IR 05000354/2011007

IR 05000354-11-007; PSEG Nuclear, LLC (Pseg); 04/18/2011 - 05/06/2011; Hope Creek Generating Station, Engineering Specialist Plant Modifications Inspection

ML111710487
Person / Time
Site:	Hope Creek, 05000345
Issue date:	06/20/2011
From:	Doerflein L Engineering Region 1 Branch 2
To:	Joyce T Public Service Electric & Gas Co
References
IR-11-007
Download: ML111710487 (25)

Text

UNITED STATES une 20, 2011

SUBJECT:

HOPE CREEK GENERATING STATION - NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT NO. 050003541201 1007

Dear Mr. Joyce:

On May 6,2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Hope Creek Generating Station. The enclosed inspection report documents the inspection iesults, which were discussed on May 6, 2011, with Mr. John F. Perry, Site Vice President, 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 were identified'

ln 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 Document Room or from the Publicly Available Records (PARS) component of the NRC's document system, Agencywide Documents Access and Management System (ADA[4S)

ADAMS is accessible from the NRC Web site at http://www.nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

--t { ,,

D/...r^^-r-.^l 1' l*

Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Docket No.: 50-354 License No.: NPF-57

Mr. Thomas President and Chief Nuclear Officer PSEG Nuclear LLC-N09 P.O. Box 236 Hancocks Bridge, NJ 08038 SUBJECT: HOPE CREEK GENERATING STATION - NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM I NSPECTION REPORT 05000354/2U 1 047

Dear Mr. Joyce:

On May 6,2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Hope Creek Generating Station. The enclosed inspection report documents the inspection results, which were discussed on May 6, 2011, with Mr. John F. Perry, Site Vice President, 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.

ln 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 were identified'

In accordance with 1O 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 Document Room or from the Publicly Available Records (PARS) component of the NRC's document system, Agencywide Documents Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RN Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Docket No.: 50-354 License No.: NPF-57 SUNSI Review Complete: LTD (Reviewer's Initials) ADAMS ACC #ML111710487 DOCUMENT NAME: GlDivision\DRS\Engineering Branch 2\Balazik\HCmodsreport2--Q1^1007-.doc Aflordodarimthisdocum6nl.AnomcialAqenryRo@rd,itwillb9r6i6sed-tolhePublic.ToEc.iw"coPy@

RECORD COPY

Enclosure:

lnspection Report No. 050003541201 1007 MAttachment: Supplemental lnformation

REGION I Docket No.: 50-354 License No.: NPF-57 Report No.: 05000354/2011007 Licensee: PSEG Nuclear, LLC (PSEG)

Facility: Hope Creek Generating Station Location: Hancocks Bridge, NJ lnspection Period: April 18 through May 6, 2011 lnspectors: M. Balazik, Reactor lnspector, Division of Reactor Safety (DRS), Team Leader K. Young, Senior Reactor lnspector, DRS J. Schoppy, Senior Reactor Inspector, DRS Approved By: Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

tR 0500035 412011007; 0411 812011 - 051061201 1; Hope Creek Generating Station; Engineering

Specialist Plant Modifications Inspection.

This report covers a two week on-site inspection period of the evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three region based engineering inspectors. 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.

No findings were identified.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: lnitiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Chanoes. Tests. or Experiments and Permanent Plant Modifications

(tP 71111.17)

Evaluations of Chanoes. Tests. or Experiments (26 samples)

a.

lnspection Scope The team reviewed three 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 changes. The team interviewed plant staff and reviewed supporting information including calculations, analyses, design change documentation, procedures, the UFSAR, the Technical Specifications (TS), 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 (NEl) 96-07, "Guidelines for 10 CFR 50.59 Evaluations," as endorsed by NRC Regulatory Guide 1.187, "Guidance for lmplementation 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-three 10 CFR 50.59 screenings 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, calculations, and procedure changes.

The team reviewed the safety evaluations that PSEG had performed and approved during the time period covered by this inspection (i.e., since the last modifications inspection) not previously reviewed by NRC inspectors. 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 and screenings are listed in the attachment.

b. Findinqs No findings were identified.

