IR 05000334/2010006: Difference between revisions

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{{Adams|number = ML102150492}}
{{Adams
| number = ML102150492
| issue date = 08/03/2010
| title = IR 05000334-10-006, 05000412-10-006 on 6/07/2010 - 6/24/2010 for Beaver Valley Power Station, Units 1 & 2, Engineering Specialist Plant Modifications Inspection
| author name = Doerflein L T
| author affiliation = NRC/RGN-I/DRS/EB2
| addressee name = Harden P
| addressee affiliation = FirstEnergy Nuclear Operating Co
| docket = 05000334, 05000412
| license number = DPR-066, NPF-073
| contact person =
| document report number = IR-10-006
| document type = Inspection Report, Letter
| page count = 18
}}


{{IR-Nav| site = 05000334 | year = 2010 | report number = 006 }}
{{IR-Nav| site = 05000334 | year = 2010 | report number = 006 }}


=Text=
=Text=
{{#Wiki_filter:Mr. Paul Harden Site Vice President August 3, 2010 FirstEnergy Nuclear Operating Company Beaver Valley Power Station P.O. Box 4, Route 168 Shippingport, PA 15077
{{#Wiki_filter:Mr. Paul Harden Site Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PA 19406-1415 August 3, 2010 FirstEnergy Nuclear Operating Company Beaver Valley Power Station P.O. Box 4, Route 168 Shippingport, PA 15077


SUBJECT: BEAVER VALLEY POWER STATION -NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000334/2010006 AND 05000412/2010006
SUBJECT: BEAVER VALLEY POWER STATION -NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000334/2010006 AND 05000412/2010006


==Dear Mr. Harden:==
==Dear Mr. Harden:==
On June 24,2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Beaver Valley Power Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on June 24, 2010, with Mr. Ray Lieb, Director Site Operations, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 Document Room or from the Publicly Available Records (PARS) component of the 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: 50-334, 50-412 License Nos: DPR-66, NPF-73  
On June 24,2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Beaver Valley Power Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on June 24, 2010, with Mr. Ray Lieb, Director Site Operations, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 Document Room or from the Publicly Available Records (PARS) component of the 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: 50-334, 50-412 License Nos: DPR-66, NPF-73  


Sincerely,Lawrence T. Doerflein, Chi Engineering Branch 2 Division of Reactor Safety  
Sincerely,Lawrence T. Doerflein, Chi Engineering Branch 2 Division of Reactor Safety  
Line 23: Line 41:


==Dear Mr. Harden:==
==Dear Mr. Harden:==
On June 24, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Beaver Valley Power Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on June 24, 2010, with Mr. Ray Lieb, Director Site Operations, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 Document Room or from the Publicly Available Records (PARS) component of the 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: 50-334, 50-412 License Nos: DPR-66, NPF-73  
On June 24, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Beaver Valley Power Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on June 24, 2010, with Mr. Ray Lieb, Director Site Operations, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 Document Room or from the Publicly Available Records (PARS) component of the 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: 50-334, 50-412 License Nos: DPR-66, NPF-73  


Sincerely,IRA! Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety  
Sincerely,IRA! Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety  
Line 31: Line 53:


===w/Attachment:===
===w/Attachment:===
Supplemental Information cc w/encl: Distribution via ListServ ADAMS ACCESSION NO. ML 102150492 SUNSI Review Complete: Initials) DOCUMENT NAME: Y:\Division\DRS\Engineering Branch 2\Arner\beaver valley mods\BVModsReport2010006.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: "ell C = Opy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE RI/DRS I RI/DRP I RI/DRS I I NAME FArnerl RBellamvl LDoerfieinl DATE 07/23/10 07/30/10 08/3110 OFFICIAL RECORD COpy Distribution w/encl: M. Dapas, Acting RA (R10RAMAIL Resource) D. Lew, Acting DRA (R10RAMAIL Resource) J. Clifford, DRP (R1DRPMAIL Resource) D. Collins, DRP (R1DRPMAIL Resource) D. Roberts, DRS (R1DRSMaii Resource) P. Wilson, DRS (R1DRSMaii Resource) R. Bellamy, DRP S. Barber, DRP C. Newport, DRP D. Werkheiser, SRI E. Bonney, RI P. Garrett, Resident OA L. Trocine, RI OEDO D. Bearde, DRS RidsNRRPMBeaverValley Resource RidsNrrDorlLpl1-1 Resource@nrc.gov ROPreportsResource@nrc.gov L. Doerflein, DRS F. Arner, DRS U.S. NUCLEAR REGULATORY COMMISSION REGION I Docket No.: 50-334, 50-412 License No.: DPR-66, NPF-73 Report No.: 05000334/2010006 and 05000412/2010006 Licensee: FirstEnergy Nuclear Operating Company (FENOC) Facility: Beaver Valley Power Station, Units 1 and 2 Location: Post Office Box 4 Shippingport, PA 15077 Inspection Period: June 7 through June 24, 2010 Inspectors: F. Arner, Senior Reactor Inspector, Division of Reactor Safety (DRS), Team Leader L. Scholl, Senior Reactor Inspector, DRS E. Burkett, Reactor Inspector, DRS A. Dugandzic, Reactor Engineer, DRS (in-training) Approved By: Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Enclosure  
Supplemental Information cc w/encl: Distribution via ListServ ADAMS ACCESSION NO. ML 102150492 SUNSI Review Complete:
Initials)
DOCUMENT NAME: Y:\Division\DRS\Engineering Branch 2\Arner\beaver valley mods\BVModsReport2010006.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: "ell C = Opy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE RI/DRS I RI/DRP I RI/DRS I I NAME FArnerl RBellamvl LDoerfieinl DATE 07/23/10 07/30/10 08/3110 OFFICIAL RECORD COpy Distribution w/encl: M. Dapas, Acting RA (R10RAMAIL Resource)
D. Lew, Acting DRA (R10RAMAIL Resource)
J. Clifford, DRP (R1DRPMAIL Resource)
D. Collins, DRP (R1DRPMAIL Resource)
D. Roberts, DRS (R1DRSMaii Resource)
P. Wilson, DRS (R1DRSMaii Resource)
R. Bellamy, DRP S. Barber, DRP C. Newport, DRP D. Werkheiser, SRI E. Bonney, RI P. Garrett, Resident OA L. Trocine, RI OEDO D. Bearde, DRS RidsNRRPMBeaverValley Resource RidsNrrDorlLpl1-1 Resource@nrc.gov ROPreportsResource@nrc.gov L. Doerflein, DRS F. Arner, DRS U.S. NUCLEAR REGULATORY COMMISSION REGION I Docket No.: 50-334, 50-412 License No.: DPR-66, NPF-73 Report No.: 05000334/2010006 and 05000412/2010006 Licensee:
FirstEnergy Nuclear Operating Company (FENOC) Facility:
Beaver Valley Power Station, Units 1 and 2 Location:
Post Office Box 4 Shippingport, PA 15077 Inspection Period: June 7 through June 24, 2010 Inspectors:
F. Arner, Senior Reactor Inspector, Division of Reactor Safety (DRS), Team Leader L. Scholl, Senior Reactor Inspector, DRS E. Burkett, Reactor Inspector, DRS A. Dugandzic, Reactor Engineer, DRS (in-training)
Approved By: Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Enclosure  


=SUMMARY OF FINDINGS=
=SUMMARY OF FINDINGS=
IR 05000334/2010006; IR 05000412/2010006; 6/07/2010 -6/24/2010; FirstEnergy Nuclear Operating Company (FENOC); Beaver Valley Power Station, Units 1 & 2; 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 and one inspector in training. 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. NRG-Identified and Self-Revealing Findings No findings of significance were identified.
IR 05000334/2010006;  
 
IR 05000412/2010006; 6/07/2010  
-6/24/2010;
FirstEnergy Nuclear Operating Company (FENOC); Beaver Valley Power Station, Units 1 & 2; 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 and one inspector in training.
 
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. NRG-Identified and Self-Revealing Findings No findings of significance were identified.


===B. Licensee-Identified Violations===
===B. Licensee-Identified Violations===
None. ii
None. ii


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==REACTOR SAFETY==
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1 R 17 Evaluations of Changes. Tests. or Experiments and Permanent Plant Modifications (IP 71111.17)
Cornerstones:
Initiating Events, Mitigating Systems, and Barrier Integrity 1 R 17 Evaluations of Changes. Tests. or Experiments and Permanent Plant Modifications (IP 71111.17)


===.1 Evaluations of Changes. Tests. or Experiments (28 samples)===
===.1 Evaluations===
 
of Changes. Tests. or Experiments (28 samples)


====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed four 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 FENOC 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-four 10 CFR 50.59 screenings for which FENOC had concluded that no safety evaluation was required. These reviews were performed to assess whether FENOC'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 FENOC had performed and approved during the time period covered by this inspection (Le., since the last modifications inspection) not previously reviewed by NRC inspectors. The screenings were selected based on the safety significance, risk significance, and complexity of the change to the facility. In addition, the team compared FENOC'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.
The team reviewed four 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 FENOC 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-four 10 CFR 50.59 screenings for which FENOC had concluded that no safety evaluation was required.
 
These reviews were performed to assess whether FENOC'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 FENOC had performed and approved during the time period covered by this inspection (Le., since the last modifications inspection)not previously reviewed by NRC inspectors.
 
The screenings were selected based on the safety significance, risk significance, and complexity of the change to the facility.
 
In addition, the team compared FENOC'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. Findings====
====b. Findings====
No findings of significance were identified. Enclosure 2
No findings of significance were identified.


===.2 Permanent Plant Modifications (9 samples) .2.1 Unit 1 Emergency Diesel Generator (EDG) K 1 Relay Replacement*===
2
 
===.2 Permanent===
 
Plant Modifications (9 samples)
 
===.2.1 Unit 1 Emergency===
 
Diesel Generator (EDG) K 1 Relay Replacement*


====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a modification (Engineering Change Package (ECP) 08-0095) that replaced the original latching style relays with an electrically equivalent relay without the latching feature. The K1 relay design function is to shunt the field circuit during the EDG shutdown sequence and thereby terminate the generator electrical output in a controlled manner. The original relay would remain latched closed until unlatched during the next EDG startup sequence. One reason for the replacement was that the original style relays were found to be potentially subject to auxiliary contact terminals coming loose during a seismic event. Additionally, Unit 2 had experienced a problem with this latching type relay when it failed to release during an EDG start sequence and thereby rendered the EDG inoperable. This modification also added indicating lights to allow operators to verify the K1 relay opened as designed at the end of the shutdown sequence indicating that the excitation system was properly aligned for the next EDG start. An indicator light was also added to allow monitoring of the K2 field flash relay in a similar manner. The team conducted the review to verify that the design bases, licensing bases, and performance capability of the EDGs were not degraded by the component replacements and circuit modifications. The team discussed the modification and design basis with design and system engineers, and reviewed the scope and results of post-modification testing to assess the capability of the EDGs to perform their safety function during a design basis event. The team also confirmed that surveillance tests, operational procedures, and drawings had been appropriately updated to reflect the design change. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.
The team reviewed a modification (Engineering Change Package (ECP) 08-0095) that replaced the original latching style relays with an electrically equivalent relay without the latching feature. The K1 relay design function is to shunt the field circuit during the EDG shutdown sequence and thereby terminate the generator electrical output in a controlled manner. The original relay would remain latched closed until unlatched during the next EDG startup sequence.
 
