currently in the surveillance tests would not allow design TOAFWP flows to be achieved at the speed set for automatic startup (3450 rpm).
Pump Performance Testing Review - Conclusion The team determined that the acceptance criteria in the AFW pump surveillance test procedures did not ensure that pump performance was capable of meeting accident analysis performance or pump operability criteria found in the AFW system bases section of the Salem Unit 1 and 2 Technical Specifications. Specifically, the procedure allowed a span of 7.5% to 10.0% pump performance degradation before considering the pumps to be inoperable or in the required action range. This allowable degradation was inconsistent with assumptions utilized within the PSEG Nuclear's AFW flow model system analysis in that the flow model had not accounted for any pump performance degradation. Although PSEG Nuclear had established in-service testing (IST)
aQceptance criteria that met the requirements specified in the ASME Code, the criteria would have allowed the AFW pumps to degrade below the performance assumed in the accident analysis. This potential inoperability issue was considered significant due to the degraded core cooling capability associated with less than required pump perfomrance in the risk significant AFW system. The issue was considered green in the significance determination process since it did not have an immediate impact on AFW system operability as determined by an evaluation of recent pump performance. The failure to establish AFW pump surveillance acceptance criteria, which ensured that each pump achieved its minimum design required performance, was considered to be the first example of a violation of 10CFR Part 50, Appendix 8, Criterion XI, "Test Control." This violation is being treated as a Non-Cited Violation (NCV), consistent with Appendix F of the NRC Enforcement Policy. (NCV 50-272/99-06-01; 50-311/99-06-01) PSEG Nuclear initiated notification number 20001328 in accordance with the Corrective Action Program (CAP) requirements to review the AFW pump technical specifications, develop a licensing change request to ensure AFW required pump performance is consistent with accident analysis assumptions, and modify AFW surveillance test acceptance criteria accordingl The team noted that the non-conforming condition, whereby surveillance test acceptance criteria was inconsistent with the AFW flow model pump performance assumptions, was not promptly incorporated into PSEG Nuclear's CAP. Specifically, on July 23, 1999, the discrepancy had been identified to PSEG Nuclear personnel. However, notification 20001328 was not entered into the CAP until July 30, 1999, after further subsequent questions and independent NRC calculations identified concerns with the operability of the 22 MDAFWP on July 29, 1999. The team considered this lack of timeliness was significant, considering the challenge to operability of a risk significant system presented by the issue. This issue was considered green in the significance determination process since it had no immediate impact on AFW system operability. This issue was the first example of a violation of NRC requirements concerning corrective action. 10 CFR 50, Appendix 8, Criterion XVI, "Corrective Action," requires that measures shall be established to assure conditions adverse to quality are promptly identified and correcte In response to the discussion held on July 23, 1999, PSEG Nuclear viewed that the current level of pump performance supported design calculation minimum flowrates and actions had been initiated to re-analyze existing flow model calculations to determine acceptable levels of pump degradation. However, the team determined that the lack of a timely notification initiation for the identified discrepancy was inconsistent with the i
J
- expectations documented within PSE&G Nuclear procedure, NC.WM-AP.ZZ-OOOO(Q),
Revision 0, "Notification Process," which stated, in part, that "when in doubt, a notification should be processed." This violation is being treated as a Non-Cited Violation, consistent with Appendix F of the NRC Enforcement Policy. (NCV 50-272/99-06-02; 50-311/99-06-02) PSEG Nuclear has included this issue appropriately into the CAP (Notification 20004451 ).
Air Operated Valve Positioners The team noted that the AFW system design utilized air-operated components to control flow during automatic system operation and for remote manual operation of the flow control valves and the turbine driven pump governor and steam stop valves. The team also found that the design bases of the system considered the control air system to be highly reliable and assumed it to be available following an acciden When responding to questions regarding the safety classification of control air system components, PSEG Nuclear informed the team that the positioners for the flow control valves had been maintained with parts that were not appropriate for use in safety-related applications. Based on an their initial review, PSEG Nuclear determined that the cause of the problem was the downgrading of the purchase class of certain piece parts during a procurement engineering change performed in 1995.
