IR 05000272/2010008
| ML100980293 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/08/2010 |
| From: | Doerflein L T Engineering Region 1 Branch 2 |
| To: | Joyce T P Public Service Enterprise Group |
| References | |
| IR-10-008 | |
| Download: ML100980293 (20) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PENNSYLVANIA 19406-1415 Mr. Thomas President and Chief Nuclear Officer PSEG Nuclear LLC -N09 P.O. Box 236 Hancock's Bridge, NJ 08038 April 8, 2010
SUBJECT: SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2-NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000272/2010008 and 05000311/2010008
Dear Mr. Joyce:
On February 25,2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Salem Nuclear Generating Station, Unit Nos. 1 and 2. The enclosed .inspection report documents the inspection results, which were discussed on February 25,2010, with Mr. C. Fricker and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. In conducting the inspection, the team reviewed selected procedures, calculations and records, observed activities, and interviewed station personnel.
Based on the results of this inspection, no findings of significance were identified . . In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice, II a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Docket Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading*rm/adams.html(the Public Electronic Reading Room). Docket Nos. 50*272; 50-311 License Nos. DPR-70; DPR*75
Sincerely,Lawrence T. Doerflein, Chi Engineering Branch 2 Division of Reactor Safety
Enclosure:
Inspection Report No. 05000272/2010008 and 05000311/2010008 wi
Attachment:
Supplemental Information cc w/encl: Distribution via ListServ ...
Mr. Thomas President and Chief Nuclear Officer PSEG Nuclear LLC -N09 P.O. Box 236 Hancock's Bridge, NJ 08038 April 8, 2010
SUBJECT: SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2-NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000272/2010008 and 05000311/2010008
Dear Mr. Joyce:
On February 25,2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Salem Nuclear Generating Station, Unit Nos. 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on February 25, 2010, with Mr. C. Fricker and other members of staff. , The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. In conducting the inspeCtion, the team reviewed selected procedures, calculations and records, observed activities, and interviewed station personnel.
- Based on the results of this inspection, no findings of significance were identified.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Docket Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). 'ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html(the Public Electronic Reading Room). Docket Nos. 50-272; 50-311 License Nos. DPR-70; DPR-75
Sincerely," IRA! Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety
Enclosure:
Inspection Report No. 05000272/2010008 and 05000311/2010008 wI
Attachment:
Supplemental Information cc wfencl: Distribution via ListServ Distribution w/encl: P. Wflson. DRS (R1DRSMall Resource)
S. Collins, RA (R10RAMAIL Resource)
A. Burritt. DRP M. Dapas. DRA (R10RAMAIL Resource)
L. Cline. DRP D. Lew. DRP (R1DRPMAll Resource)
A. Turilin. DRP J. Clifford.
DRP (R1DRPMAll,Resource)
R. Moore, DRP D. Roberts. DRS (R1DRSMall Resource)
D. Schroeder.
DRP. SRI H. BaHan, DR?, RI L. Trocine, RI OEDO RldsNrrPMSalem Resource ROPreportSResource@nrv.gov SUNSI Review Complete:
LTD (Reviewer'S Initials)
ML100980293 DOCUMENT NAME: G:\DRS\Engineering Branch 2\McKenna\Salem_ModsJeporC2010008.doc After declaring this document "An Official Agency Record" it will be released to the Public. To receive a copy of this document, Indicate in the box: "C" = Copy wtthout attachment/enclosure "En = Copy wtth tt ch tJ cl "N" N a a men en osure = 0 copy OFFICE RIIDRS I RItDRP I RIIDRS I NAME PMcKennal ABurritt/
LDoerfieinl DATE 3118/10 ,4/6/10 4/8/10 OFFICIAL RECORD COpy 1 i* u.s. NUCLEAR REGULATORY COMMISSION REGION I Docket Nos.: 50-272,50-311 License No.: DPR-70, DPR-75 Report No.: 05000272/2010008 and 05000311/2010008 Licensee:
PSEG Nuclear LLC (PSEG) . Facility:
Salem Nuclear Generating Station, Unit Nos. 1 and 2 Location:
P.O. Box.236 Hancocks Bridge, NJ 08038 Inspection Period: February 8 -25, 2010 Inspectors:
P. McKenna. Reactor Inspector, Division of Reactor Safety (DRS), Team Leader M. Balazik, Reactor Inspector, DRS L. Scholl. Senior Reactor Inspector.
DRS Approved By:. Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Enclosure
SUMMARY OF FINDINGS
IR 05000272/2010008 and 05000311/2010008; 02/08/2010
-02/25/2010;
Salem Nuclear Generating Station Unit Nos.1 and 2; Engineering Specialist Plant Modifications Inspection.
The report covers a two week on-site inspection of the evaluations of changes, tests, or experiments and permanent plant modifications.
The inspection was conducted by three region based engineering inspectors.
No findings of significance were identified.
The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4,dated December 2006.
A. NRC-Identified
and Self-Revealing Findings No firidings of significance were identified.