One reason for the replacement was that the original style relays were found to be potentially subject to auxiliary contact terminals coming loose during a seismic event. Additionally, Unit 2 had experienced a problem with this latching type relay when it failed to release during an EDG start sequence and thereby rendered the EDG inoperable.
 
This modification also added indicating lights to allow operators to verify the K1 relay opened as designed at the end of the shutdown sequence indicating that the excitation system was properly aligned for the next EDG start. An indicator light was also added to allow monitoring of the K2 field flash relay in a similar manner. The team conducted the review to verify that the design bases, licensing bases, and performance capability of the EDGs were not degraded by the component replacements and circuit modifications.
 
The team discussed the modification and design basis with design and system engineers, and reviewed the scope and results of post-modification testing to assess the capability of the EDGs to perform their safety function during a design basis event. The team also confirmed that surveillance tests, operational procedures, and drawings had been appropriately updated to reflect the design change. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a modification (ECP-10-00010) that replaced an obsolete containment sump discharge flow instrumentation system with an equivalent model. The scope of the modification included the replacement of the flow tube and flow transmitter, and changed the location of its electrical power feed. The instrument provides a method of determining the amount of unidentified leakage from systems located in the containment building as required by plant technical specification 3.4.15, RCS Leakage Detection Instrumentation. Enclosure 3 The team conducted the review to ensure that the design bases, licensing bases, and performance capability of the leakage monitoring system had not been adversely affected by the modification. The team reviewed the associated work order packages, and conducted interviews with design and system engineers regarding the design, installation, calibration, and testing of the instrumentation to verify that the modification was adequate. The team walked down the accessible portions of the new equipment to ensure the system configuration was in accordance with the design instructions. The team also verified that drawings and operating procedures had been appropriately revised to reflect the change. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.
The team reviewed a modification (ECP-10-00010)that replaced an obsolete containment sump discharge flow instrumentation system with an equivalent model. The scope of the modification included the replacement of the flow tube and flow transmitter, and changed the location of its electrical power feed. The instrument provides a method of determining the amount of unidentified leakage from systems located in the containment building as required by plant technical specification 3.4.15, RCS Leakage Detection Instrumentation.
 
3 The team conducted the review to ensure that the design bases, licensing bases, and performance capability of the leakage monitoring system had not been adversely affected by the modification.
 
The team reviewed the associated work order packages, and conducted interviews with design and system engineers regarding the design, installation, calibration, and testing of the instrumentation to verify that the modification was adequate.
 
The team walked down the accessible portions of the new equipment to ensure the system configuration was in accordance with the design instructions.
 
The team also verified that drawings and operating procedures had been appropriately revised to reflect the change. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a modification (ECP-08-0134) performed by FENOC to install a local test switch in the reactor trip switchgear. The test switch allows local closing of the individual reactor trip breakers during surveillance testing instead of both trip breakers (the breaker under test and the other in-service breaker) receiving a closing signal when the breaker is closed from the control room switch. As a result, the in-service breaker is not unnecessarily subjected to the reverse inductive voltage spike during testing. The team evaluated the change to confirm that the design bases, licensing bases, and performance capability of the reactor trip system would not be affected by the change. The team also walked down accessible portions of the system to assess the configuration and material condition of the system, and reviewed updated surveillance procedures to ensure they included appropriate directions for the technicians operating the switches during testing. The team interviewed design engineers to review the design change and the potential impact on proper system operation. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.
The team reviewed a modification (ECP-08-0134)performed by FENOC to install a local test switch in the reactor trip switchgear.
 
The test switch allows local closing of the individual reactor trip breakers during surveillance testing instead of both trip breakers (the breaker under test and the other in-service breaker) receiving a closing signal when the breaker is closed from the control room switch. As a result, the in-service breaker is not unnecessarily subjected to the reverse inductive voltage spike during testing. The team evaluated the change to confirm that the design bases, licensing bases, and performance capability of the reactor trip system would not be affected by the change. The team also walked down accessible portions of the system to assess the configuration and material condition of the system, and reviewed updated surveillance procedures to ensure they included appropriate directions for the technicians operating the switches during testing. The team interviewed design engineers to review the design change and the potential impact on proper system operation.
 
The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
No findings of significance were identified. Enclosure 4
No findings of significance were identified.


===.2.4 Refueling Water Storage Tank (RWST) Level Interlock to Recirculation Spray (RS) Pump Start===
4
 
===.2.4 Refueling===
 
Water Storage Tank (RWST) Level Interlock to Recirculation Spray (RS) Pump Start


====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a modification (ECP 06-0227) that revised the automatic start signal for the RS pumps at both units to be initiated from an RWST level input. FENOC implemented the modification to regain RS pump net positive suction head (NPSH) margin that was reduced following the installation of modified containment sump strainers, which had been installed to address potential debris accumulation. Prior to this modification, the RS pumps started automatically following a containment high-high pressure signal and a time delay. The modification included adding main control room benchboard status lights and changing electrical wiring for the RS pump start circuit and relay logic. The team conducted the review to verify that the design bases, licensing bases and performance capability of the recirculation spray pumps had not been adversely affected by the modification. The team reviewed drawings, calculations, operator training modules, and procedures to ensure they had been properly updated to incorporate the changes to the RS pump start logic. In addition, post-modification testing was reviewed to verify the logic sequence and automatic RS pump start were appropriately tested. The team walked down portions of the modification to verify that status lights and annunciators were added to the main control room benchboards. The team also discussed the modification and design basis with design engineers to assess the adequacy of the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R 17.1 of this report. The documents reviewed are listed in the attachment.
The team reviewed a modification (ECP 06-0227) that revised the automatic start signal for the RS pumps at both units to be initiated from an RWST level input. FENOC implemented the modification to regain RS pump net positive suction head (NPSH) margin that was reduced following the installation of modified containment sump strainers, which had been installed to address potential debris accumulation.
 
Prior to this modification, the RS pumps started automatically following a containment high-high pressure signal and a time delay. The modification included adding main control room benchboard status lights and changing electrical wiring for the RS pump start circuit and relay logic. The team conducted the review to verify that the design bases, licensing bases and performance capability of the recirculation spray pumps had not been adversely affected by the modification.
 
The team reviewed drawings, calculations, operator training modules, and procedures to ensure they had been properly updated to incorporate the changes to the RS pump start logic. In addition, post-modification testing was reviewed to verify the logic sequence and automatic RS pump start were appropriately tested. The team walked down portions of the modification to verify that status lights and annunciators were added to the main control room benchboards.
 
The team also discussed the modification and design basis with design engineers to assess the adequacy of the modification.
 
The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R 17.1 of this report. The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed an analysis (ECP 09-03277) performed by FENOC to evaluate the fouling factor of the Unit 2 RS heat exchangers and the potential for air entrapment within the heat exchangers. This analysis was performed to ensure the design fouling factor assumption for the two RS heat exchangers cooled by service water was appropriate. FENOC performed the analysis using the Electric Power Research Institute (EPRI) Modular Accident Analysis Program (MAAP) to demonstrate the fouling factor was acceptable for component performance during design basis accident conditions. Additionally, FENOC evaluated the internal shell side velocities of the heat exchangers to determine if they were sufficient to sweep air from the system during Enclosure 5 initial fill. Using a Froude number criterion, FENOC determined that air would be swept from the system upon initial fill during system initiation, and not impede HX performance. The team reviewed the analysis and associated calculations to verify the assumptions and parameters used were valid and considered worst case scenarios. The parameters included service water temperature, containment temperature and pressure, number of plugged tubes, sump temperature, and heat exchanger baffle arrangement. The documents reviewed are listed in the attachment.
The team reviewed an analysis (ECP 09-03277)performed by FENOC to evaluate the fouling factor of the Unit 2 RS heat exchangers and the potential for air entrapment within the heat exchangers.
 
This analysis was performed to ensure the design fouling factor assumption for the two RS heat exchangers cooled by service water was appropriate.
 
FENOC performed the analysis using the Electric Power Research Institute (EPRI) Modular Accident Analysis Program (MAAP) to demonstrate the fouling factor was acceptable for component performance during design basis accident conditions.
 
Additionally, FENOC evaluated the internal shell side velocities of the heat exchangers to determine if they were sufficient to sweep air from the system during Enclosure 5 initial fill. Using a Froude number criterion, FENOC determined that air would be swept from the system upon initial fill during system initiation, and not impede HX performance.
 
The team reviewed the analysis and associated calculations to verify the assumptions and parameters used were valid and considered worst case scenarios.
 
The parameters included service water temperature, containment temperature and pressure, number of plugged tubes, sump temperature, and heat exchanger baffle arrangement.
 
The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a modification (ECP 07-0076) that replaced the selector switches for the TDAFW pump steam admission solenoid operated valves in the Unit 2 main control room. FENOC implemented this modification to address a postulated accident scenario such as a steam generator tube rupture (SGTR) event which would require plant operators to shut down the TDAFW pump from the control room using class 1 E safety related controls. Prior to the modification, the selector switches had" return-to-center" operators that prevented the steam admission valves from maintaining a closed position which may be necessary during certain design basis events. The replacement switches were designed to maintain the position in which they have been placed. The modification included installing an annunciator in the main control room, and replacing and rewiring the control switches. The team reviewed operating procedures, operator training modules, and drawings to ensure they had been properly updated. The team reviewed work order packages and conducted interviews with design and system engineers regarding the design, installation, and testing to verify that the modification was adequate. The team performed a walkdown of the modification in the main control room and discussed the alarm response procedure with the operators. The documents reviewed are listed in the attachment.
The team reviewed a modification (ECP 07-0076) that replaced the selector switches for the TDAFW pump steam admission solenoid operated valves in the Unit 2 main control room. FENOC implemented this modification to address a postulated accident scenario such as a steam generator tube rupture (SGTR) event which would require plant operators to shut down the TDAFW pump from the control room using class 1 E safety related controls.
 