This issue was considered significant since the functionality of the valves in a risk significant system was being challenged. PSEG Nuclear performed an evaluation to assess the potential effect of the non-safety parts on the affected valves operability. This evaluation concluded that the affected valves were operable but that actions were necessary to restore the valves to full qualification as a safety-related component. The team evaluated this condition and concluded that, since the use of non-qualified parts did not affect the valve operability and the issue was being addressed by PSEG Nuclear's CAP, this issue was green in accordance with the criteria of the significance determination process. PSEG Nuclear was also performing an extent of condition review to determine if there are any similar problems with other installation CFR 50, Appendix B, Criterion Ill, "Design Control", requires, in part, that measures be established to ensure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures and instructions. It further requires that measures be established for the selection and review for suitability of application of materials, parts, equipment and processes that are essential to the safety-related functions of the structures, systems and component *
Contrary to the above, PSEG Nuclear had not ensured that the applicable design basis was correctly translated into specifications, drawings, procedures or instructions, and had not ensured that measures were established for the selection and review of materials, in that repair parts for positioners on the flow control valves were not properly selected and reviewed to ensure their suitability for performing a safety-related functio In accordance with Appendix* F of the NRC Enforcement Policy, this violation is being treated as a Non-Cited Violation. PSEG Nuclear has entered this issue into the CAP (Notification #20002187). (NCV 50-272199-06-03;50-311/99-06-03)
/
Although PSEG Nuclear entered the issue into. the CAP, the team found that this action was not timely in that the issue was identified on July 23, 1999, but the corrective action notification was not initiated until August 3, 1999. The team considered that this lack of timeliness was significant, considering the challenge to operability of a risk significant system presented by the issue. The issue was considered green in the significance determination process since it did not have an immediate impact on AFW system operability. This was the second example of a violation of NRC requirements concerning corrective action. 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," requires that measures shall be established to assure conditions adverse to quality are promptly identified and corrected. This violation is being treated as a Non-Cited Violation; consistent with Appendix F of the NRC Enforcement Policy. (NCV 50-272/99-06-02; 50-311-99-06-02) PSEG Nuclear has included this issue appropriately into the CAP (Notification 20004451 ).
. 2 Operations and Maintenance Inspection Scope The team reviewed a number of activities to verify that the AFW system was installed, opera_ted and maintained consistent with the design and licensing bases. The operational standby readiness and material condition of the AFW system was assessed by conducting system walkdowns and reviewing procedures, operator logs, design and vendor documents, component maintenance history records, and system health report The team also interviewed licensed and non-licensed operators and engineers. As part of this review, the team evaluated a sample of licensee-identified problems in the CAP as well as some emergent problems to assess the effectiveness of PSE&G Nuclear's corrective action Observations and Findings The team identified some instances where PSEG Nuclear did not effectively implement the CA *
A prior performance weakness resulted in a governor control oil system needle valve setting for the 23 AFW pump turbine that was not optimal for actual operating condition *
- During operation of the 23 TDAFWP, the team identified that the turbine outboard bearing temperature exhibited an increasing trend, which represented an operability question, and PSEG Nuclear was not aware of this potential proble Consequently, this problem was not promptly entered into the CA *
Other examples of ineffective implementation of the CAP were identified, including deficiencies not promptly entered into the CAP or not adequately correcte These issues were considered significant since they involved weaknesses in the CAP that negatively impacted the performance of risk significant equipment. Furthermore, the issues indicated an overall CAP weakness since problems were noted in not identifying corrective action items and the lack of timeliness and effectiveness of corrective action The issues were considered green in the significance determination process because they did not have an immediate impact on AFW system operability and were ultimately included in PSEG Nuclear's CAP. The issues collectively represented the third example of a violation of NRC requirements concerning inadequate corrective action. 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action", requires that measures shall be established to assure conditions adverse to quality are promptly identified and correcte This violation is being treated as a Non-Cited Violation, consistent with Appendix F of the NRC Enforcement Policy. (NCV 50-272/99-06-02; 50-311/99-06-02) PSEG Nuclear has included these issues appropriately into the CAP (Notification 20004451).