B. licensee-Identified Violations None. ii
REPORT DETAILS
REACTOR SAFETY
Cornerstones:
Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations
of Changes. Tests. or Experiments and Permanent Plant Modifications (IP 71111.17)
.1 Evaluations
of Changes. Tests. or Experiments (27 samples)
a. Inspection Scope
The team reviewed five safety evaluations to determine whether the changes to the facility or procedures, as described in the Updated Final Safety Analysis Report (UFSAR), had been reviewed and documented in accordance with 10 CFR 50.59 requirements.
In addition, the team evaluated whether PSEG had been required to obtain NRC approval prior to implementing the change. The team interviewed plant staff and reviewed supporting information including calculations, analyses, design change documentation, procedures, the UFSAR, the Technical Specifications (TSs), and plant drawings, to assess the adequacy of the safety evaluations.
The team compared the safety evaluations and supporting documents to the guidance and methods provided in Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Evaluations," as endorsed by NRC Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," to determine the adequacy of the safety evaluations.
The team also reviewed a sample of twenty two 10 CFR 50.59 screenings and applicability determinations for which PSEG had concluded that no safety evaluation was required.
These reviews were performed to assess whether PSEG's threshold for performing safety evaluations was consistent with 10 CFR 50.59. The sample included design changes, calclilations, procedure changes, and setpoint changes. The team reviewed the safety evaluations that PSEG had performed during the time period covered by this inspeCtion (i.e. since the last modifications inspection).
The screenings and applicability determinations were selected based on the safety significance, risk significance, and complexity of the change to the facility.
In addition, the team compared PSEG's administrative procedures used to control the screening, preparation, review, and approval of safety evaluations to the guidance in NEI 96-07 to determine whether those procedures adequately implemented the requirements of 10 CFR 50.59. The reviewed safety evaluations, screenings, and applicability determinations are listed in the attachment.
b. Findings
No findings of significance were 2
.2 Permanent
Plant Modifications (12 samples) ".2.1 Replacement of Service Water Valves 12SW33, 12SW35, 12SW37, 12SW46, and 12SW469, Unit 1 a. " Inspection Scope The team reviewed a modification (Design Change Package (DCP) 80096213)that replaced several 6 inch service water (SW) valves to include 12SW33, 12SW35, 12SW37, 12SW46, and 12SW469. The modification was implemented because the existing valves had a history of seat leakage. The valves were replaced with valves" made out of a more corrosion resistant material.
The team's review was performed to verify that the design bases, licensing bases, and performance capability of the service water system had not been degraded by the modification.
The team reviewed PSEG's installation work order, including the adequacy" of the post-modification results. The team interviewed engineering staff and' conducted a walk down of the installed valves to determine if the material condition and performance of the SW system was acceptable and in accordance with design "assumptions.
The team also reviewed stress calculations and conducted a walk down of the additional pipe supports installed for this modification to assess the installed configuration.
Additionally, the team reviewed the 10 CFR 50.59 screen associated with this modification as described in section 1 R1"7.1 of this report. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
,2,2 Emergency Diesel Generator Field Flashing Relay Replacement.
Unit 2
a. Inspection Scope
The team reviewed a modification (DCP 80095529)that replaced the K1 C field flashing relay and installed an additional relay to the field flash supervisory circuit of the Unit 2 "An emergency diesel generator (EDG). The field flash relay was replaced because the original style relay was no longer available and the second relay was installed to improve the electrical separation of the supervisory circuit from the" K1 C relay. The supervisory circuit provides indication to plant operators tl1at the field flash relay had not properly reset or that the operating coil had failed open, Either condition, if left uncorrected, would prevent proper field flashing during the next start of the EDG. The team assessed the modification to verify that the design bases and performance of the EDG had not been adversely impacted by the relay and circuitry "changes.
The team also discussed the impact of the modification on the EDG operation with responsible engineers and reviewed the status of these changes for the remaining EDGs. The engineers confirmed that all of the EDGs had received the same modification under other similar design change packages.
The team performed a field " Enclosure 3 inspection of accessible portions of the circuits to assess the quality of the modification work and the overall material condition of the equipment.
The adequacy of the modification testing was verified and affected design documents were reviewed to ensure they had been properly updated. The team also reviewed the 10 CFR 50.59 screening associated with this design change. The documents reviewed are listed in the attachment.
'b. Findings No findings of significance were identified . . 2.3 Replacement of Auxiliary Feedwater Storage Tank (AFST)
Check Valve (2DR71. Unit 2
a. Inspection Scope
The team reviewed a modification (DCP 2EE00337)that replaced the AFST isolation check valve (2DR7). Check valve 2DR7 serves as an isolation between the AFST and the demineralized water system, which is used to fill the AFST. The design function of the check valve is to prevent inadvertent draining of the AFST during a pipe break or loss-of-offsite' power, thereby ensuring AFST operability.
The modification was , implemented address obsolescence of check valve parts and the seat leakage history of the installed valve. In the modification installed an upstream vent valve , (2DR174) to allow for testing of check valve 2DR7. The review was performed to verify that the design and licensing bases and performance capability of the AFST had not been degraded by the modification.