Prior to the modification, the selector switches had" return-to-center" operators that prevented the steam admission valves from maintaining a closed position which may be necessary during certain design basis events. The replacement switches were designed to maintain the position in which they have been placed. The modification included installing an annunciator in the main control room, and replacing and rewiring the control switches.
 
The team reviewed operating procedures, operator training modules, and drawings to ensure they had been properly updated. The team reviewed work order packages and conducted interviews with design and system engineers regarding the design, installation, and testing to verify that the modification was adequate.
 
The team performed a walkdown of the modification in the main control room and discussed the alarm response procedure with the operators.
 
The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
No findings of significance were identified. Enclosure 6
No findings of significance were identified.


===.2.7 Unit 2 Joint Owners Group (JOG) Motor Operated Valve (MOV) Periodic Verification Program Implementation===
6
 
===.2.7 Unit 2 Joint Owners Group (JOG) Motor Operated Valve (MOV) Periodic Verification===
 
Program Implementation


====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a modification (ECP 08-0504) that changed the setup for motor operated valve (MOV) actuators associated with the Unit 2 safety injection (SI) system. FENOC implemented this modification to incorporate requirements of the JOG MOV program which is an industry-wide response to Generic Letter (GL) 1996-05, " Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves." The team selected a sample of two valves to review the changes performed. The valves reviewed were 2SIS-MOV867 A; high head SI valve, and 2SIS-MOV869A, hot leg SI isolation valve. The actuator control switch setup was modified to account for increased torque and thrust requirements, and reduced stress relative to the weak link components. Additionally, the control switch for the closure function for valve 2SIS-MOV867 A was changed from torque switch control to limit switch control. The team conducted the review to verify that the design bases and performance capability of the safety injection system had not been adversely affected by the modification. The team reviewed associated drawings and calculations to ensure they were properly updated and the correct assumptions were used. In addition, the team reviewed completed work orders to verify the appropriateness of the testing and acceptance criteria. The team also discussed the modification and design bases with design and system engineers to assess the adequacy of the modification. The documents reviewed are listed in the attachment.
The team reviewed a modification (ECP 08-0504) that changed the setup for motor operated valve (MOV) actuators associated with the Unit 2 safety injection (SI) system. FENOC implemented this modification to incorporate requirements of the JOG MOV program which is an industry-wide response to Generic Letter (GL) 1996-05, " Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves." The team selected a sample of two valves to review the changes performed.
 
The valves reviewed were 2SIS-MOV867 A; high head SI valve, and 2SIS-MOV869A, hot leg SI isolation valve. The actuator control switch setup was modified to account for increased torque and thrust requirements, and reduced stress relative to the weak link components.
 
Additionally, the control switch for the closure function for valve 2SIS-MOV867 A was changed from torque switch control to limit switch control. The team conducted the review to verify that the design bases and performance capability of the safety injection system had not been adversely affected by the modification.
 
The team reviewed associated drawings and calculations to ensure they were properly updated and the correct assumptions were used. In addition, the team reviewed completed work orders to verify the appropriateness of the testing and acceptance criteria.
 
The team also discussed the modification and design bases with design and system engineers to assess the adequacy of the modification.
 
The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
Line 101: Line 236:


====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a modification (ECP 07-0145) associated with the installation of a plug in the vent hole on the two quench spray (as) system loop seals and the repositioning of drain valves in the QS piping from open to normally closed. The modification was implemented in order to prevent air ingestion into the recirculation spray (RS) and low head safety injection (LHSI) pumps through the containment sump plenums. This was required due to the revised containment sump design, which had created the potential for air ingestion from the as system. Specifically, several of the as branch lines penetrate and discharge into the containment sump plenum. The design intent of those lines was to increase NPSH margin by adding the cooler QS water initially to the containment-sump. This modification also identified that the QS Enclosure 7 piping loop seal to the containment sump would be required to be kept full to prevent air ingestion during a postulated accident which would require containment sump recirculation. The team reviewed the associated calculations to ensure the methods used were reasonable in determining that the loop seal would be able to prevent air ingestion during postulated accident scenarios. Specifically, the team reviewed the design input assumption for containment pressure to ensure the calculated height of the loop seal would be adequate at the time the as pumps are secured after RWST depletion. The team reviewed emergency operating procedures to verify that the containment pressure design input was reasonable. Additionally, the team reviewed operating procedures established to ensure that the as loop seal would be adequately filled prior to operation as well as FENOC' s assessment for potential leakage impacting the design conclusions. The team also reviewed the response time change associated with the as system because of the change regarding the loop seal being maintained water solid prior to operation. This review was performed to ensure that surveillance tests and calculations had been revised appropriately. The team interviewed design engineers to review the design change and the potential impact on proper system operation. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.
The team reviewed a modification (ECP 07-0145) associated with the installation of a plug in the vent hole on the two quench spray (as) system loop seals and the repositioning of drain valves in the QS piping from open to normally closed. The modification was implemented in order to prevent air ingestion into the recirculation spray (RS) and low head safety injection (LHSI) pumps through the containment sump plenums. This was required due to the revised containment sump design, which had created the potential for air ingestion from the as system. Specifically, several of the as branch lines penetrate and discharge into the containment sump plenum. The design intent of those lines was to increase NPSH margin by adding the cooler QS water initially to the containment-sump.
 
This modification also identified that the QS Enclosure 7 piping loop seal to the containment sump would be required to be kept full to prevent air ingestion during a postulated accident which would require containment sump recirculation.
 
The team reviewed the associated calculations to ensure the methods used were reasonable in determining that the loop seal would be able to prevent air ingestion during postulated accident scenarios.
 
Specifically, the team reviewed the design input assumption for containment pressure to ensure the calculated height of the loop seal would be adequate at the time the as pumps are secured after RWST depletion.
 
The team reviewed emergency operating procedures to verify that the containment pressure design input was reasonable.
 
Additionally, the team reviewed operating procedures established to ensure that the as loop seal would be adequately filled prior to operation as well as FENOC' s assessment for potential leakage impacting the design conclusions.
 
The team also reviewed the response time change associated with the as system because of the change regarding the loop seal being maintained water solid prior to operation.
 
This review was performed to ensure that surveillance tests and calculations had been revised appropriately.
 
The team interviewed design engineers to review the design change and the potential impact on proper system operation.
 
The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
Line 107: Line 260:


====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a revision to design calculation 10080-N-584 that changed the differential pressure (DP) and line pressure requirements for SI valves due to the replacement of the Unit 2 charging pump rotating assemblies. The team reviewed FENOC' s methodology used in determining the revised DP parameters for various valves within the system. The team reviewed selected design inputs used in the revised calculation to verify they were reasonable. The design inputs verified included the static pressure affect from the refueling water storage tank, the containment pressure assumption at the time of swapover to sump recirculation, and the static elevation differences between the pumps and valves. The charging pump performance curve design input was also verified to have been correctly incorporated within the calculations with respect to the change of performance due to the replaced rotating assembly. The emergency operating procedures were reviewed to verify equipment alignments assumed in the determination of valve differential pressures were in accordance with procedure direction. Enclosure 8 The team sampled a few of the revised design DP outputs to ensure that the information was appropriately translated into torque calculations for Unit 2 SI valve 2SIS-MOV8811A, which automatically opens to provide a recirculation path from the containment sump via the recirculation pumps, during the recirculation phase of-safety injection. Additionally, the team verified that the revised DP output was appropriately translated to the torque calculation for 2SIS-MOV863B, which automatically opens during the recirculation phase to divert the flow of the recirculation spray pumps to the suction of the high head safety injection (HHSI)/charging pumps. The team performed a walkdown of the SI inlet isolation valve, to assess the material condition of the valve. The team also performed interviews with the responsible motor operated valve engineers to determine whether the design changes appeared reasonable. The documents reviewed are listed in the attachment.
The team reviewed a revision to design calculation 10080-N-584 that changed the differential pressure (DP) and line pressure requirements for SI valves due to the replacement of the Unit 2 charging pump rotating assemblies.
 
The team reviewed FENOC' s methodology used in determining the revised DP parameters for various valves within the system. The team reviewed selected design inputs used in the revised calculation to verify they were reasonable.
 
The design inputs verified included the static pressure affect from the refueling water storage tank, the containment pressure assumption at the time of swapover to sump recirculation, and the static elevation differences between the pumps and valves. The charging pump performance curve design input was also verified to have been correctly incorporated within the calculations with respect to the change of performance due to the replaced rotating assembly.
 
The emergency operating procedures were reviewed to verify equipment alignments assumed in the determination of valve differential pressures were in accordance with procedure direction.
 
8 The team sampled a few of the revised design DP outputs to ensure that the information was appropriately translated into torque calculations for Unit 2 SI valve 2SIS-MOV8811A, which automatically opens to provide a recirculation path from the containment sump via the recirculation pumps, during the recirculation phase of-safety injection.
 
Additionally, the team verified that the revised DP output was appropriately translated to the torque calculation for 2SIS-MOV863B, which automatically opens during the recirculation phase to divert the flow of the recirculation spray pumps to the suction of the high head safety injection (HHSI)/charging pumps. The team performed a walkdown of the SI inlet isolation valve, to assess the material condition of the valve. The team also performed interviews with the responsible motor operated valve engineers to determine whether the design changes appeared reasonable.
 
The documents reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
Line 116: Line 281:


====a. Inspection Scope====
====a. Inspection Scope====
The team reviewed a sample of the condition reports (CRs) associated with 10 CFR 50.59 and plant modification issues to determine whether FENOC 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 CRs written on issues identified during the inspection to verify adequate problem identification and incorporation of the problem into the corrective action system. The CRs reviewed are listed in the attachment.
The team reviewed a sample of the condition reports (CRs) associated with 10 CFR 50.59 and plant modification issues to determine whether FENOC 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 CRs written on issues identified during the inspection to verify adequate problem identification and incorporation of the problem into the corrective action system. The CRs reviewed are listed in the attachment.


====b. Findings====
====b. Findings====
No findings of significance were identified. 40A6 Meetings, including Exit The team presented the inspection results to Mr. Ray Lieb, Director, Site Operations, and other members of FENOC's staff at an exit meeting on June 24, 2010. The team returned the proprietary information reviewed during the inspection to the licensee and verified that this report does not contain proprietary information. Enclosure A-1 ATTACHMENT
No findings of significance were identified.