.2.1 Service Water Strai ner Structural mprovements

I lnspection Scope The team reviewed a modification (80097547) that implemented structural improvements to the safety-related service water (SW) strainers. The SW strainers (H1EA -14, B, C, D

-F-S09) had a history of strainer element breakthrough events (failure of the mesh screening when in service during heavy loading). These failures occurred when heavy grass accumulated on the strainer, and increased the differential pressure (DP) across the screen. PSEG modified the construction of the SW strainers to improve the ability of the screen to withstand these types of grassing events. Specifically, the structural improvements included the installation of an additional mid ring and additional tack welding on the mesh elements.

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 interviewed engineering staff and reviewed technical evaluations associated with the modification to determine if the SW strainers and SW system would function in accordance with the design assumptions. Although PSEG implemented the modification on all four SW strainers, the team focused their review on the 'A' SW strainer. The team performed several walkdowns of the SW pump bays and control room instrumentation to independently assess PSEG's configuration control, the material condition of the SW strainers, and the relative grass loading on the SW traveling water screens. The team reviewed the associated post modification test (PMT) results and SW strainer DP trend data to verify that the strainers functioned as designed following the modification. In addition, the team reviewed several post-modification periodic preventive maintenance (PM) inspections of the SW strainer internals to assess the effectiveness of PSEG's modification and the condition of the strainers. The team also reviewed corrective action notifications (NOTF) to determine if there were reliability or performance issues that may have resulted from the modification. The documents reviewed are listed in the attachment.

b. Findinqs No findings were identified.

.2.2 Reactor Buildinq Floor Pluq Hatch Hold Down Anchors Reconfiouration

a. Inspection Scope

The team reviewed a modification (80097553) that revised the anchoring configuration for the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC)floor plug hatches in the reactor building (above the HPCI and RCIC pump rooms).

PSEG implemented the modification in response to an updated high energy line break analysis to ensure that the floor plugs continued to function as designed during design basis accident conditions. ln addition, PSEG upgraded and revised the anchoring configuration to address an adverse condition (corroded bolting) associated with the pre-existing hold down anchors.

The team reviewed the modification to verify that the design bases, licensing bases and structural integrity of the HPCI/RCIC floor plugs had not been degraded by the modification. The team reviewed calculations and technical evaluations to verify that the hold down anchors and floor plugs would function in accordance with design assumptions. The team reviewed the associated work order instructions and documentation to verify that maintenance implemented the modification as designed.

The team also walked down the floor plugs and adjacent pump rooms to verify that PSEG had adequately implemented the modification, maintained configuration control, and had not impacted the operation of other safety-related structures, systems, and components (SSC) located in the vicinity. The team conducted a plant walkdown of the modification with the respective design engineer to discuss the critical design aspects and to independently verify that maintenance installed the anchors in accordance with engineering's design assumptions and associated PSEG procedural requirements.

Additionally, the team reviewed the 10 CFR 50.59 screen and engineering evaluation associated with this modification. The documents reviewed are listed in the attachment.

b. Findinqs No findings were identified.

.2.3 Hiqh Pressure Coolant Iniection Main Pump and Booster Pump Tie Down Bolt

Replacement

a. Inspection Scope

The team reviewed a modification (80097504) that replaced the HPCI main and booster pump tie down bolts. The HPCI main and booster pumps were designed to mount using dowels and ASTM (American Society for Testing and Materials) 4307 mounting bolts.

The dowel pins were designed to bear the horizontal anchor shear loads while the four 4307 bolts, per pump, would resist the vertical lifting tension loads. However, during an extent-of-condition review (NUCR 70074025), PSEG had previously identified that the dowel pins, shown on vendor drawings and included in the seismic analysis for resisting horizontal seismic forces, were not installed in the field. PSEG's operability evaluation (DEH08-0135) for this condition had determined that the dowel pins were not required providing the correct material was used for the mounting bolts. The original 4307 bolts were bearing the full horizontal shear loads in addition to vertical lifting tension loads.