The team assessed whether the component safety classification and specific safety functions were maintained.
The team reviewed various technical evaluations to assess wbether the modification was consistent with assumptions in the design and licensing bases related to the operation of the auxiliary feedwater system. Surveillance and post-modification, test results were reviewed to verify the check valve would function in accordance with the design and to verify that test results appropriately supported system operability.
The team performed a walkdown to assess the system material condition ,and the installed configuration of the check valve and vent valve. The team also reviewed affected pla!')t documents and drawings to verify they were Cippropriately updated. Finally, the team conducted interviews with engineering staff to determine if the valve would fUnction in accordance with technical and design assumptions.
The , documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.4 Replacement
of Inboard Containment Purge Supply and Exhaust 'Isolation Valves (2VC2 and 2VC3) with Blind Flanges. Unit 2
a. Inspection Scope
The team reviewed a modification (DCP 80091075)that replaced the inboard containment purge supply isolation valve (2VC2) and the inboard containment exhaust isolation valve (2VC3) with double testable, 36 inch, blind flanges. PSEG implemented this modification because the containment isolation valves in the containment purge system have had a history of requiring repair to pass leak rate tests and that spare parts were not readily available.
The containment purge system is a normally closed, deactivated system that is manually energized as required to perform purging of the containment atmosphere following a plant shutdown.
The blind flanges are removed during modes 5 and 6 to allow containment purge system operation.
The outboard supply and exhaust containment isolation valves (2VC1 and 2VC4) serve as containment closure valves if an isolation of containment is required during modes 5 or 6. The blind flanges are reinstalled and leak rate tested prior to changing to mode 4. The team reviewed the modification to verify that the design bases, licensing bases, and peiformance capability of the containment purge system and the containment had not been degraded by the modification.
The team also reviewed the 10 CFR 50.59 screen, as described in section 1 R17.1 of this report, and a previously NRC approved license amendment (No. 2(0) associated with this modification.
The team interviewed engineering staff and reviewed technical evaluations associated with the modification to determine if the blind flanges would function in accordance with the design assumptions.
The team also reviewed PSEG's installation work order including post-modification testing results to ensure appropriate acceptance criteria had been applied. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified . . 2.5 Motor Control Center Feeder Circuit Breaker Replacement.
Unit 2
a. Inspection Scope
The team reviewed a modification (OCP 80095528)that replaced the K225 frame circuit breaker with a K600 frame circuit breaker in Unit 2 vital bus 2B position . S2230-2BY1AX3Y.
The replacement was necessary to make the circuit breaker configuration consistent with the electrical bus configuration.
Specifically, the bus bars in the affected location where the circuit breaker line side connected were 0.5 inch thick and therefore designed for a 600 amp rated circuit breaker. However, the design drawing improperly specified a 225 amp breaker (designed to connect to a 0.25 inch thick bus bar), resulting in the mismatch between the breaker and bus configuration.
This mismatch could result in excessive stresses on the circuit breaker line side connections and/or the bus bars. Enclosure 5 The team reviewed the change to ensure that the design bases and performance capability of the vital bus were not affected by the change. This included a review of the replacement circuit breaker over current protection trip unit set points and time/current characteristics to ensure they remained consistent with design bases information (e.g. circuit breaker coordination calculations).
The inspectors discussed the change with the . responsible design and system engineers to evaluate the extent-of-condition and verify consistency between the design documentation and installed configuration on similar load centers. The team also verified affected procedures and design drawings had been . properly updated. A field walk down was performed to verify the circuit breaker configuration in the affected Unit 1 and 2 vital buses was consistent with the design documents.
The team also reviewed the 10 CFR 50.59 screening associated with this design change. The documents reviewed are listed in the attachment.
b. . Findings No findings of Significance were identified . . 2.5 Actuator Modification of Residual Heat Removal (RHRl System Suction Valves, Unit 2
a. Inspection Scope
The team reviewed a modification (DCP 80090480)that replaced the actuator gearing of the containment sump suction valves (21SJ44 and 22SJ44). The modification was implemented to address low thrust margin of the valves when subjected to a potentially higher differential pressure developed across the valves due to operation of the RHR pumps on minimum flow during certain small break loss-of-coolant accidents.
Also, a gearing replacement was performed on the refueling water storage tank (RWST) suction valves, 21 RH4 and 22RH4, to decrease the stroke time to ensure the RWST operability was not impacted.
In addition, the team reviewed the 10 CFR 50.59 screen associated with this modification as described in section 1R17.1 ofthis report. The review was performed to verify that the design and licensing bases and performance capability of the RHR system had not been degraded by the modification.
The team assessed whether the modification was consistent with assumptions in the design and licensing bases. Additionally, post-modification, diagnostic, and stroke-time testing data was reviewed to verify the operability and thrust margin of the valves. The review included verifying the UFSAR, calculations, test and operating procedures were updated to incorporate the modification.
The team verified that the operator training plan was also appropriately updated to incorporate the modification, Finally, the team conducted interviews with engineering staff to determine if the valves would function in accordance with technical and design assumptions.