=SUPPLEMENTAL INFORMATION=
40A6 Meetings, including Exit The team presented the inspection results to Mr. Ray Lieb, Director, Site Operations, and other members of FENOC's staff at an exit meeting on June 24, 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 ATTACHMENT
 
=SUPPLEMENTAL
INFORMATION=


==KEY POINTS OF CONTACT==
==KEY POINTS OF CONTACT==
FENOC Personnel  
FENOC Personnel  
: [[contact::R. Fedin]], Regulatory Compliance  
: [[contact::R. Fedin]], Regulatory
: [[contact::C. Mancuso]], Design Engineering Manager  
Compliance  
: [[contact::D. Price]], Design Engineering Supervisor  
: [[contact::C. Mancuso]], Design Engineering
Manager  
: [[contact::D. Price]], Design Engineering
Supervisor  
: [[contact::K. Woessner]], Nuclear Staff Engineer  
: [[contact::K. Woessner]], Nuclear Staff Engineer  
==LIST OF ITEMS==
==LIST OF ITEMS==
OPENED, CLOSED AND DISCUSSED None  
 
==LIST OF DOCUMENTS REVIEWED==
===OPENED, CLOSED AND DISCUSSED===
10
 
: CFR 50.59 Evaluations 07-00157, Impact of Reduced Atmospheric Relief Dump Valve (ASDV) Capacity, Rev. 0 07 -05204, Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project, Rev. 0 08-05643, BV2 Primary Zinc Addition, Rev. 0 09-02468, BV2 UFSAR Change -Containment Sump Screen Passive Failure, Rev. 0 10
None  
: CFR 50.59 Screened-out Evaluations 07-00275, Reactor Coolant Temperature Loop 2RCS-T412 Delta T-Tavg Protection Channell Calibration, Rev. 0 07-00521, Containment Coatings Walkdown, Rev. 0 07-05146, 18-Month Slave Relay Testing (Train B), Rev. 0 08-02593, Elimination of Recirculation Spray Pump Response Time SR 3.3.2.9, Rev. 1 08-05721, Replace Valves 2RCS-633 & 2RCS-634 with Pipe Spool Piece, Rev. 0 09-03164, Quench Spray System Trouble, Rev. 0 09-04690, RHS Fill and Vent from RWST, Rev. 0 08-05359, BV2 EDG ASP Transfer Circuit Relay Change, Rev. 0 09-00048, Station Battery Jumper Installation and Removal, Rev. 0 09-00238, Calculation 8700-DMC-1430 Revision, Rev. 0 09-00374, Incipient Fault Monitor Repair, Rev. 0 09-01089, 1
==LIST OF DOCUMENTS==
REVIEWED 10
: CFR 50.59 Evaluations  
: 07-00157, Impact of Reduced Atmospheric Relief Dump Valve (ASDV) Capacity, Rev. 0 07 -05204, Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project, Rev. 0 08-05643, BV2 Primary Zinc Addition, Rev. 0 09-02468, BV2 UFSAR Change -Containment Sump Screen Passive Failure, Rev. 0 10
: CFR 50.59 Screened-out Evaluations  
: 07-00275, Reactor Coolant Temperature Loop 2RCS-T412  
: Delta T-Tavg Protection Channell Calibration, Rev. 0 07-00521, Containment Coatings Walkdown, Rev. 0 07-05146, 18-Month Slave Relay Testing (Train B), Rev. 0 08-02593, Elimination of Recirculation Spray Pump Response Time SR 3.3.2.9, Rev. 1 08-05721, Replace Valves 2RCS-633 & 2RCS-634 with Pipe Spool Piece, Rev. 0 09-03164, Quench Spray System Trouble, Rev. 0 09-04690, RHS Fill and Vent from RWST, Rev. 0 08-05359, BV2 EDG ASP Transfer Circuit Relay Change, Rev. 0 09-00048, Station Battery Jumper Installation and Removal, Rev. 0 09-00238, Calculation  
: 8700-DMC-1430  
: Revision, Rev. 0 09-00374, Incipient Fault Monitor Repair, Rev. 0 09-01089, 1
: OST-1.22 Procedure Revision, Rev. 0 10-00126, Recirculation Spray Pump Test, Rev. 0 10-01483, Residual Heat Removal System Train A Valve Exercise, Rev. 0 10-00326, 1
: OST-1.22 Procedure Revision, Rev. 0 10-00126, Recirculation Spray Pump Test, Rev. 0 10-01483, Residual Heat Removal System Train A Valve Exercise, Rev. 0 10-00326, 1
: MSP-9.05-1 Calibration Procedure Revision, Rev. 0 10-00342, Emergency Diesel Generator Monthly Test Procedure Change, Rev. 0 Attachment   
: MSP-9.05-1  
: 10-00473, 1CMP-60-CR-15-1M Procedure Revision, Rev. 0 10-00731, Procedure 1
: Calibration
: MSP-36.43-E Revision, Rev. 0 10-00687, Replacement of BV1 Vital Bus Alternate Source Regulating Transformers, Rev. 0 Modification Packages (* designates Modification and 10
===Procedure===
: Revision, Rev. 0 10-00342, Emergency Diesel Generator Monthly Test Procedure Change, Rev. 0 Attachment   
: 10-00473, 1CMP-60-CR-15-1M
===Procedure===
: Revision, Rev. 0 10-00731, Procedure
: MSP-36.43-E  
: Revision, Rev. 0 10-00687, Replacement of BV1 Vital Bus Alternate Source Regulating Transformers, Rev. 0 Modification Packages (* designates Modification and 10
: CFR 50.59 screen sample) *ECP 06-0227, Add RWST Level Interlock to RS Pump Start, Rev. 1
: CFR 50.59 screen sample) *ECP 06-0227, Add RWST Level Interlock to RS Pump Start, Rev. 1
: ECP 07-0076, TDAFW Pump Steam Supply SOVs -Benchboard Control Switch Replacements, Rev. 1 *ECP 07-0145, Modification of U1 OS lines, Rev. 0 *ECP 08-0095, Unit 1 Emergency Diesel Generator K1 Relay Replacement, Rev. 1 *ECP 08-0134, Reclosing Reactor Trip Breaker Under Test, Rev. 0
: ECP 07-0076, TDAFW Pump Steam Supply SOVs -Benchboard Control Switch Replacements, Rev. 1 *ECP 07-0145, Modification of U1 OS lines, Rev. 0 *ECP 08-0095, Unit 1 Emergency Diesel Generator  
: ECP 08-0504, Unit 2 Joint Owners Group Motor Operated Valve Periodic Verification (JOG MOV PV) Program Implementation -2R14
: K1 Relay Replacement, Rev. 1 *ECP 08-0134, Reclosing Reactor Trip Breaker Under Test, Rev. 0
: ECP 09-03277, Recirculation Spray Heat Exchanger Inputs for MAAP Containment Analysis, Rev. 0 *ECP 10-0010, Unit 1 Containment Sump Discharge Flow Transmitter Replacement, Rev. 1 10080-N-584, Revision to Maximum Differential Pressure Across the Category I Motor Operated Valves In the Unit 2 Safety Injection (SI) System, Rev. 0 Calculations & Analysis 8700-DMC-1651, Containment Coatings Walkdown, Rev. 0 8700-DMC-1683, Required Height of OS Loop Seals, Rev. 0 10080-N-661, Torque Calculations for 2SIS-MOV867A, 2SIS-MOV867B, 2SIS-MOV867C, 2SIS-MOV867D, Rev. 12 10080-N-662, Torque Calculations for 2SIS-MOV869A and 2SIS-MOV869B, Rev. 8 10080-N-824, Recirculation Spray Heat Exchanger Inputs for MAAP Containment Analysis, Rev. 5 10080-US(B)-239, Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project, Rev. 5 12241-NP(B)-X70D, Pipe Stress Calculation -Reactor Coolant By-Pass -2 Inch Pipe Loop 21 -Reactor Containment Building, Rev. 3 12241-NP(B)-X70E, Pipe Stress Calculation -Reactor Coolant By-Pass -2 Inch Pipe Loop 22-Reactor Containment Building, Rev. 3 Corrective Action Reports 06-06553 07-28237  
: ECP 08-0504, Unit 2 Joint Owners Group Motor Operated Valve Periodic Verification (JOG MOV PV) Program Implementation  
-2R14
: ECP 09-03277, Recirculation Spray Heat Exchanger Inputs for MAAP Containment Analysis, Rev. 0 *ECP 10-0010, Unit 1 Containment Sump Discharge Flow Transmitter Replacement, Rev. 1 10080-N-584, Revision to Maximum Differential Pressure Across the Category I Motor Operated Valves In the Unit 2 Safety Injection (SI) System, Rev. 0 Calculations  
& Analysis 8700-DMC-1651, Containment Coatings Walkdown, Rev. 0 8700-DMC-1683, Required Height of OS Loop Seals, Rev. 0 10080-N-661, Torque Calculations for 2SIS-MOV867A, 2SIS-MOV867B, 2SIS-MOV867C, 2SIS-MOV867D, Rev. 12 10080-N-662, Torque Calculations for 2SIS-MOV869A
and 2SIS-MOV869B, Rev. 8 10080-N-824, Recirculation Spray Heat Exchanger Inputs for MAAP Containment Analysis, Rev. 5 10080-US(B)-239, Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project, Rev. 5 12241-NP(B)-X70D, Pipe Stress Calculation  
-Reactor Coolant By-Pass -2 Inch Pipe Loop 21 -Reactor Containment Building, Rev. 3 12241-NP(B)-X70E, Pipe Stress Calculation  
-Reactor Coolant By-Pass -2 Inch Pipe Loop 22-Reactor Containment Building, Rev. 3 Corrective Action Reports 06-06553 07-28237  
: 07-28510  
: 07-28510  
: 08-38463  
: 08-38463  
: 08-39206 (* denotes NRC identified during this inspection) 08-40329 10-78035 09-57719 10-78627* 10-71378 10-78665* 10-77661 10-78705* 10-77667 10-78734* Attachment
: 08-39206 (* denotes NRC identified during this inspection)  
: 08-40329 10-78035 09-57719 10-78627*  
: 10-71378 10-78665*  
: 10-77661 10-78705*  
: 10-77667 10-78734*
===Drawings===
===Drawings===
: 8700-RE-0018F, Sht. 5, Wiring Diagram -120V AC Wiring Details, Rev. 8 8700-RE-0011 F, Sht. 2, Wiring Diagram -120 VAC DIST PNLS 7, 8, 9, 10, 11, 12, and 13, Rev. 36 8700-RE-21 BU, Elementary Diagram -Diesel Gen Engine Controls Sheet 2 of 4, Rev. 20 8700-RE-0021TZ, Elementary Diagram -Reactor Trip Switchgear 52/RTA, 52/RTB, 52/BYA, 52/BYB, Rev. 7 10080-E-6HT, Boron Injection Isolation Valves, Rev. 15 10080-E-11 LV, Train A Signal Isolators, Rev. 8 1 0080-RE-4AF, Sht. 1, Wiring Diagram Reactor Protection Rack Input Cabinet Train B, Rev. 12 10080-RE-9GK, Sht. 4, 480V MCC*2-E03 Aux Bldg, Rev. 25 1 0080-RM-0087 A, BV2 Flow Diagram Safety Injection Piping, Rev. 28 10080-RM-0087B, BV2 Flow Diagram Safety Injection Piping, Rev. 22 1 0080-RM-0406-001, Valve Oper. No Diagram Reactor Coolant System, Rev. 23 10080-RM-0413-002, Valve Diagram Quench Spray System, Rev. 19 12241-E-5DQ, Recirculation Spray Pump (2RSS*P21A), Rev. 13 Procedures 1
: 8700-RE-0018F, Sht. 5, Wiring Diagram -120V AC Wiring Details, Rev. 8 8700-RE-0011  
: F, Sht. 2, Wiring Diagram -120 VAC DIST PNLS 7, 8, 9, 10, 11, 12, and 13, Rev. 36 8700-RE-21  