23 TDAFWP Governor Needle Valve Setting PSEG Nuclear initiated performance of the quarterly surveillance test of the 23 TDAFWP on July 23. After about 45 minutes of operation, operators noticed a speed oscillation of about 200 rpm peak to peak. Also, one of the U-bolt type supports on the two-inch recirculation line broke during the test. Operators declared the pump inoperable pending further evaluatio PSEG Nuclear attributed the cause for the speed oscillations to be related to the setting of the governor oil compensating needle valve, which affects the responsiveness of the governor. The needle valve was last adjusted during an overspeed test using compressed air rather than steam during the niost recent refueling outage (May 1999).
The turbine response is different when using compressed air as the turbine motive forc The use of compressed air resulted in a different position of the governor oil needle valve, causing the speed instability during the July 23 test. The needle valve is sensitive to oil viscosity changes as well as actual machine responsiveness. PSEG Nuclear subsequently evaluated the condition of the 23 TDAFWP and determined that the setting would not have resulted in a loss of governor function since the pump was able to deliver the required pressure and flow to perform its safety function. Although the team considered that this determination was reasonable, the issue was considered green in the significance determination process since the evaluation supporting the determination was largely dependent on engineering judgment (e.g., 200 rpm oscillation was acceptable while higher oscillations would not be). An inspection of the Unit 1 TDAFWP revealed its governor needle valve to be set correctly after its last overspeed test which J
was performed with compressed air. PSEG Nuclear had recognized that tuning the governor using compressed air in lieu of steam would result in a different needle valve adjustment; however, an activity was not scheduled and performed to implement the necessary adjustment. As additional corrective action, PSEG Nuclear issued Notifications 20002900 and 20002878 to add an activity to the governor actuator replacement recurring tasks (411001 and 461001) to require final tuning using steam as the motive forc *
Turbine Outboard Temperature Trend During operation of the 23 TDAFWP on August 4 to address oscillation problems previously experienced on July 23, the team requested a trend graph of its parameters to verify that the turbine governor had sufficient cooling. All parameters were normal with the exception of the turbine outboard bearing, which was about 165 °F and still increasing at the end of the one-hour run. The team questioned PSEG Nuclear regarding the ability of the 23 TDAFWP to meet its design function of operating for an extended time period considering there was a computer point high temperature alarm set at 180 °F for this bearing. PSEG Nuclear conducted an additional run of the pump on August 10 and documented a technical evaluation, which included vendor information indicating the bearing damage point to be 220 °F. This test demonstrated that a stable outboard bearing temperature was achieved at 164 °F. Although operability was not compromised, PSEG Nuclear did not initiate prompt actions to evaluate the August 4 temperature data that potentially degraded the design function of the 23 TDAFWP. The team considered that this lack of timeliness was significant, considering the challenge to operability of a risk significant system presented by the *issue. The issue was considered green in the significance determination process because it did not have an immediate impact on system operability and was ultimately resolved by inplant testin Other Examples of Ineffective Implementation of the Corrective Action Program The team identified two instances where PSEG Nuclear's entries of deficiencies into the corrective action program were untimely. The 23 TDAFWP experienced excessive speed oscillations during a routine surveillance on July 23; however, this discrepancy was not entered into the CAP until July 30. The team considered that this lack of timeliness was significant, considering the challenge to ope~ability of a risk significant system presented by the issue. The second instance involved the team's identification on July 20 that a large (about 3 foot diameter) manway cover was inappropriately stored adjacent to a r~turn line to the AFW storage tank, presenting a seismic interaction concern.. Although PSEG Nuclear implemented prompt actions to remove the cover, the*