The documents reviewed are listed in the attachment.
b, Findings No findings of significance were identified.
.2.7 Relocation
of Service Water Accumulator Injection Line Check Valves. Unit 2
a. Inspection Scope
The team reviewed a modification (DCP 80092250)that relocated the SW accumulator injection line check valves (21SW536 and 22SW536).
Two 15,000 gallon pressurized storage tanks are connected to the SW piping downstream of the SW pumps in order to keep the containment fan cooler units (CFCU) SW piping full of water following a offsite power (LOOP). The check valves in this piping prevent backflow from the SW headers to the accumulators during normal operations.
PSEG implemented this . modification to reduce the horizontal sections of SW pipe that were exposed to silt downstream of the check valves. The presence of silt adjacent to the valves discs could potentially impact valve opening. The team reviewed the modification to verify that the design bases, licensing bases, and . performance capability of the SW system had not been degraded by the modification.
The team reviewed the documentation supporting PSEG's evaluation and determination that it was acceptable to relocate the check valves next to the 90 degree elbows in the SW injection piping. The team reviewed calculation S-C-SW-MEE-191 0, Salem Units 1 & 2 CFCU Accumulator Injection Piping -Allowable Levels of Silt Accumulation during Plant Operation, to assess the impact of silt accumulation in CFCU accumulator injection piping during plant operation.
The team reviewed PSEG's installation work order, modification testing results, and revised pipe stress calculations for adequacy.
The team also interviewed engineering staff and conducted a walk down of the installed valves to determine if the material condition and of the SW system WClS acceptable and in accordancewith design Clssumptions
.. In addition, the team reviewed the 10 CFR 50.59 screen associated with this modification as described in section 1 R 17.1 otthis report. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified . . 2.8 Reactor Coolant System (RCS) Cold and Hot Leg Thermowell and RTD Replacement.
Unit 2 .
a. Inspection Scope
tearn reviewed a modification (DCP 80091019)that replaced the Unit 2 reactor coolant system hot and cold leg piping thermowells and associated narrow range resistance temperature detectors (RTDs). The thermowells are part of the reactor coolant system (RCS) pressure boundary and house the narrow range RTDs. The material of the original thermowells was Alloy 600 and the material of the replacement thermowells was 316 stainless steel. This change was implemented to eliminate the Alloy 600 material in order to reduce the plant's susceptibility to potential primary water stress corrosion cracking (PWSCC) issues. Enclosure 7 The team discussed the change with the responsible design engineers and evaluated the change to verify it did not adversely impact the design function of the thermowells and RTDs. The inspectors reviewed the results of the post*modification testing to verify the integrity of the primary system and the accuracy and operability of the RTDs. In addition, the team verified affected design documents and instrumentation calibration procedures had been updated to incorporate the modification.
The team also reviewed the 10 CFR 50.59 screening associated with this design change as described in section 1R17.1 of this report. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified . . 2.9 Modification of Pressurizer Spray Valves Internals (PS-1 & PS-3), Unit 2
a. Inspection Scope
The team reviewed a modification (DCP 80098324)that replaced the valve internals of the pressurizer spray valves (PS-1 and PS-3). The modification was implemented to address valve performance issues and obsolescence of vendor parts. Specifically, the modification eliminated the valve internal bellows, replaced the valve actuator, removed the valve bonnet extension, and upgraded the valve flow control characteristics.
The review was performed to verify that the design bases, licensing bases, and performance capability of the pressurizer system had not been degraded.
by the modification.
In addition, the team reviewed the 10 CFR 50.59 screen associated with this modification as described in .section 1 R 17.1 ofthis report. The team assessed whether the modification was consistent with assumptions in the design and licensing bases. Additionally, diagnostic and testing data was reviewed to verify the operability of the valves. The team reviewed selected plant procedures, calculations, training plans, and drawings to verify they were properly updated to incorporate the modification.
Finally, the team conducted interviews with engineering staff to determine if the valve would function in accordance with technical and design assumptions.
The documents reviewed are listed in the attachment.
b. . Findings No findings of significance were identified . . 2.10 MOV Margin Recovery Modifications for Valves 2RH2 & 2RH26, Unit 2 The team reviewed a modification (DCP 80093314)that enhanced motor operated valve (MOV) margin for valve 2RH26 (RHR to RCS hot leg recirculation isolation valve) and valve 2RH2 (RCS hot leg to RHR suction header valve). This modification changed the packing arrangement in both valves from two sets of six rings each to one set of five rings. Additionally, a leak off port in-between the two sets of packing was cut and capped for each valve. The modification to the valve packing reduced friction loads on . Enclosure
", '" the valve stem. Valve 2RH26 was also modified to replace the torque switch with a limit switch to control seating force. The team reviewed the, modification to verify that the design bases, licensing bases, and performance capability of the RHR system had not been degraded by the modification.
The review included verifying the UFSAR, calculations, and test and operating procedures were updated to incorporate the modification.