: BU, Elementary Diagram -Diesel Gen Engine Controls Sheet 2 of 4, Rev. 20 8700-RE-0021TZ, Elementary Diagram -Reactor Trip Switchgear  
: 2/RTA, 52/RTB, 52/BYA, 52/BYB, Rev. 7 10080-E-6HT, Boron Injection Isolation Valves, Rev. 15 10080-E-11  
: LV, Train A Signal Isolators, Rev. 8 1 0080-RE-4AF, Sht. 1, Wiring Diagram Reactor Protection Rack Input Cabinet Train B, Rev. 12 10080-RE-9GK, Sht. 4, 480V MCC*2-E03  
: Aux Bldg, Rev. 25 1 0080-RM-0087  
: A, BV2 Flow Diagram Safety Injection Piping, Rev. 28 10080-RM-0087B, BV2 Flow Diagram Safety Injection Piping, Rev. 22 1 0080-RM-0406-001, Valve Oper. No Diagram Reactor Coolant System, Rev. 23 10080-RM-0413-002, Valve Diagram Quench Spray System, Rev. 19 12241-E-5DQ, Recirculation Spray Pump (2RSS*P21A), Rev. 13 Procedures
: MSP-1.04-1, Reactor Protection System Train A Test, Rev. 47 1
: MSP-1.04-1, Reactor Protection System Train A Test, Rev. 47 1
: MSP-9.05-1, Containment Sump Flow Measuring System Calibration, Rev. 9 1MSP-36A3-E, 1N 480 Volt Emergency Bus Degraded Voltage Relays 27-RN2100AB and 27-RN2100BC Test, Rev. 23 1
: MSP-9.05-1, Containment Sump Flow Measuring System Calibration, Rev. 9 1MSP-36A3-E, 1N 480 Volt Emergency Bus Degraded Voltage Relays 27-RN2100AB
and 27-RN2100BC  
: Test, Rev. 23 1
: OM-9.1. E, Specific Instrumentation and Controls, Rev. 4 10M-9.3.C, Power Supply and Control Switch List, Rev. 6 1
: OM-9.1. E, Specific Instrumentation and Controls, Rev. 4 10M-9.3.C, Power Supply and Control Switch List, Rev. 6 1
: OM-53A.1.E-0 (ISS1 C), Reactor Trip or Safety Injection, Rev. 11 10M-53A.1.ES-1.3 (ISS1C), Transfer to Cold Leg Recirculation, Rev. 7 1
: OM-53A.1.E-0 (ISS1 C), Reactor Trip or Safety Injection, Rev. 11 10M-53A.1.ES-1.3 (ISS1C), Transfer to Cold Leg Recirculation, Rev. 7 1
Line 155: Line 362:
: OM-37.4.X, 480V Buses 1 G and 1 H Maintenance Support, Rev. 3 1
: OM-37.4.X, 480V Buses 1 G and 1 H Maintenance Support, Rev. 3 1
: OM-37 A.Y, 480V Buses 1 E and 1 F Maintenance Support, Rev. 2 10M-38.5.B.2, Table 38-2 AC Distribution Panels Load List, Rev. 30 10M-45GA.AAA, Seismic Accelograph Operation, Rev. 5 1 OST -1.22, Manual ESF Actuation Circuitry Test, Rev. 5 10ST-13.7C, 1A Recirculation Spray Pump Auto Start Test, Rev. 6 10ST-36.2, Diesel Generator No.2 Monthly Test, Rev. 54 1
: OM-37 A.Y, 480V Buses 1 E and 1 F Maintenance Support, Rev. 2 10M-38.5.B.2, Table 38-2 AC Distribution Panels Load List, Rev. 30 10M-45GA.AAA, Seismic Accelograph Operation, Rev. 5 1 OST -1.22, Manual ESF Actuation Circuitry Test, Rev. 5 10ST-13.7C, 1A Recirculation Spray Pump Auto Start Test, Rev. 6 10ST-36.2, Diesel Generator No.2 Monthly Test, Rev. 54 1
: PMP-36EE-EG-1-2-1E, Emergency Diesel Generator Relay Cleaning and Inspection, Rev. 12 1/2-CMP-E-39-366, Station Battery Jumper Installation and Restoration, Rev. 7 2MSP-1.04-1, Reactor Protection System Train A Test, Rev. 44 2BVT 1.13.5, Recirculation Spray Pump Test, Rev. 15 2MSP-1.04-1, Reactor Protection System Train A Test, Rev. 44 2MSP-6.38-1, Reactor Coolant Temperature Loop 2RCS-T412 Delta T-Tavg Protection Channel I Calibration, Rev. 12 2MSP-13.05-1, 2QSS-L 104A, Refueling Water Storage Tank 2QSS-TK21 Level Loop Channell Calibration, Rev. 15 20M-6A.AH, Filling the RCS/RHS Piping with the Temporary Reactor Vessel Cover Installed, Rev. 0 Attachment   
: PMP-36EE-EG-1-2-1E, Emergency Diesel Generator Relay Cleaning and Inspection, Rev. 12 1/2-CMP-E-39-366, Station Battery Jumper Installation and Restoration, Rev. 7 2MSP-1.04-1, Reactor Protection System Train A Test, Rev. 44 2BVT 1.13.5, Recirculation Spray Pump Test, Rev. 15 2MSP-1.04-1, Reactor Protection System Train A Test, Rev. 44 2MSP-6.38-1, Reactor Coolant Temperature Loop 2RCS-T412  
: 20M-13.4.AAB, Quench Spray System Trouble, Rev. 8 20M-24.4.ABM, AFW Pump SOV SS Closed, Rev. 0 20M-53A.1.E-0 (ISS1 C), Reactor Trip or Safety Injection, Rev. 8 20M-53E.1.SAG-6, Control Containment Conditions, Rev. 3 20M-53A.1.E-3 (ISS1 C), Steam Generator Tube Rupture, Rev. 15 20M-53A.1.ES-1.3 (ISS1 C), Transfer to Cold Leg Recirculation, Rev. 6 20ST -1.10, Cold Shutdown Valve Exercise Test, Rev. 33 20ST-10.3, Residual Heat Removal System Train A Valve Exercise, Rev. 24 20ST-13.3, Recirculation Spray Pump [2RSS*P21A] Dry Test, Rev. 12 20ST-13.3A, Recirculation Spray Pump [2RSS*P21A] Automatic Start Circuit Test, Rev. 3 20ST-24.4, Steam Driven Auxiliary Feed Pump [2FWE*P22] Quarterly Test, Rev. 65 20ST-47.3M, Containment Penetration and ASME Valve Test, Rev. 19
: Delta T-Tavg Protection Channel I Calibration, Rev. 12 2MSP-13.05-1, 2QSS-L 104A, Refueling Water Storage Tank 2QSS-TK21  
: Level Loop Channell Calibration, Rev. 15 20M-6A.AH, Filling the RCS/RHS Piping with the Temporary Reactor Vessel Cover Installed, Rev. 0 Attachment   
: 20M-13.4.AAB, Quench Spray System Trouble, Rev. 8 20M-24.4.ABM, AFW Pump SOV SS Closed, Rev. 0 20M-53A.1.E-0 (ISS1 C), Reactor Trip or Safety Injection, Rev. 8 20M-53E.1.SAG-6, Control Containment Conditions, Rev. 3 20M-53A.1.E-3 (ISS1 C), Steam Generator Tube Rupture, Rev. 15 20M-53A.1.ES-1.3 (ISS1 C), Transfer to Cold Leg Recirculation, Rev. 6 20ST -1.10, Cold Shutdown Valve Exercise Test, Rev. 33 20ST-10.3, Residual Heat Removal System Train A Valve Exercise, Rev. 24 20ST-13.3, Recirculation Spray Pump [2RSS*P21A]  
: Dry Test, Rev. 12 20ST-13.3A, Recirculation Spray Pump [2RSS*P21A]  
: Automatic Start Circuit Test, Rev. 3 20ST-24.4, Steam Driven Auxiliary Feed Pump [2FWE*P22]  
: Quarterly Test, Rev. 65 20ST-47.3M, Containment Penetration and ASME Valve Test, Rev. 19
: NOP-CC-2004, Design Interface Reviews and Evaluations, Rev. 7 Work Orders
: NOP-CC-2004, Design Interface Reviews and Evaluations, Rev. 7 Work Orders
: 200249586
: 200249586
Line 167: Line 379:
: 200286520
: 200286520
: 200286513
: 200286513
: 200313752 Vendor Manuals 07.561-0023, Instructions and Parts Lists for Magnetic Flowmeter Instrumentation, Rev. F Miscellaneous 10BD-36A, Design Basis Document for Emergency Diesel Generators, Rev. 11 20BO-11, Design Basis Document for Safety Injection System, Rev. 12 1
: 200313752  
: Vendor Manuals 07.561-0023, Instructions and Parts Lists for Magnetic Flowmeter Instrumentation, Rev. F Miscellaneous  
: 10BD-36A, Design Basis Document for Emergency Diesel Generators, Rev. 11 20BO-11, Design Basis Document for Safety Injection System, Rev. 12 1
: SQS-36.2, Unit 1 Diesel Generator Operator Training Lesson Plan, Rev. 16 Licensed Operator Training Module 1
: SQS-36.2, Unit 1 Diesel Generator Operator Training Lesson Plan, Rev. 16 Licensed Operator Training Module 1
: SQS-13.1, Containment Depressurization System, Rev. 1 Licensed Operator Training Module 2SQS-24.1, Feedwater Systems, Rev. 1 Preparation for
: SQS-13.1, Containment Depressurization System, Rev. 1 Licensed Operator Training Module 2SQS-24.1, Feedwater Systems, Rev. 1 Preparation for
: NRC 10CFR50.59/Modification Inspection Self-Assessment, dated 06/04/10 Technical Specifications, Beaver Valley Power Station Unit 1, Amendment 280 Technical Specifications, Beaver Valley Power Station Unit 2, Amendment 164 Attachment   
: NRC 10CFR50.59/Modification Inspection Self-Assessment, dated 06/04/10 Technical Specifications, Beaver Valley Power Station Unit 1, Amendment  
: 280 Technical Specifications, Beaver Valley Power Station Unit 2, Amendment  
: 164 Attachment   
: ADAMS CFR CR DBA  
: ADAMS CFR CR DBA  
: DP DRS ECP EDG EPRI FENOC GL HX IP JOG LHSI MAAP MOV NEI NPSH OS RS RWST SGTR SI SOV TDAFW TS UFSAR A-5
: DP DRS ECP EDG EPRI FENOC GL HX IP JOG LHSI MAAP MOV NEI NPSH OS RS RWST SGTR SI SOV TDAFW TS UFSAR A-5
==LIST OF ACRONYMS==
==LIST OF ACRONYMS==
: [[NRC]] [[Document System Code of Federal Regulations Condition Report Design Basis Accident Differential Pressure Division of Reactor Safety Engineering Change Package Emergency Diesel Generator Electric Power Research Institute FirstEnergy Nuclear Operating Company Generic Letter Heat Exchanger Inspection Procedure Joint Owners Group Low Head Safety Injection Modular Accident Analysis Program Motor Operated Valve Nuclear Energy Institute Net Positive Suction Head Quench Spray Recirculation Spray Refueling Water Storage Tank Steam Generator Tube Rupture Safety Injection Solenoid Operated Valves Turbine Driven Auxiliary Feedwater Technical Specifications Updated Final Safety Analysis Report Attachment]]
NRC Document System Code of Federal Regulations
Condition
Report Design Basis Accident Differential
Pressure Division of Reactor Safety Engineering
Change Package Emergency
Diesel Generator
Electric Power Research Institute
FirstEnergy
Nuclear Operating
Company Generic Letter Heat Exchanger
Inspection
Procedure
Joint Owners Group Low Head Safety Injection
Modular Accident Analysis Program Motor Operated Valve Nuclear Energy Institute
Net Positive Suction Head Quench Spray Recirculation
Spray Refueling
Water Storage Tank Steam Generator
Tube Rupture Safety Injection
Solenoid Operated Valves Turbine Driven Auxiliary
Feedwater
Technical
Specifications
Updated Final Safety Analysis Report Attachment
}}
}}