However, ASTM 4307 material is a low carbon steel bolt and, therefore, subject to a potential torque relaxation over time which is detrimental to its shear carrying ability. ln order to address this potential longterm torque relaxation issue, PSEG upgraded the 4307 bolts to A325 bolts torqued for a friction type connection per American Institute of Steel Construction (AISC) requirements. The new HPCI pump anchoring system using the ASTM 4325 bolts performs the same design function as the ASTM 4307 bolts and dowel pins, and resolved the corrective actions identified in the associated operability evaluation.

The team reviewed the modification to verify that the design bases, licensing bases and performance capability of the HPCI system had not been degraded by the tie down bolt replacement. The team reviewed calculations and technical evaluations to verify that the tie down bolts would function in accordance with design assumptions. The team reviewed the associated work order instructions and documentation to verify that maintenance implemented the modification as designed. The team reviewed the associated PMT results, HPCI pump vibration and DP trend data, and post-modification HPCI surveillance test results to verify proper operation of the HPCI system. The team walked down the HPCI main and booster pumps to verify that PSEG had adequately implemented the modification, maintained configuration control, and had not impacted the operation of other safety-related SSCs located in the vicinity. The team also reviewed corrective action NOTFs to determine if there were reliability or performance issues that may have resulted from the modification. Additionally, the team reviewed the 10 CFR 50.59 screen and engineering evaluation associated with this modification. The documents reviewed are listed in the attachment.

b. Findinqs No findings were identified.

.2.4 Emerqencv Diesel Generator Fuel lniector Coolino Water Svstem Removal

a. Inspection Scope

The team reviewed a modification (80083515) that removed the fuel injector cooling water system from the'B'emergency diesel generator (EDG). The fuel injector cooling system was designed to minimize the effect of fuel varnishing on the fuel injector nozzle tips. Fuel varnishing is a particular concern for diesel engines that use a heavy fuel that must be heated to remain a liquid. The EDG vendor determined that fuel injector cooling was of marginal value for Hope Creek's diesel engines because they use a light diesel fuel. The vendor determined that the fuel used to run the EDGs does not need to be heated and the flow of fuel alone provided adequate cooling to minimize the effects of fuel varnishing. PSEG identified that this marginal valued fuel injector cooling system was the source of minor coolant leakage on all four EDGs and implemented the modification on each of the EDGs to improve EDG reliability, reduce cooling water leakage, and improve EDG availability.

The team reviewed the modification to verify that the design bases, licensing bases and performance capability of the EDG had not been degraded by the modification. The team interviewed engineering staff and reviewed technical evaluations associated with the modification to determine if the EDG and its support systems would function in accordance with the design assumptions. The team reviewed the associated work order instructions and documentation to verify that maintenance implemented the modification as designed. The team performed several walkdowns of the four EDGs to independently assess PSEG's configuration control, the material condition of the EDGs, and the integrity of the cooling water systems. In particular, the team directly observed portions of the 'A' and 'B' EDG monthly surveillances on May 2, 2011 , and April 19, 2011, respectively, to assess the cooling water system integrity and performance during

=nclosure EDG operation at rated conditions. The team also took advantage of an opportunity to directly observe the as-installed condition of 'B' EDG cylinder No. 2 fuel injector cooling plugs and lock wire, a normally inaccessible area, during an operations inspection activity that temporarily removed the cylinder head cover on May 5,2011. The team reviewed the associated PMT results and several subsequent EDG surveillance test results, including a 24-hour run, to verify that the 'B' EDG and coolant systems functioned as designed following the modification. The team also reviewed corrective action NOTFs to determine if there were reliability or performance issues that may have resulted from the modification. Additionally, the team reviewed the 10 CFR 50.59 screen and engineering evaluation associated with this modification. The documents reviewed are listed in the attachment.

b. Findinqs No findings were identified.

.2.5 lnstallation of New Vibration Sensors to Monitor Hope Creek Control Room Supplv and

Control Equipment Room Supplv Heatinq Ventilation and Air Conditionino Units

a. Inspection Scope

The team reviewed a modification (80072712)that mounted vibration sensors on motor and fan bearings, and installed cable from the sensors through the outside wall of the air handler to a connection box mounted outside the control area heating, ventilation, and air conditioning (HVAC) equipment (Control Room Supply (CRS) units HlGK-1A-VH403 and HlGK-18-VH403, and Control Equipment Room Supply (CERS) units HlGK-1A-VH4O7 and Hl GK-18-VH407). The modification was installed to enhance the ability of the operations department and maintenance technicians to periodically collect vibration data for the control area HVAC equipment without removing it from service. The modification provided an external test connection for easy interface of measuring and test equipment during vibration data gathering activities for the CRS and CERS systems.