item was not entered into the CAP until August 5. The team considered that this lack of timeliness was significant, considering the challenge to operability of a risk significant system presented by the issue and specific NRC prompting to enter the problem into the CAP. Both instances were considered green in the significance determination process because they did not have an immediate impact on AFW system operability and the problems were ultimately included in PSEG Nuclear's CA The team also identified two instances of ineffective corrective actions concerning past deficiencies. In the first instance, the team reviewed action requests (ARs) 990105086, 990222151, and 990323164 related to freezing and water intrusion concerns for the Unit 2 AFW storage tank overflow line loop seal and identified that PSEG Nuclear failed to implement the recommended corrective actions identified in the three ARs. Secondly, AR 980502123 recommended that TDAFWP IST performance procedure changes be implemented to prevent operational problems with the AFW pump turbine trip valve (MS52). However, the team identified that additional procedures that provided detailed
- operation of the valve were not similarly revised. PSEG Nuclear subsequently initiated additional items in the CAP to address these problems. The team considered that these two instances were significant since they indicated an overall CAP weakness which affects actions on multiple risk significant systems. Both instances were considered green in the significance determination process because they did not have an immediate impact on AFW system operability.
. 3 Surveillance and Testing Inspection Scope The inspectors reviewed test procedures and recent performance data to verify that the following AFW components met their design and licensing bases:
motor and turbine driven pumps
steam generator inlet control valves (AF21 s) and recirculation valves (AF40s)
....
autornatic actuation circuitry Observations and Findings The AF40 valves have a safety function to automatically close during emergency motor driven pump operation. This function limits pump recirculation flow to ensure that the required flow rates are delivered to the steam generators. Technical specification (TS)
surveillance 4. 7.1.2.c.1 requires verification that each automatic AFW valve not secured in position actuates to the correct position during an actual or simulated actuation signa The corresponding test procedure, S1 (2).0P-ST.AF-0009, "Plant Systems - Auxiliary Feedwater," does not verify that the AF40 valves close at the specified pump discharge flow rate. However, PSEG Nuclear verified that the valves operate properly during quarterly inservice testing using a simulated signal, in addition to a periodic calibration of the associated flow instruments that send the actuating signal to the valves during operation. The team concluded that this approach adequately satisfied TS surveillance 4.7.1.2. *
OTHER ACTIVITIES [OA]
40A2 Performance Indicator Verification Inspection Scope (IP 71151)
The team verified the accuracy and completeness of data used to calculate and report safety system unavailability for high pressure safety injection, auxiliary feedwater, emergency AC power and residual heat removal systems performance indicators (Pis)
for both Salem units. To assess the reported Pl data accuracy the team reviewed control room operating logs, condition reports, system health reports and interviewed the individuals compiling and trending the Pl dat Observations and Findings *
From a review of safety system unavailability data for the first and second quarters in 1999 and through discussions with cognizant personnei, the team determined that the Pl data was accurately reported for these safety systems using guidance contained in the Nuclear Energy Institute (NEI) draft Pl guideline document, 99-02, revision 8. None of the unavailability data reviewed for these systems exceeded the Pl thresholds for increased regulatory response. With respect to the AFW system, the Unit 1 TDAFWP was unavailable for about 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br /> in January 1999 due to governor performance problems as well as human performance problem.s during the related testing activitie No Pl threshold was exceeded as a result of this unavailabilit Pl unavailability performance data deficiencies identified by PSEG Nuclear of previously submitted data for 1997 and 1998 was properly entered into the CAP. PSEG Nuclear plans to submit the revised Pl unavailability data to the NR A4 Other (IP 71152) (Closed) I Fl 50-311197-16-03: Conformance of Calculations and Procedures with Technical Specification 4. 7.6.1.d (5). The inspectors determined that previous discrepancies identified within calculation S-C-CAV-MDC-1569, "Units 1 & 2 Control Room Envelope Cooling and Heating Load," and documented within inspection report 97-16, had no significant impact on the capability of the system to perform its design function. Surveillance procedures, S 1 (2). RA-ST. CAV-0001, "Control Room Emergency Ventilation System Surveillance Test," S1.RA-ST.CAV-0004(0), "Unit 1 Control Room Emergency Air Conditioning System (EACS) Surveillance," and S2.RA-ST.CAV-0003(0), "Unit 2 Control Room Emergency Air Conditioning System (EACS)
Surveillance," were reviewed and found to adequately ensure the capability of the Control Room EACS to remove the assumed heat load in accordance with TS 4. 7.6.1.d(5). No violations were identified with this issue.