Revision as of 19:56, 21 August 2018

IR 05000334-10-006, 05000412-10-006 on 6/07/2010 - 6/24/2010 for Beaver Valley Power Station, Units 1 & 2, Engineering Specialist Plant Modifications Inspection
ML102150492
Person / Time
Site: Beaver Valley
Issue date: 08/03/2010
From: Doerflein L T
Engineering Region 1 Branch 2
To: Harden P
FirstEnergy Nuclear Operating Co
References
IR-10-006
Download: ML102150492 (18)
v • d • e


Text

Mr. Paul Harden Site Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PA 19406-1415 August 3, 2010 FirstEnergy Nuclear Operating Company Beaver Valley Power Station P.O. Box 4, Route 168 Shippingport, PA 15077

SUBJECT: BEAVER VALLEY POWER STATION -NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000334/2010006 AND 05000412/2010006

Dear Mr. Harden:

On June 24,2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Beaver Valley Power Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on June 24, 2010, with Mr. Ray Lieb, Director Site Operations, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 Document Room or from the Publicly Available Records (PARS) component of the 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: 50-334, 50-412 License Nos: DPR-66, NPF-73

Sincerely,Lawrence T. Doerflein, Chi Engineering Branch 2 Division of Reactor Safety

Enclosure:

Inspection Report No. 05000334/2010006; 05000412/2010006

w/Attachment:

Supplemental Information cc w/encl: Distribution via ListServ ,

Mr. Paul Harden Site Vice President August 3, 2010 FirstEnergy Nuclear Operating Company Beaver Valley Power Station P.O. Box 4, Route 168 Shippingport, PA 15077

SUBJECT: BEAVER VALLEY POWER STATION -NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000334/2010006 AND 05000412/2010006

Dear Mr. Harden:

On June 24, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Beaver Valley Power Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on June 24, 2010, with Mr. Ray Lieb, Director Site Operations, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 Document Room or from the Publicly Available Records (PARS) component of the 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: 50-334, 50-412 License Nos: DPR-66, NPF-73

Sincerely,IRA! Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety

Enclosure:

Inspection Report No. 05000334/2010006; 05000412/2010006

w/Attachment:

Supplemental Information cc w/encl: Distribution via ListServ ADAMS ACCESSION NO. ML 102150492 SUNSI Review Complete:

Initials)

DOCUMENT NAME: Y:\Division\DRS\Engineering Branch 2\Arner\beaver valley mods\BVModsReport2010006.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: "ell C = Opy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE RI/DRS I RI/DRP I RI/DRS I I NAME FArnerl RBellamvl LDoerfieinl DATE 07/23/10 07/30/10 08/3110 OFFICIAL RECORD COpy Distribution w/encl: M. Dapas, Acting RA (R10RAMAIL Resource)

D. Lew, Acting DRA (R10RAMAIL Resource)

J. Clifford, DRP (R1DRPMAIL Resource)

D. Collins, DRP (R1DRPMAIL Resource)

D. Roberts, DRS (R1DRSMaii Resource)

P. Wilson, DRS (R1DRSMaii Resource)

R. Bellamy, DRP S. Barber, DRP C. Newport, DRP D. Werkheiser, SRI E. Bonney, RI P. Garrett, Resident OA L. Trocine, RI OEDO D. Bearde, DRS RidsNRRPMBeaverValley Resource RidsNrrDorlLpl1-1 Resource@nrc.gov ROPreportsResource@nrc.gov L. Doerflein, DRS F. Arner, DRS U.S. NUCLEAR REGULATORY COMMISSION REGION I Docket No.: 50-334, 50-412 License No.: DPR-66, NPF-73 Report No.: 05000334/2010006 and 05000412/2010006 Licensee:

FirstEnergy Nuclear Operating Company (FENOC) Facility:

Beaver Valley Power Station, Units 1 and 2 Location:

Post Office Box 4 Shippingport, PA 15077 Inspection Period: June 7 through June 24, 2010 Inspectors:

F. Arner, Senior Reactor Inspector, Division of Reactor Safety (DRS), Team Leader L. Scholl, Senior Reactor Inspector, DRS E. Burkett, Reactor Inspector, DRS A. Dugandzic, Reactor Engineer, DRS (in-training)

Approved By: Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000334/2010006;

IR 05000412/2010006; 6/07/2010

-6/24/2010;

FirstEnergy Nuclear Operating Company (FENOC); Beaver Valley Power Station, Units 1 & 2; 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 and one inspector in training.

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. NRG-Identified and Self-Revealing Findings No findings of significance were identified.

B. Licensee-Identified Violations

None. ii

REPORT DETAILS

REACTOR SAFETY

Cornerstones:

Initiating Events, Mitigating Systems, and Barrier Integrity 1 R 17 Evaluations of Changes. Tests. or Experiments and Permanent Plant Modifications (IP 71111.17)

.1 Evaluations

of Changes. Tests. or Experiments (28 samples)

a. Inspection Scope

The team reviewed four 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 FENOC 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-four 10 CFR 50.59 screenings for which FENOC had concluded that no safety evaluation was required.

These reviews were performed to assess whether FENOC'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 FENOC had performed and approved during the time period covered by this inspection (Le., since the last modifications inspection)not previously reviewed by NRC inspectors.

The screenings were selected based on the safety significance, risk significance, and complexity of the change to the facility.

In addition, the team compared FENOC'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. Findings

No findings of significance were identified.

2

.2 Permanent

Plant Modifications (9 samples)

.2.1 Unit 1 Emergency

Diesel Generator (EDG) K 1 Relay Replacement*

a. Inspection Scope

The team reviewed a modification (Engineering Change Package (ECP) 08-0095) that replaced the original latching style relays with an electrically equivalent relay without the latching feature. The K1 relay design function is to shunt the field circuit during the EDG shutdown sequence and thereby terminate the generator electrical output in a controlled manner. The original relay would remain latched closed until unlatched during the next EDG startup sequence.

One reason for the replacement was that the original style relays were found to be potentially subject to auxiliary contact terminals coming loose during a seismic event. Additionally, Unit 2 had experienced a problem with this latching type relay when it failed to release during an EDG start sequence and thereby rendered the EDG inoperable.

This modification also added indicating lights to allow operators to verify the K1 relay opened as designed at the end of the shutdown sequence indicating that the excitation system was properly aligned for the next EDG start. An indicator light was also added to allow monitoring of the K2 field flash relay in a similar manner. The team conducted the review to verify that the design bases, licensing bases, and performance capability of the EDGs were not degraded by the component replacements and circuit modifications.

The team discussed the modification and design basis with design and system engineers, and reviewed the scope and results of post-modification testing to assess the capability of the EDGs to perform their safety function during a design basis event. The team also confirmed that surveillance tests, operational procedures, and drawings had been appropriately updated to reflect the design change. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.2 Unit 1 Containment Sump Discharge Flow Transmitter Replacement

a. Inspection Scope

The team reviewed a modification (ECP-10-00010)that replaced an obsolete containment sump discharge flow instrumentation system with an equivalent model. The scope of the modification included the replacement of the flow tube and flow transmitter, and changed the location of its electrical power feed. The instrument provides a method of determining the amount of unidentified leakage from systems located in the containment building as required by plant technical specification 3.4.15, RCS Leakage Detection Instrumentation.

3 The team conducted the review to ensure that the design bases, licensing bases, and performance capability of the leakage monitoring system had not been adversely affected by the modification.

The team reviewed the associated work order packages, and conducted interviews with design and system engineers regarding the design, installation, calibration, and testing of the instrumentation to verify that the modification was adequate.

The team walked down the accessible portions of the new equipment to ensure the system configuration was in accordance with the design instructions.