The team conducted the review to ensure that the design bases, licensing bases, and performance capability of the CRS and CERS HVAC systems had not been adversely affected by the modification. The team reviewed PSEG's installation work orders, which included review of the adequacy of the post-modification testing results. The team interviewed the engineering staff regarding the design, installation, and testing of the new vibration sensors and associated wire harnesses to assess the adequacy of the modification. The team walked down the accessible portions of the new equipment to determine material condition of the system and ensure the vibration sensors and wire harnesses were installed in accordance with design assumptions and instructions. The team also confirmed that surveillance tests, operational procedures, and drawings had been appropriately updated to reflect the design change. The team reviewed corrective action notifications, control room HVAC system health reports, and completed surveillance procedures to determine if reliability or performance issues resulted from the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findinqs No findings were identified.

.2.6 Replacement of Control Room Chiller Bearino Oil and Refrioerant Discharqe

Temperature Switches a. lnspection Scope The team reviewed a modification (80093784) that upgraded temperature switches and the associated temperature sensor on the 'A' control room chiller (H1GJ-1A-K-400). The new temperature switches (H1GJ-IGJTISH-9652A0 and HlGJ-IGJTlSH-9552A2) and sensor (H1GJ-1GJTE-9652A13) are associated with measuring and indicating thrust bearing oil temperatures and refrigerant discharge temperature for the control room chiller. The upgraded components provided the same function as the previous components. The modification was implemented because the original components were unreliable and caused operational challenges with the control room chiller system.

The team conducted the review to ensure that the design bases, licensing bases, and performance capability of the 'A' control room chiller system had not been adversely affected by the modification. The team reviewed PSEG's installation work orders, which included review of the adequacy of the post-modification testing results. The team interviewed the engineering staff regarding the design, installation, calibration, and testing of the new components to assess the adequacy of the modification. The team walked down the accessible portions of the installed equipment to determine material condition of the system and ensure the temperature switches and sensors were installed in accordance with design assumptions and instructions. The team also confirmed that surveillance tests, operational procedures, and drawings had been appropriately updated to reflect the design change. The team reviewed corrective action notifications, control room chiller water system health reports, and completed surveillance procedures to determine if reliability or performance issues resulted from the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findinqs No findings were identified.

.2.7 Replacement of the 'B' Emerqencv Diesel Generator Fuel Oil Dav Tank Level Alarm

Switch and Raise the Alarm Setpoint a. lnspection Scooe The team reviewed a modification (80080423)that replaced the fuel oil day tank (FODT)level alarm switch (l JELSHL-7501B) and raised the high level alarm setpoint from 9.25 inches to 8.75 inches (measured from the top of the tank) for the 'B' EDG FODT. The FODT level alarm switch (H1JE-lJELSHL-7501B) provides high and low level alarms individually at the EDG remote control panel and provides a common trouble alarm in the main control room. The switch also provides a back-up starUstop function for the fuel oil transfer pumps in case the normal switch fails to operate the pumps. The modification was installed to correct numerous nuisance alarms and frequent switch failures during testing of the switch for the FODT. The new component and setpoint change was needed to enhance system reliability and eliminate nuisance alarms.

The team conducted the review to ensure that the design bases, licensing bases, and the performance capability of the EDG FODT system had not been adversely affected by the modification. The team reviewed PSEG's installation work orders, which included review of the adequacy of the post-modification testing results. The team interviewed the engineering staff regarding the design, installation, and testing of the new FODT level switch and revised FODT level setpoint to assess the adequacy of the modification.