11. (Closed) URI 98-09-04: weak resolution of carbon dioxide discharge valve seal failur The inspectors reviewed and discussed the issues and corrective *actions with PSEG Nuclear system engineering personnel. PSEG Nuclear's evaluation of the valve seal failure determined it to be an isolated failure. Subsequent inspections of similar valves did not reveal any seal degradation. The inspectors concluded that PSEG Nuclear's corrective actions were reasonable, and that no violation or deviation from regulatory requirements existe A5 Management Meetings PSEG Nuclear representatives were informed of the purpose and scope of the inspection at an entrance meeting conducted on July 19, 1999. The team presented the preliminary inspection findings to Messrs. D. Garchow and F. Sullivan and other members of PSEG Nuclear management on August 6, 1999, who acknowledged the findings presente Several subsequent telephone conversations were held with PSEG Nuclear personnel to further discuss the inspection findings. The last conversation occurred on September 1, 1999. No proprietary information was identified.
PARTIAL LIST OF PERSONS CONTACTED Public Service Enterprise Group T. Carrier D. Garchow A. Garcia J. Grant M. Kafantans S. Mannon G. Nagy T. Ross G. Salamon F.Sullivan E. Villar J.Zudans PSA Supervisor Vice President, Engineering AFW System Engineer Salem Maintenance Manager Salem Assistant Operations Manager Assistant Manager, Salem System Engineering *
Manager, Salem System Engineering Nuclear Fuels Supervisor Manager, Licensing Director, Nuclear Plant Engineering Licensing Engineer Manager, Mechanical Design New Jersey Department of Environmental Protection R. Pinney Nuclear Engineer ITEMS OPENED AND CLOSED
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Opened and Closed 50-27.2/99-06-01; 50-311/99-06-01 NCV Inadequate acceptance criteria in surveillance test procedures
50-272/99-06-02; 50-311/99-06-02 NCV Multiple examples of ineffective implementation of corrective action program 50-272/99-06-03; 50-311/99-06-03 NCV In adequate design control concerning replacement parts for positioners for AFW flow control valves Closed 50-311197-16-03 IFI Conformance of Calculations and Procedures with Technical Specification 4.7.6.1.d(5).
50-272/98-09-04; 50-311/99-09-04 URI Weak Resolution of Carbon Dioxide Discharge Valve Seal Failur *
AFW ASME CAP CFR gpm IST LLC MDAFWP NRC Pl PSEG psi a psig rpm TDAFWP TS UFSAR VP
LIST OF ACRONYMS USED Auxiliary Feedwater American Society of Mechanical Engineers Corrective Action Program Code of Federal Regulations gallons per minute lnservice Testing Limited Liability Corporation Motor Driven Auxiliary Feedwater Pump Nuclear regulatory Commission Performance Indicator Public Service Enterprise Group pounds per square inch absolute pounds per square inch gauge revolutions per minute Turbine Driven Auxiliary Feedwater Pump Technical Specification Updated Final Safety Analysis Report Vice President