The team also verified that drawings and operating procedures had been appropriately revised to reflect the change. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.3 Reclosing Reactor Trip Breaker Under Test

a. Inspection Scope

The team reviewed a modification (ECP-08-0134)performed by FENOC to install a local test switch in the reactor trip switchgear.

The test switch allows local closing of the individual reactor trip breakers during surveillance testing instead of both trip breakers (the breaker under test and the other in-service breaker) receiving a closing signal when the breaker is closed from the control room switch. As a result, the in-service breaker is not unnecessarily subjected to the reverse inductive voltage spike during testing. The team evaluated the change to confirm that the design bases, licensing bases, and performance capability of the reactor trip system would not be affected by the change. The team also walked down accessible portions of the system to assess the configuration and material condition of the system, and reviewed updated surveillance procedures to ensure they included appropriate directions for the technicians operating the switches during testing. The team interviewed design engineers to review the design change and the potential impact on proper system operation.

The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

4

.2.4 Refueling

Water Storage Tank (RWST) Level Interlock to Recirculation Spray (RS) Pump Start

a. Inspection Scope

The team reviewed a modification (ECP 06-0227) that revised the automatic start signal for the RS pumps at both units to be initiated from an RWST level input. FENOC implemented the modification to regain RS pump net positive suction head (NPSH) margin that was reduced following the installation of modified containment sump strainers, which had been installed to address potential debris accumulation.

Prior to this modification, the RS pumps started automatically following a containment high-high pressure signal and a time delay. The modification included adding main control room benchboard status lights and changing electrical wiring for the RS pump start circuit and relay logic. The team conducted the review to verify that the design bases, licensing bases and performance capability of the recirculation spray pumps had not been adversely affected by the modification.

The team reviewed drawings, calculations, operator training modules, and procedures to ensure they had been properly updated to incorporate the changes to the RS pump start logic. In addition, post-modification testing was reviewed to verify the logic sequence and automatic RS pump start were appropriately tested. The team walked down portions of the modification to verify that status lights and annunciators were added to the main control room benchboards.

The team also discussed the modification and design basis with design engineers to assess the adequacy of the modification.

The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R 17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of Significance were identified . . 2.5 Recirculation Spray Heat Exchanger (HX) Inputs for Modular Accident Analysis Program (MAAP) Containment Analysis

a. Inspection Scope

The team reviewed an analysis (ECP 09-03277)performed by FENOC to evaluate the fouling factor of the Unit 2 RS heat exchangers and the potential for air entrapment within the heat exchangers.

This analysis was performed to ensure the design fouling factor assumption for the two RS heat exchangers cooled by service water was appropriate.

FENOC performed the analysis using the Electric Power Research Institute (EPRI) Modular Accident Analysis Program (MAAP) to demonstrate the fouling factor was acceptable for component performance during design basis accident conditions.

Additionally, FENOC evaluated the internal shell side velocities of the heat exchangers to determine if they were sufficient to sweep air from the system during Enclosure 5 initial fill. Using a Froude number criterion, FENOC determined that air would be swept from the system upon initial fill during system initiation, and not impede HX performance.

The team reviewed the analysis and associated calculations to verify the assumptions and parameters used were valid and considered worst case scenarios.

The parameters included service water temperature, containment temperature and pressure, number of plugged tubes, sump temperature, and heat exchanger baffle arrangement.

The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.6 Turbine Driven Auxiliary Feedwater (TDAFW) Pump Steam Supply Solenoid Operated Valves (SOVs) -Benchboard Control Switch Replacements

a. Inspection Scope

The team reviewed a modification (ECP 07-0076) that replaced the selector switches for the TDAFW pump steam admission solenoid operated valves in the Unit 2 main control room. FENOC implemented this modification to address a postulated accident scenario such as a steam generator tube rupture (SGTR) event which would require plant operators to shut down the TDAFW pump from the control room using class 1 E safety related controls.

Prior to the modification, the selector switches had" return-to-center" operators that prevented the steam admission valves from maintaining a closed position which may be necessary during certain design basis events. The replacement switches were designed to maintain the position in which they have been placed. The modification included installing an annunciator in the main control room, and replacing and rewiring the control switches.

The team reviewed operating procedures, operator training modules, and drawings to ensure they had been properly updated. The team reviewed work order packages and conducted interviews with design and system engineers regarding the design, installation, and testing to verify that the modification was adequate.

The team performed a walkdown of the modification in the main control room and discussed the alarm response procedure with the operators.

The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

6

.2.7 Unit 2 Joint Owners Group (JOG) Motor Operated Valve (MOV) Periodic Verification

Program Implementation

a. Inspection Scope

The team reviewed a modification (ECP 08-0504) that changed the setup for motor operated valve (MOV) actuators associated with the Unit 2 safety injection (SI) system. FENOC implemented this modification to incorporate requirements of the JOG MOV program which is an industry-wide response to Generic Letter (GL) 1996-05, " Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves." The team selected a sample of two valves to review the changes performed.

The valves reviewed were 2SIS-MOV867 A; high head SI valve, and 2SIS-MOV869A, hot leg SI isolation valve. The actuator control switch setup was modified to account for increased torque and thrust requirements, and reduced stress relative to the weak link components.

Additionally, the control switch for the closure function for valve 2SIS-MOV867 A was changed from torque switch control to limit switch control. The team conducted the review to verify that the design bases and performance capability of the safety injection system had not been adversely affected by the modification.

The team reviewed associated drawings and calculations to ensure they were properly updated and the correct assumptions were used. In addition, the team reviewed completed work orders to verify the appropriateness of the testing and acceptance criteria.

The team also discussed the modification and design bases with design and system engineers to assess the adequacy of the modification.

The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.8 Unit 1 Quench Spray Loop Seal Modification

a. Inspection Scope

The team reviewed a modification (ECP 07-0145) associated with the installation of a plug in the vent hole on the two quench spray (as) system loop seals and the repositioning of drain valves in the QS piping from open to normally closed. The modification was implemented in order to prevent air ingestion into the recirculation spray (RS) and low head safety injection (LHSI) pumps through the containment sump plenums. This was required due to the revised containment sump design, which had created the potential for air ingestion from the as system. Specifically, several of the as branch lines penetrate and discharge into the containment sump plenum. The design intent of those lines was to increase NPSH margin by adding the cooler QS water initially to the containment-sump.

This modification also identified that the QS Enclosure 7 piping loop seal to the containment sump would be required to be kept full to prevent air ingestion during a postulated accident which would require containment sump recirculation.

The team reviewed the associated calculations to ensure the methods used were reasonable in determining that the loop seal would be able to prevent air ingestion during postulated accident scenarios.

Specifically, the team reviewed the design input assumption for containment pressure to ensure the calculated height of the loop seal would be adequate at the time the as pumps are secured after RWST depletion.

The team reviewed emergency operating procedures to verify that the containment pressure design input was reasonable.

Additionally, the team reviewed operating procedures established to ensure that the as loop seal would be adequately filled prior to operation as well as FENOC' s assessment for potential leakage impacting the design conclusions.

The team also reviewed the response time change associated with the as system because of the change regarding the loop seal being maintained water solid prior to operation.

This review was performed to ensure that surveillance tests and calculations had been revised appropriately.

The team interviewed design engineers to review the design change and the potential impact on proper system operation.

The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.9 Revision to Maximum Differential Pressure Across the Category I Motor Operated Valves in the Unit 2 Safety Injection System

a. Inspection Scope

The team reviewed a revision to design calculation 10080-N-584 that changed the differential pressure (DP) and line pressure requirements for SI valves due to the replacement of the Unit 2 charging pump rotating assemblies.

The team reviewed FENOC' s methodology used in determining the revised DP parameters for various valves within the system. The team reviewed selected design inputs used in the revised calculation to verify they were reasonable.

The design inputs verified included the static pressure affect from the refueling water storage tank, the containment pressure assumption at the time of swapover to sump recirculation, and the static elevation differences between the pumps and valves. The charging pump performance curve design input was also verified to have been correctly incorporated within the calculations with respect to the change of performance due to the replaced rotating assembly.

The emergency operating procedures were reviewed to verify equipment alignments assumed in the determination of valve differential pressures were in accordance with procedure direction.

8 The team sampled a few of the revised design DP outputs to ensure that the information was appropriately translated into torque calculations for Unit 2 SI valve 2SIS-MOV8811A, which automatically opens to provide a recirculation path from the containment sump via the recirculation pumps, during the recirculation phase of-safety injection.

Additionally, the team verified that the revised DP output was appropriately translated to the torque calculation for 2SIS-MOV863B, which automatically opens during the recirculation phase to divert the flow of the recirculation spray pumps to the suction of the high head safety injection (HHSI)/charging pumps. The team performed a walkdown of the SI inlet isolation valve, to assess the material condition of the valve. The team also performed interviews with the responsible motor operated valve engineers to determine whether the design changes appeared reasonable.

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)

a. Inspection Scope

The team reviewed a sample of the condition reports (CRs) associated with 10 CFR 50.59 and plant modification issues to determine whether FENOC 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 CRs written on issues identified during the inspection to verify adequate problem identification and incorporation of the problem into the corrective action system. The CRs reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

40A6 Meetings, including Exit The team presented the inspection results to Mr. Ray Lieb, Director, Site Operations, and other members of FENOC's staff at an exit meeting on June 24, 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 ATTACHMENT

=SUPPLEMENTAL

INFORMATION=

KEY POINTS OF CONTACT

FENOC Personnel

R. Fedin, Regulatory

Compliance

C. Mancuso, Design Engineering

Manager

D. Price, Design Engineering

Supervisor

K. Woessner, Nuclear Staff Engineer

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

None

LIST OF DOCUMENTS

REVIEWED 10

CFR 50.59 Evaluations
07-00157, Impact of Reduced Atmospheric Relief Dump Valve (ASDV) Capacity, Rev. 0 07 -05204, Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project, Rev. 0 08-05643, BV2 Primary Zinc Addition, Rev. 0 09-02468, BV2 UFSAR Change -Containment Sump Screen Passive Failure, Rev. 0 10
CFR 50.59 Screened-out Evaluations
07-00275, Reactor Coolant Temperature Loop 2RCS-T412
Delta T-Tavg Protection Channell Calibration, Rev. 0 07-00521, Containment Coatings Walkdown, Rev. 0 07-05146, 18-Month Slave Relay Testing (Train B), Rev. 0 08-02593, Elimination of Recirculation Spray Pump Response Time SR 3.3.2.9, Rev. 1 08-05721, Replace Valves 2RCS-633 & 2RCS-634 with Pipe Spool Piece, Rev. 0 09-03164, Quench Spray System Trouble, Rev. 0 09-04690, RHS Fill and Vent from RWST, Rev. 0 08-05359, BV2 EDG ASP Transfer Circuit Relay Change, Rev. 0 09-00048, Station Battery Jumper Installation and Removal, Rev. 0 09-00238, Calculation
8700-DMC-1430
Revision, Rev. 0 09-00374, Incipient Fault Monitor Repair, Rev. 0 09-01089, 1
OST-1.22 Procedure Revision, Rev. 0 10-00126, Recirculation Spray Pump Test, Rev. 0 10-01483, Residual Heat Removal System Train A Valve Exercise, Rev. 0 10-00326, 1
MSP-9.05-1
Calibration