The team also reviewed PSEG's installation work order including post-modification, diagnostic, and stro,ke-time testing data to verify the operability and thrust margin of the valves. In addition, the team conducted interviews with 'engineering staff to determine if the valves would function in accordance with technical and design assumptions.
Finally, the team reviewed the 10 CFR 50.59 screen' associated with this modification as described in section 1R17.1 of this report. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified . . 2.11 Containment Atmosphere Hydrogen Monitor Replacement.
Unit 2
a. Inspection Scope
The team reviewed a modification (DCP 80091022)that replaced the Unit 2 containment atmosphere monitoring hydrogen analyzers.
This change was implemented because the original analyzers had become ob,solete ,and replacement parts were no loriger available.
The team reviewed the change to verify that the design bases and performance capability of the hydrogen monitors had not been adversely impacted by the change. The review included verification that ,the accuracy of ,the new monitors met the value specified in the UFSAR. The inspectors discussed the change with the responsible design engineer to assess any potential impacts on system operation and to ensure the design functions were not adversely affected.
The team also verified post-modification calibration and testing were adequate to ensure system operability.
The team also verified affected procedures and design documents had been appropriately updated to incorporate the modification.
A field walkdown of the new monitors was performed to verify the installed configuration was as described in the design change documentation.
The team also reviewed the 10 CFR 50.59 screening associated with this design change. The documents reviewed are listed in the attachment.
'
b. Findings
No findings of significance were identified.
.2.12 Modification
of Containment Spray Injection Valves {21CS2 & 22CS2}. Unit 2
a. Inspection Scope
The team reviewed a modification (DCP 80093311)that upgraded the capability of the containment spray (CS) injection valves (21 CS2, and 22CS2). The CS2 motor operated valves are normally closed valves that receive an automatic signal to open on high containment pressure, The modification was implemented to address low thrust margin of the valve actuator during design basis events: Specifically, the modification replaced the valve 10 foot-pound (ft-Ib) actuator motors with 15 ft-Ib actuator motors, as well as an actuator gearing replacement.
The new motors provided grea,ter thrust margin to ensure operation during conditions of high differential pressure across the valve. In addition, the team reviewed the 10 CFR 50.59 screen associated with this modification as described in section 1 R 17.1 of this report. The review was performed to verify that the design and licensing bases and performance capability of the containment spray system had not been degraded by the modification.
The team assessed whether the modification was consistent with assumptions in the design and licensing bases. Additionally, post-modification, diagnostic, and stroke-time testing data was reviewed to verify the operability and thrust margin of the valves. The team's review included verifying the UFSAR, calculations, and test and operating procedures were appropriated updated to incorporate the modification.
The team performed a walkdown to assess the material condition' and installed configuration of the valve actuators.
Finally, the team conducted interviews with engineering staff to determine if the valves would function in accordance with technical and design assumptions.
The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
40A2 Identification and Resolution of Problems (IP 71152) The team reviewed a sample of problems that PSEG had previously identified and entered into the corrective action program. The team reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions. In addition, condition reports written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment.
b. Findings
No findings of significance were identified.
10 40A6 Meetings.
Including Exit The team presented the inspection results to Mr. C. Fricker, Site Vice-President, and other members of PSEG's. staff at an exit meeting on February 25, 2010. The team returned the proprietary information reviewed during the inspection to the licensee and verified that this report does not contain proprietary information.
A-1
=SUPPLEMENTAL
INFORMATION=
KEY POINTS OF CONTACT
PSEG Personnel
- M. Ahmed, Design Engineering
- R. DeSanctis, Director Maintenance
- A. Garcia, System Engineering
- F. Hummel, System Engineering
- A. Johnson, Manager, Design Engineering .
- D. Johnson, Programs Engineering .
- W. Kittle, Programs Engineering
- K. Mathur, Design Engineering
- W. Mattingly, Manager, Regulatory
Assurance
- R. Moore, System Engineering
- N. Ortiz, Design Engineering
- R. Page, Design Engineering
J. Patel; Design Engineering
- M. Puher, Design Engineering
- L. Rajkowski, Engineering
Director
- T. Ram, System Engineering
- F. Szanxi, Programs Engineering
- E. Villar, Regulatory
Compliance
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
None.