The team walked down the accessible portions of the new equipment to determine material condition of the system and ensure the FODT level switch was installed in accordance with design assumptions and instructions. The team also confirmed that surveillance tests, operational procedures, calibration procedures, and drawings had been appropriately updated to reflect the design change. The team reviewed corrective action notifications, the EDG system health reports, and completed surveillance procedures to determine if reliability or performance issues resulted from the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findinqs No findings were identified.

.2.8 Revision to Calculation E-1.4. Hope Creek Class 1E 125 and 250 Vdc Svstems: Short

Circuit and Voltaoe Drop Studies. to Address Batterv lnter-cell Resistance

a. Inspection Scope

The team reviewed a change (S0095526) to calculation E-1.4 which added a basis for the inter-cell resistance values used in surveillance procedures HC.MD-ST.PJ-0002, "250 Volt Quarterly Surveillance," and HC.MD-ST.PK-0002, "125 Volt Quarterly Surveillance." The procedures provide administrative limits of battery inter-cell resistance of less than 50 micro-ohms. This value is conservative when compared to the Hope Creek's TS requirement of 150 micro-ohms. The administrative limits in the surveillance procedures allow PSEG to track, trend, and correct the battery inter-cell resistance values prior to them reaching the TS limit and becoming an operability concern. The calculation change did not impact the inter-cell resistance requirements of the TS. PSEG performed this change based on review of an issue identified at another nuclear power plant.

The team evaluated the calculation revision to confirm that the 125 and 250 Vdc batteries' design bases, licensing bases, and performance capability would not be affected by the change. The team reviewed the calculation and associated analysis to verify the assumptions used in the calculation were valid. The team reviewed associated calculations to ensure they had been updated based on the limits analyzed.

The team reviewed procedures to ensure battery inter-cell resistance limits were valid, had been implemented, and were conservative when compared to the applicable TS sections. The team interviewed the engineering staff to review the calculation assumptions, applicable surveillance procedures, and calculation methodology to verify their adequacy. The team walked down the battery systems to determine material condition of the batteries. The team reviewed corrective action notifications, the 125 and 250 Vdc batteries system health reports, and completed surveillance procedures to determine if reliability or performance issues existed. The 10 CFR 50.59 screening determination associated with this revised calculation was also reviewed as described in section 1R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findinqs No findings were identified.

.2.9 Service Water Strainer Allowable Differential

Pressure Analvsis

a. Inspection Scope

The team reviewed a new analysis (80101059) that ensured the SW strainers (HIEA -

1A, B, C, D -F-S09) design function was maintained at elevated differential pressures.

PSEG replaced the existing strainers with a more robust design due to the high differential pressures experienced during heavy grassing events. The analysis provided a stress calculation to determine the limiting differential pressure across the SW strainer.

The analysis resulted in a limiting differential pressure of 101.7 pounds per square inch differential. PSEG identified within the analysis that cyclic fatigue failure of the strainers was not modeled; therefore, maintenance plans were created to replace the strainers at a six-year interval based on past data from strainer failures.

The team evaluated the analysis to ensure that the licensing bases, design bases, and performance capability of the SW system had not been adversely affected by the analysis. The team reviewed operating procedures to ensure they had been properly updated to incorporate the results of the analysis. Drawings were reviewed to verify inputs into the analysis to ensure proper modeling of the strainers. The team discussed the analysis and system design basis with both design and system engineers to assess the adequacy and results of the analysis. In addition, the team verified the SW strainer instrumentation was adequate for the increased differential pressure. The team reviewed strainer differential pressure trending data and NOTFs to ensure that the SW strainer design function was maintained and that margin existed to the limiting differential pressure. Finally, the team reviewed the strainer maintenance plan to ensure the strainer replacement is within the six-year interval. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findinos No findings were identified.