Procedure

Revision, Rev. 0 10-00342, Emergency Diesel Generator Monthly Test Procedure Change, Rev. 0 Attachment
10-00473, 1CMP-60-CR-15-1M

Procedure

Revision, Rev. 0 10-00731, Procedure
MSP-36.43-E
Revision, Rev. 0 10-00687, Replacement of BV1 Vital Bus Alternate Source Regulating Transformers, Rev. 0 Modification Packages (* designates Modification and 10
CFR 50.59 screen sample) *ECP 06-0227, Add RWST Level Interlock to RS Pump Start, Rev. 1
ECP 07-0076, TDAFW Pump Steam Supply SOVs -Benchboard Control Switch Replacements, Rev. 1 *ECP 07-0145, Modification of U1 OS lines, Rev. 0 *ECP 08-0095, Unit 1 Emergency Diesel Generator
K1 Relay Replacement, Rev. 1 *ECP 08-0134, Reclosing Reactor Trip Breaker Under Test, Rev. 0
ECP 08-0504, Unit 2 Joint Owners Group Motor Operated Valve Periodic Verification (JOG MOV PV) Program Implementation

-2R14

ECP 09-03277, Recirculation Spray Heat Exchanger Inputs for MAAP Containment Analysis, Rev. 0 *ECP 10-0010, Unit 1 Containment Sump Discharge Flow Transmitter Replacement, Rev. 1 10080-N-584, Revision to Maximum Differential Pressure Across the Category I Motor Operated Valves In the Unit 2 Safety Injection (SI) System, Rev. 0 Calculations

& Analysis 8700-DMC-1651, Containment Coatings Walkdown, Rev. 0 8700-DMC-1683, Required Height of OS Loop Seals, Rev. 0 10080-N-661, Torque Calculations for 2SIS-MOV867A, 2SIS-MOV867B, 2SIS-MOV867C, 2SIS-MOV867D, Rev. 12 10080-N-662, Torque Calculations for 2SIS-MOV869A

and 2SIS-MOV869B, Rev. 8 10080-N-824, Recirculation Spray Heat Exchanger Inputs for MAAP Containment Analysis, Rev. 5 10080-US(B)-239, Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project, Rev. 5 12241-NP(B)-X70D, Pipe Stress Calculation

-Reactor Coolant By-Pass -2 Inch Pipe Loop 21 -Reactor Containment Building, Rev. 3 12241-NP(B)-X70E, Pipe Stress Calculation

-Reactor Coolant By-Pass -2 Inch Pipe Loop 22-Reactor Containment Building, Rev. 3 Corrective Action Reports 06-06553 07-28237

07-28510
08-38463
08-39206 (* denotes NRC identified during this inspection)
08-40329 10-78035 09-57719 10-78627*
10-71378 10-78665*
10-77661 10-78705*
10-77667 10-78734*

Drawings

8700-RE-0018F, Sht. 5, Wiring Diagram -120V AC Wiring Details, Rev. 8 8700-RE-0011
F, Sht. 2, Wiring Diagram -120 VAC DIST PNLS 7, 8, 9, 10, 11, 12, and 13, Rev. 36 8700-RE-21
BU, Elementary Diagram -Diesel Gen Engine Controls Sheet 2 of 4, Rev. 20 8700-RE-0021TZ, Elementary Diagram -Reactor Trip Switchgear
2/RTA, 52/RTB, 52/BYA, 52/BYB, Rev. 7 10080-E-6HT, Boron Injection Isolation Valves, Rev. 15 10080-E-11
LV, Train A Signal Isolators, Rev. 8 1 0080-RE-4AF, Sht. 1, Wiring Diagram Reactor Protection Rack Input Cabinet Train B, Rev. 12 10080-RE-9GK, Sht. 4, 480V MCC*2-E03
Aux Bldg, Rev. 25 1 0080-RM-0087
A, BV2 Flow Diagram Safety Injection Piping, Rev. 28 10080-RM-0087B, BV2 Flow Diagram Safety Injection Piping, Rev. 22 1 0080-RM-0406-001, Valve Oper. No Diagram Reactor Coolant System, Rev. 23 10080-RM-0413-002, Valve Diagram Quench Spray System, Rev. 19 12241-E-5DQ, Recirculation Spray Pump (2RSS*P21A), Rev. 13 Procedures
MSP-1.04-1, Reactor Protection System Train A Test, Rev. 47 1
MSP-9.05-1, Containment Sump Flow Measuring System Calibration, Rev. 9 1MSP-36A3-E, 1N 480 Volt Emergency Bus Degraded Voltage Relays 27-RN2100AB

and 27-RN2100BC

Test, Rev. 23 1
OM-9.1. E, Specific Instrumentation and Controls, Rev. 4 10M-9.3.C, Power Supply and Control Switch List, Rev. 6 1
OM-53A.1.E-0 (ISS1 C), Reactor Trip or Safety Injection, Rev. 11 10M-53A.1.ES-1.3 (ISS1C), Transfer to Cold Leg Recirculation, Rev. 7 1
OM-53E.1.SAG-3, Inject Into the RCS, Rev. 3 10M-36A.AH, Diesel Generator No.2 Start-Up and Shutdown, Rev. 11 1
OM-37.4.X, 480V Buses 1 G and 1 H Maintenance Support, Rev. 3 1
OM-37 A.Y, 480V Buses 1 E and 1 F Maintenance Support, Rev. 2 10M-38.5.B.2, Table 38-2 AC Distribution Panels Load List, Rev. 30 10M-45GA.AAA, Seismic Accelograph Operation, Rev. 5 1 OST -1.22, Manual ESF Actuation Circuitry Test, Rev. 5 10ST-13.7C, 1A Recirculation Spray Pump Auto Start Test, Rev. 6 10ST-36.2, Diesel Generator No.2 Monthly Test, Rev. 54 1
PMP-36EE-EG-1-2-1E, Emergency Diesel Generator Relay Cleaning and Inspection, Rev. 12 1/2-CMP-E-39-366, Station Battery Jumper Installation and Restoration, Rev. 7 2MSP-1.04-1, Reactor Protection System Train A Test, Rev. 44 2BVT 1.13.5, Recirculation Spray Pump Test, Rev. 15 2MSP-1.04-1, Reactor Protection System Train A Test, Rev. 44 2MSP-6.38-1, Reactor Coolant Temperature Loop 2RCS-T412
Delta T-Tavg Protection Channel I Calibration, Rev. 12 2MSP-13.05-1, 2QSS-L 104A, Refueling Water Storage Tank 2QSS-TK21
Level Loop Channell Calibration, Rev. 15 20M-6A.AH, Filling the RCS/RHS Piping with the Temporary Reactor Vessel Cover Installed, Rev. 0 Attachment
20M-13.4.AAB, Quench Spray System Trouble, Rev. 8 20M-24.4.ABM, AFW Pump SOV SS Closed, Rev. 0 20M-53A.1.E-0 (ISS1 C), Reactor Trip or Safety Injection, Rev. 8 20M-53E.1.SAG-6, Control Containment Conditions, Rev. 3 20M-53A.1.E-3 (ISS1 C), Steam Generator Tube Rupture, Rev. 15 20M-53A.1.ES-1.3 (ISS1 C), Transfer to Cold Leg Recirculation, Rev. 6 20ST -1.10, Cold Shutdown Valve Exercise Test, Rev. 33 20ST-10.3, Residual Heat Removal System Train A Valve Exercise, Rev. 24 20ST-13.3, Recirculation Spray Pump [2RSS*P21A]
Dry Test, Rev. 12 20ST-13.3A, Recirculation Spray Pump [2RSS*P21A]
Automatic Start Circuit Test, Rev. 3 20ST-24.4, Steam Driven Auxiliary Feed Pump [2FWE*P22]
Quarterly Test, Rev. 65 20ST-47.3M, Containment Penetration and ASME Valve Test, Rev. 19
NOP-CC-2004, Design Interface Reviews and Evaluations, Rev. 7 Work Orders
200249586
200252480
200253464
200253465
200338908
200338912
200284820
200286520
200286513
200313752
Vendor Manuals 07.561-0023, Instructions and Parts Lists for Magnetic Flowmeter Instrumentation, Rev. F Miscellaneous
10BD-36A, Design Basis Document for Emergency Diesel Generators, Rev. 11 20BO-11, Design Basis Document for Safety Injection System, Rev. 12 1
SQS-36.2, Unit 1 Diesel Generator Operator Training Lesson Plan, Rev. 16 Licensed Operator Training Module 1
SQS-13.1, Containment Depressurization System, Rev. 1 Licensed Operator Training Module 2SQS-24.1, Feedwater Systems, Rev. 1 Preparation for
NRC 10CFR50.59/Modification Inspection Self-Assessment, dated 06/04/10 Technical Specifications, Beaver Valley Power Station Unit 1, Amendment
280 Technical Specifications, Beaver Valley Power Station Unit 2, Amendment
164 Attachment
ADAMS CFR CR DBA
DP DRS ECP EDG EPRI FENOC GL HX IP JOG LHSI MAAP MOV NEI NPSH OS RS RWST SGTR SI SOV TDAFW TS UFSAR A-5

LIST OF ACRONYMS

NRC Document System Code of Federal Regulations

Condition

Report Design Basis Accident Differential

Pressure Division of Reactor Safety Engineering

Change Package Emergency

Diesel Generator

Electric Power Research Institute

FirstEnergy

Nuclear Operating

Company Generic Letter Heat Exchanger

Inspection

Procedure

Joint Owners Group Low Head Safety Injection

Modular Accident Analysis Program Motor Operated Valve Nuclear Energy Institute

Net Positive Suction Head Quench Spray Recirculation

Spray Refueling

Water Storage Tank Steam Generator

Tube Rupture Safety Injection

Solenoid Operated Valves Turbine Driven Auxiliary

Feedwater

Technical

Specifications

Updated Final Safety Analysis Report Attachment

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