LIST OF DOCUMENTS
REVIEWED 10
- CFR 50.59 Evaluations
- S2006-143, Replacement of Steam Generators, Rev. 0 . 82007-213, Calculation Methodology Change for Determining
- RHR Pump NPSH during Post LOCA Recirculation Operation, Rev. 0 . 82008-028, Implementation of the Equivalent Closure Device for TS 3/4.9.4 and Containment Closure for Loss of RHR Events During S2R16, Rev. 0 82008-105, Steam Generator Supports, Rev. 2 S2008-212, Steam Generator Carryover Test, Rev. 0 10
- CFR 50.59 Screened-out Evaluations
- 2007-149, Replace 2VC2 and 2VC3 Valves with Blind Flanges, Rev. 0 S2007-188, Salem Unit 2 8W Accumulator Injection Line Modification, Rev. 0 82007-258, Procedure Revision/1-EOP-TRIP-2, Rev. 0 82007-280, Install Temporary Jumper Across Enable Switches S02, Rev. 0 S2007-288, Replacement of RC8 Hot & Cold Leg PipingRTDsand Thermowells, Rev. 0 S2007-315, Change Stroke Times on SJ44 & RH4 Valves, Rev. 1 82007-350, Salem Unit 1 & 2 Pressurizer.Level, Rev. 0
- S2007-352, Salem Unit 1 & 2 Reactor Flux Power Range, Rev. 0 S2007-360, Replace Valve Actuator Motor MOVs 21 CS2, 22CS2, and Limit Seat MOV Valve 22CS36, Rev. 0 S2007-383, MOV Margin Recovery Modifications for Valves 2RH2 & 2RH26, Rev. 0 S2008-013, 12 Service Water Strainer Non-Conformance, Rev. 0 S2008-155, Unit 1 & 2 CVC Letdown Heat Exchanger Outlet Control Valve, Rev. 0 S2008-216.
- Replacement of Service Water Valves 12SW33, 12SW35, 12SW37, 12SW46, 12SW469, Rev. 0 S20 08-245 , SSPS Circuit Card Upgrade, Rev: 0 S2008-268, Salem Service Water Pump Mechanical Seal Modification, Rev. 0 S2008-315, Reactor Coolant System Water Inventory Balance, Rev. 0 S2009-041, Unit 1 NIS Power Range Negative Flux Rate Trip Elimination, Rev. 0 S2009-092,.1
- B Diesel Generator Auxiliaries Air and Turbo Boost Air Check Valve Test, Rev. 0 S2009-148, Modify Pressurizer Spray Valve Internals, Rev; 0 S2009-210, Hot Standby to Cold Shutdown, Rev. 0 52009-230, Temporary Power to 22 Containment Spray Motor, Rev. 0 S2009-261, Unit 1 Thermal Overload and Circuit Breaker Changes, Rev. 0 Calculations
- 267733, Service Water Return Piping, Rev. 5
- 267735, Service Water Return Piping, Rev. 4
- 267788, Service Water Piping, Rev. 4 56'71386, Service Water Piping, Rev. 12
- 5671387, Service Water Piping, Rev. 15
- 567570, Auxiliary Feedwater Piping Stress Calculation, Rev. 4
- 5675247, Service Water Piping -Accumulator Fill Line Pump 21, Rev. 2
- 5675248, Service Water Piping -Accumulator Fill Line Pump 22, Rev. 1 1A-DWA-0190, Pipe Support Calculation for
- DWA-190, Rev. 1 S-2-SJ-MDC-2125, Salem Unit 2 RWST to Containment Sump Switch over Drain Down, Rev. 0
- SC-CB-V001-01, Salem Unit 1&2 Containment Ventilation Hydrogen Analyzer, Rev. 0
- SC-RC-P001-01, Over Temperatlire
- DT Scaling Salem Unit 2, Rev. 6 S-C-DG-MDC-11
- 01, Comparison of Emergency Diesel Generator Turbo Boost Air System Demand. Versus Design, Rev. 1 S-C-RC-SEE-0545, Station Blackout Analysis in Support of Unit 2 Steam Generator Replacement, Rev. 0 S-C-SJ-MDC-2124, Pressure Increase on RHR Side of SJ44 Following Small Break LOCA, Rev. 0 S-C-SJ-MDC-i127, Design Basis Differential Pressure (21SJ44 & 22SJ44), Rev. 0
- ES-4.006(Q), 125 Volt DC Component Study and Voltage Drop Calculation, Rev. 6 MIDACALC Results, 21CS2, Rev. 2 MIDACALC Results, 22CS2, Rev. 1 MIDACALC Results, 2RH2, Rev. 1 MIDACALC Results, 21 RH4, Rev. 1 MIDACALC Results, 22RH4, Rev. 1 MIDACALC Results, 2RH26, Rev. 1 MIDACALC Results, 21 SJ44, Rev. 1 MIDACALC Results, 22SJ44, Rev. 1 Attachment Completed Work and Test Procedures
- S2.0P-LR.MP-0001(Q), 2VC2 & 2VC3 Blind Flange Type B Mechanical Penetration Leak Rate Testing, performed
- 5/03/08 S2.0P-ST.CS-0003(Q), Inservice Testing Containment Valves, performed
- 11/17/09 S2.0P-ST.PZR-0003(Q), Inservice Testing Pressurizer
- PORV and Spray and Reactor Head Vent Valves, performed
- 10/20109 S2.0P-ST.RHR-0003, Inservice Testing Residual Heat Removal Flow Path Valves, performed
- 11/12/09 S2.0P-ST.RHR-0004(Q), Inservice Testing Residual Heat Removal Valves, performed
- 11/02/09 S2.0P-ST.SJ-0005(Q), Inservice Testing Safety Injection Valves Modes 5-6, performed
- 10/29/09 S2.0P-ST.ZZ-0003(Q), Inservice Testing Miscellaneous Valves, performed
- 06/19/08, 12/17/09, & 01/07110 .