2j0 Safetv and Turbine Auxiliaries Coolino Svstem Looic Modification a. lnspection Scope The team reviewed a modification (80098513) that added an automatic function to close the turbine auxiliary cooling return valve (HV-2496) when the associated supply valve (HV- 2522) inadvertently closes due to failure. These valves function to isolate the non-safety related from the safety-related loads. PSEG implemented this modification to prevent a complete loss of the safety and turbine auxiliaries cooling system (STACS)due to a low expansion tank level, which occurred during past failures of HV-2522. The complete loss of STACS results in a reactor trip. The STACS expansion tank low level, which isolates the corresponding loop, is caused by system sluicing from the operating loop to the standby loop that occurs during a system load swap when HV-2522 closes' The team's review was performed to verify that the design and licensing bases, and performance capability of the STACS had not been degraded by the modification. The team reviewed a technical evaluation and the system operating parameters, such as valve sequencing, expansion tank levels, system flowrate, and instrumentation setpoints to verify that margin existed between the low expansion tank level isolation during loop swap. in addition post-modification testing was reviewed to verify proper operation of STACS. The team reviewed drawings, procedures, and training documents to ensure that they were properly updated. Also, the team reviewed documentation to ensure the plant simulator was updated with the modification. The team discussed the modification and design basis witfr design and system engineers to assess the adequacy of the modification. In addition, the 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1R17.1of this report' The documents reviewed are listed in the attachment.

b.

Findinqs No findings were identified.

2.11 lmpact

a. Inspection Scope

The team reviewed a modification (80102144) that revised calculation H-1-ZZ-MDC-1880 to incorporate a closure time reduction in numerous primary containment isolation valves along with a change in the methodology of radiological transport of main steam isolation valve leakage. ine modification was implemented to support the approval of License AmgndmeniRequest H09-01 for the use of Cobalt-60 lsotope Test Assemblies in the reactor core. Calculation H-l-ZZ-MDC-1880 evaluates the post loss-of-coolant accident (LOCA) offsite and control room radiological impact. This modification also incorporated revisions to H-1-ZZ-MDC-1923 and H-1-ZZ-MDC-1927 , due to the change in H-1-ZZ-MDC-1880. Calculation H-1-ZZ-MDC-1923 addresses radiological impact to areas requiring continuous occupancy during a design bases event and H-1-ZZ-MDC-1927 addresses post-LOCA mission doses to various vital areas in the plant.

The team conducted the review to assess whether the calculation revision was consistent with assumptions in the design and licensing bases. The team reviewed the associated revision to the calculations to assess their adequacy and results. The team verified the results and methodology were in accordance with 10 CFR 50.67 and Regulatory Guide 1

.183 . In addition, the calculations and results were discussed with

the responsible engineer to verify inputs and assumptions were appropriate. The 10 CFR 50.59 safety evaluation associated with this modification was also reviewed as described in section 1R17.1of this report. The documents reviewed are listed in the attachment.

b. Findin,os No findings were identified.

OTHER ACTIVITIES

4OA2 ldentification and Resolution of Problems (lP 71152)a. lnspection Scope The team reviewed a sample of NOTFs associated with 10 CFR 50.59 and plant modification issues to determine whether PSEG was appropriately identifying, characterizing, and correcting problems associated with these areas, and whether the planned or completed corrective actions were appropriate. In addition, the team reviewed NOTFs written on issues identified during the inspection to verify adequate problem identification and incorporation of the problem into the corrective action system.

The NOTFs reviewed are listed in the attachment.

b. Findinqs No findings were identified.

4OAO Meetinos. includinq Exit The team presented the inspection results to Mr. J. Perry, Site Vice President, and other members of PSEG's staff at an exit meeting on May 6,2011. The team returned the proprietary information reviewed during the inspection and verified that this report does not contain proprietary information.

ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

PSEG Personnel

O. Arroyo, Systems Engineer

R. Bhai, Design Engineer

A. Bhuta, Design Engineer

W. Bischoff, Systems Engineer

J. Boyer, Mechanical Engineering Design Manager

D. Boyle, Operations Support Manager

D. Bush, Systems Engineer

P. Duca, Regulatory Assurance Engineer

J. Duffy, Design Engineer

M. Fowler, Engineering Design Manager

A. Ghose, Design Engineer

Y. Ghotok, Systems Engineer

K. Knaide, Engineering Director

J. Moss, Design Engineer

J. Lane, Design Engineer

J. Perry, Hope Creek Site Vice President

G. Siefert, Design Engineer

B. Tarr, Design Engineer

K. Torres, Systems Engineer

K. Yeanrood, Regulatory Assurance Engineer

M. Zimmerman, Design Engineer

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

None