- SH.MD-GP.ZZ-0003(Q}, General Instructions for Valve Packing, performed
- 3/24/08 Design & Licensing Bases LCR S06-06, Request for Change to Technical Specifications, Containment Ventilation System, Containment Isolation Valves, Salem Nuclear Generating Station -Unit 1 and 2, Docket Nos. 50-272 and 50-311, Facility Operating License Nos.
- DPR-70 and
- DPR-75, dated 08/04/06
- LR-N07-0032, Response to RAls on LCR S06-06 Request for Change to Technical Specifications Containment Ventilation System Containment Isolation Valves, dated 02/20/07
- LA 260 and 277, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendments Re: Containment Purge System, dated 03/19/07
- SC.DE-BD.CS-0001(Q}
- Design Bases Documentation for Containment Spray System, Rev. 0
- SC.DE-BD.RC-0001(Q)
- Design Bases Documentation for Reactor Coolant System, Rev. 0
- SC.DE-BD.AF-0001 (Q) Design Bases Documentation for Auxiliary Feedwater System, Rev. 0 Drawings 205246-A-8761, Units 1 & 2 Demineralized Water PI&D, Sh. 2,
- 205334-A-8763, Unit 2 Safety Injection
- PI&D, Sh.2, Rev. 51 205334-A-8763, Unit 2 Safety Injection
- PI&D, Sh.3,Rev.
- 205332-A-8763, Unit 2 Residual Heat Removal PI&D, Sh. 2, Rev. 33 205332-A-8763, Unit 2 Residual Heat Removal PI&D, Sh. 1, Rev. 36 205301-A-8762, Unit 2 Reactor Coolant P&ID, Sh.1, Rev. 58
- 205342, No.2 Unit Service Water Nuclear Area, Rev. 7 222485-A-1779, Unit 2 Auxiliary Building 2C West Valves One-Line Electrical Drawing, Sh. 1, Rev. 50
- 601401, Auxiliary Building Control Area 2B-230V Vital Bus One-Line, Rev. 22
- 904093, Schematic Diagram Insitu Micro, Rev. A
- C-401522, Trim Assembly, Sh. 1, Rev. 0 0-401194, Series 0-100 Valve Assembly with Model 1000-160.
- RA Actuator, Sh. 1, Rev. 4 0-401193, Series 1000-160 RA Actuator with Travel Stops, Sh. 1, Rev. 1 0-5587-326-001, Hot Leg Thermowell, Rev. 4 0-5587-326-002, Cold Leg Thermowell, Rev. 4 E-5587-326-004, Cold Leg Installation, Rev. 4 E-5587 .. 326-005, Hot Leg Installation, Rev. 5 Attachment Engineering Evaluations
- 3SC-025, Technical Evaluation of Replacement
- RCS Hot & Cold Leg Thermowells, Rev. 0 S-C-ZZ-CEE-0673
- Justification for Exempting
- VC1, VC2, VC3, and VC4 Limit Switches from _ Environmental Qualification, dated 04/15/92 S-C-SW-MDC-1705, Service-Water System Transient Analysis Model, Rev. 1 S-C-SW-MDC-1967, Service Water System Thermal Hydraulic Model, Rev. 4 S-C-SW-MEE-1138, Salem Generic Letter 96-06 Evaluation Salem Units 1 & 2, Rev. 2 S-C-SW-MEE-1910
- Salem Units 1 & 2 CFCU Accumulator Injection Piping -Allowable Levels -of Silt Accumulation during Plant Operation, Rev. 1 --,Miscellaneous
- 09-026, 2DR7 Operability Evaluation, dated 12/23/09
- 70105383, Permanent Plant Modifications Check-In Self-Assessment Report, dated 01/18/10 Flowserve Technical Update 08-01, Reliance/Magnesium Rotors, dated 12/19/08 Flowserve Technical Update 06-01, Reliance/Magnesium Rotors, dated 12/26/06 _ MPR Associates, Inc.
- Memorandum on Salem CFCU Simplification Project -Modification Scope for Tank Elimination, dated 12/29/06 . MPR Associates, Inc.
- Memorandum on Alternatives to Prevent Silting in Service Water Accumulator Piping, dated 07/22/07
- NEI 96-07, Guidelines for 10
- CFR 50.59 Implementation, Rev. 1 NRC Regulatory Guide 1.187, Guidance for Implementation of 10
- CFR 50.59, Changes. Tests, and Experiments, dated November 2000 NRC Regulatory Guide 1.186, Guidance and Examples for Identifying
- CFR 50.2 Design Bases, dated December 2000 Salem Inservice Testing Program Basis Data Sheets for 2RH2 & 2RH26. dated 12/01/97 Salem Licensed Operator Training Requalification Segment 0910-5'& 0708-4 ' Modifications
- DCP 2EE00337, Replacement of Demineralized Water Check Valve 2DR7, Rev. 1
- DCP 80090480, Change*Stroke times on SJ44 & RH4 Valves, Rev. 1
- DCP 80092250, Salem Unit 2 SW Accumulator Injection Line Modification, Rev. 2
- DCP 80093311, Replace Valve Actuator Motor
- DCP 80095529, Emergency Diesel Generator Field Flashing Relay Replacement, Rev. 1
- DCP 80095528, Unit 2 Motor Control Center Feeder Circuit Breaker Replacement, Rev. 0 bC?
- 80096213, Replacement of Service Water Valves 12SW33, 12SW35, 12SW37.12SW46, 12SW469, Rev. 0 'DCP
- 80098324, Modify Pressurizer Spray Valve Internals.
- Rev. 2
- DCP 80091019, Reactor Coolant System (RCS) Cold and Hot Leg Thermowell and RTD Replacement, Rev. 1 -DCP
- 80091022, Unit 2 Containment Atmosphere Hydrogen Monitor Replacement, Rev. 0 Attachment
Notifications
- 237971
- 20438556
- 20450729*
- 20451987*
- 236105
- 20444369
- 20450730*
- 20452105*
- 20362020
- 20444685
- 20450731*
- 20452175*
- Notification written as a result of inspection effort Procedures
- 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation, Rev. 28 2-EOP-LOCA-1, Loss of Reactor Coolant, Rev. 28 2-EOP-Trip-1;
- Reactor Trip or Safety Injection, Rev. 27
- 20366041
- 20450728*
- 20451971*
- 20452273*
- S2.IC-CC.CBV-0002{Q), 22 (2XA3359)
- Containment Hydrogen Analyzer, Rev. 8 S2.0P-SO.DG-0001(Q}, 2A Diesel Generator Operation, Rev. 36 S2.0P-SO.SW-0006, Service Water Accumulator Operation, Rev. 9 S2.0P-ST.SW-0015, Inservice Testing Service Water System CFCU and Accumulator Check Valves, Rev. 7 S2.PI-SP.SG-0004(Q), Steam Generator Moisture Carryover Test, Rev. 0
- CC-M-103-1005, Evaluating and Mitigating Electrically Induced Noise In Instrumentation and Control Circuits, Rev. 1 SC.lC-PM.RC-002(Q), Pressurized Spray Valve Operator Maintenance, Rev. 8 .
- SC.MD-CM.230-0001Q), 230 and 460 Volt ITE K-Series Breaker Maintenance, Rev. 5
- SC.MD-CM.RC-0002(Q), Pressurizer Spray Valve PS1 and PS3 Repair, Rev. 6
- SC.MD-PM.230-0003(Q), 230 and 460 Volt ABB K-Line Circuit Breaker Preventive Maintenance, Rev. 5
- SC.MD-PM.ZZ-0123, Disassembly, Inspection and Reassembly of Dual Plate Check Valves, Rev. 13
- SC.MD-ST.230-0003(Q), 230 and 460 Volt ABB K-Line Circuit Breaker Refurbishment, Rev. 24
- SH.MD-GR.ZZ-0240(Q), System Pressure Test at Normal Operating Pressure and Temperature, . . Rev. 10
- SH.MD-IT.ZZ-3907(Q), General Mechanical Equipment Inspection and Functional Testing, Rev. 1 Vendor Technical
Documents
- VTD 303254, Seismic Report for 6-inch 150 Dual Plate Check Valve, Rev. 5
- VTD 316968, Engineering.
- DeSign/Seismic Report 10" Dual Plate Check Valve, Rev. 4
- VTD 322681, Design/Seismic Report for 1" Fullport Flanged End Ball Valve, Rev. 2
- VTD 326317,36
inch Testable Blind Flange,.Rev.
- VTD 901603, Steam Generator Moisture Carryover Evaluation Test Results, dated
- 01/08/09 Work Orders .
- 30087129
- 30135254
- 60068308
- 60075842
- 60083557
- 70062703
- 70088801
- 70104163
- 30119589
- 30162529
- 60069077
- 60075941
- 60083558
- 70062810 .70092440
- 70105463
- 30135248
- 60007404
- 60071106
- 60078148
- 70048711 .70082970
- 70092441
- 80098851
- 30135255
- 60059037
- 60071107
- 60078149
- 70056975
- 70088222
- 70100505 Attachment
LIST OF ACRONYMS
Agency-Wide
Documents
Access and Management
System Auxiliary
Storage Tank Containment
Fan Cooling Unit Code of Federal Regulations
Containment
Spray Chemical and Volume Control Design Change Package Division of Reactor Safety Emergency
Diesel Generator
Emergency
Operating
Procedure
Loss-of-Coolant
Accident Loss-of-Offsite
Power Motor Operated Valve Nuclear Energy Institute
Nuclear Regulatory
Commission
Piping and Instrument
Diagram Pressurizer
Spray
LLC Primary Water Stress Corrosion
Cracking Request for Additional
Information
Reactor Coolant System Residual Heat Removal Resistance
Temperature
Detector Refueling
Water Storage Tank Service Water Technical
SpeCifications
Updated Final Safety Analysis Report Vapor Containment
.