Text

2017 Q1-Q4 ROP Inspection Findings
	ML20311A364
Person / Time
Site:	Surry
Issue date:	11/06/2017
From:	; Office of Nuclear Reactor Regulation
To:	;
References
	Download: ML20311A364 (316)
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1Q/2000 Inspection Findings - Surry 1 Page 1 of 5 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparrent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B

1Q/2000 Inspection Findings - Surry 1 Page 2 of 5 modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected.

Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

1Q/2000 Inspection Findings - Surry 1 Page 3 of 5 Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis

1Q/2000 Inspection Findings - Surry 1 Page 4 of 5 frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred.

This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April

1Q/2000 Inspection Findings - Surry 1 Page 5 of 5 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Miscellaneous Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate.

The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Last modified : April 01, 2002

2Q/2000 Inspection Findings - Surry 1 Page 1 of 5 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

2Q/2000 Inspection Findings - Surry 1 Page 2 of 5 Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparrent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected.

Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I

2Q/2000 Inspection Findings - Surry 1 Page 3 of 5 A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to

2Q/2000 Inspection Findings - Surry 1 Page 4 of 5 analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred.

This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the

2Q/2000 Inspection Findings - Surry 1 Page 5 of 5 willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Miscellaneous Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate.

The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Last modified : April 01, 2002

3Q/2000 Inspection Findings - Surry 1 Page 1 of 5 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

3Q/2000 Inspection Findings - Surry 1 Page 2 of 5 Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparrent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected.

Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I

3Q/2000 Inspection Findings - Surry 1 Page 3 of 5 A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to

3Q/2000 Inspection Findings - Surry 1 Page 4 of 5 analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred.

This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the

3Q/2000 Inspection Findings - Surry 1 Page 5 of 5 willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Miscellaneous Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate.

The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Last modified : March 29, 2002

4Q/2000 Inspection Findings - Surry 1 Page 1 of 5 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

4Q/2000 Inspection Findings - Surry 1 Page 2 of 5 Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparrent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected.

Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001

4Q/2000 Inspection Findings - Surry 1 Page 3 of 5 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability

4Q/2000 Inspection Findings - Surry 1 Page 4 of 5 requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred.

This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is

4Q/2000 Inspection Findings - Surry 1 Page 5 of 5 low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Miscellaneous Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate.

The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Last modified : March 28, 2002

1Q/2001 Inspection Findings - Surry 1 Page 1 of 5 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

1Q/2001 Inspection Findings - Surry 1 Page 2 of 5 Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparrent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected.

Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001

1Q/2001 Inspection Findings - Surry 1 Page 3 of 5 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred.

1Q/2001 Inspection Findings - Surry 1 Page 4 of 5 This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Physical Protection Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is

1Q/2001 Inspection Findings - Surry 1 Page 5 of 5 low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Miscellaneous Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate.

The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Last modified : March 28, 2002

2Q/2001 Inspection Findings - Surry 1 Page 1 of 5 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed

2Q/2001 Inspection Findings - Surry 1 Page 2 of 5 procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparrent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected.

Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

2Q/2001 Inspection Findings - Surry 1 Page 3 of 5 Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee

2Q/2001 Inspection Findings - Surry 1 Page 4 of 5 failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred.

This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Physical Protection Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April

2Q/2001 Inspection Findings - Surry 1 Page 5 of 5 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Miscellaneous Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate.

The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Last modified : March 27, 2002

3Q/2001 Inspection Findings - Surry 1 Page 1 of 5 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparrent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected.

Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001 Identified By: NRC Item Type: AV Apparent Violation

3Q/2001 Inspection Findings - Surry 1 Page 2 of 5 Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

3Q/2001 Inspection Findings - Surry 1 Page 3 of 5 Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis

3Q/2001 Inspection Findings - Surry 1 Page 4 of 5 frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred.

This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April

3Q/2001 Inspection Findings - Surry 1 Page 5 of 5 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Miscellaneous Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate.

The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Last modified : March 26, 2002

4Q/2001 Inspection Findings - Surry 1 Page 1 of 5 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparrent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected.

Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not

4Q/2001 Inspection Findings - Surry 1 Page 2 of 5 brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

4Q/2001 Inspection Findings - Surry 1 Page 3 of 5 Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred.

This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

4Q/2001 Inspection Findings - Surry 1 Page 4 of 5 Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Miscellaneous Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate.

The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding

4Q/2001 Inspection Findings - Surry 1 Page 5 of 5 THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Last modified : March 01, 2002

1Q/2002 Inspection Findings - Surry 1 Page 1 of 6 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided.

Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision

12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant

1Q/2002 Inspection Findings - Surry 1 Page 2 of 6 system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1.

Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected. Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3. A supplemental inspection of this issue, documented in report 50-280, 281/02-08, determined that the licensee had performed an overall adequate evaluation of performance deficiencies related to the failure, and that the corrective actions were appropriately prioritized and consistent with the identified root cause and contributing factors and provided reasonable assurance to prevent recurrence.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998. Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown. This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: N/A Apr 12, 2002 Identified By: NRC Item Type: FIN Finding Supplemental Inspection Results For Unit 1 and 2 White Finding, Emergency Diesel Generator Bearing Failures, and Unit 2 White Performance Indicator, Safety System Unavailability - Emergency AC Power This supplemental inspection was performed by the NRC to assess the licensee's evaluation and corrective actions associated with a low to moderate risk significance (White) finding applicable to Units 1 and 2 and a Unit 2 White performance indicator (PI). The White finding and PI are in the mitigating systems cornerstone in the reactor safety strategic performance area. The White finding is described in NRC Final Significance Determination letter dated December 21, 2001, and was associated with the Emergency Diesel Generator (EDG) 3 wrist pin failures. The PI, Safety System Unavailability - Emergency AC Power, crossed the White threshold in the third quarter of calendar year 2001 and remained there through the current quarter. The White PI resulted mainly from the EDG 3 piston wrist pin failures and from EDG 2 output breaker problems. During this supplemental inspection, which was performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee performed an overall adequate evaluation of performance deficiencies related to the EDG 3 piston wrist pin failure and the EDG 2 output breaker problems. The depth of root cause evaluations was adequate. The corrective actions were appropriately prioritized and consistent with the identified root cause and contributing factors and provided reasonable assurance to prevent recurrence.

Inspection Report# : 2002008(pdf)

Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation

1Q/2002 Inspection Findings - Surry 1 Page 3 of 6 NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power.

This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously.

This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire. This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02], 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position.

This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

1Q/2002 Inspection Findings - Surry 1 Page 4 of 6 Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry),"

specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred. This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7

1Q/2002 Inspection Findings - Surry 1 Page 5 of 6 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP).

Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP).

Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024. Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Miscellaneous Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate. The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate. Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

1Q/2002 Inspection Findings - Surry 1 Page 6 of 6 Last modified : July 22, 2002

2Q/2002 Inspection Findings - Surry 1 Page 1 of 8 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Jun 29, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide appropriate instructions to ensure proper operation of the emergency switchgear room chillers following a loss of instrument air A Non-Cited Violation of Technical Specification 6.4.A was identified due to an inadequate abnormal procedure.

Abnormal procedure (AP) - 40, "Non-Recoverable Loss of Instrument Air," did not contain adequate guidance to ensure continued operation of the emergency switchgear room chillers following a loss of instrument air. The finding was of very low safety significance due to the combination of events that would have to occur for the emergency switchgear room components to be adversely affected by the loss of the chillers. The combination of events included a medium or large break loss of coolant accident coupled with a loss of offsite power during the winter months (cold service water temperatures).

Inspection Report# : 2002002(pdf)

Significance: N/A Apr 12, 2002 Identified By: NRC Item Type: FIN Finding Supplemental Inspection Results For Unit 1 and 2 White Finding, Emergency Diesel Generator Bearing Failures, and Unit 2 White Performance Indicator, Safety System Unavailability - Emergency AC Power This supplemental inspection was performed by the NRC to assess the licensee's evaluation and corrective actions associated with a low to moderate risk significance (White) finding applicable to Units 1 and 2 and a Unit 2 White performance indicator (PI). The White finding and PI are in the mitigating systems cornerstone in the reactor safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Surry 1 Page 2 of 8 strategic performance area. The White finding is described in NRC Final Significance Determination letter dated December 21, 2001, and was associated with the Emergency Diesel Generator (EDG) 3 wrist pin failures. The PI, Safety System Unavailability - Emergency AC Power, crossed the White threshold in the third quarter of calendar year 2001 and remained there through the current quarter. The White PI resulted mainly from the EDG 3 piston wrist pin failures and from EDG 2 output breaker problems. During this supplemental inspection, which was performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee performed an overall adequate evaluation of performance deficiencies related to the EDG 3 piston wrist pin failure and the EDG 2 output breaker problems. The depth of root cause evaluations was adequate. The corrective actions were appropriately prioritized and consistent with the identified root cause and contributing factors and provided reasonable assurance to prevent recurrence.

Inspection Report# : 2002008(pdf)

Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC informed the licensee of its final significance determination for the proposed apparent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Surry 1 Page 3 of 8 corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected. Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3. A supplemental inspection of this issue, documented in report 50-280, 281/02-08, determined that the licensee had performed an overall adequate evaluation of performance deficiencies related to the failure, and that the corrective actions were appropriately prioritized and consistent with the identified root cause and contributing factors and provided reasonable assurance to prevent recurrence.

Inspection Report# : 2001006(pdf)

Significance: TBD Sep 25, 2001 Identified By: NRC Item Type: AV Apparent Violation Number 3 emergency diesel generator inoperable for 6 days longer than the 7 days allowed by Technical Specification 3.16.B.1.a.3.

See apparent violation 05000280,05000281/2001006-01. This apparent violation was combined into Violation 05000280,05000281/2001006-01 when the NRC issued its final significance determination by letter dated December 21, 2001. All follow up activities associated with the issue will be performed as part of the follow up for the issued violation. For administrative tracking purposes, this item is consider closed. See Violation 05000280,05000281/2001006-01.

Inspection Report# : 2001006(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998.

Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown.

This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: Dec 16, 2000 file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Surry 1 Page 4 of 8 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

Inspection Report# : 2000005(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire.

This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Surry 1 Page 5 of 8 S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02],

2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Surry 1 Page 6 of 8 Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred. This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Surry 1 Page 7 of 8 individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024.

Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Miscellaneous Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate. The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate.

Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Surry 1 Page 8 of 8 evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/03/2003

3Q/2002 Inspection Findings - Surry 1 Page 1 of 7 Surry 1 Initiating Events Significance: Aug 11, 2000 Identified By: NRC Item Type: FIN Finding TWO OF THE THREE CABLES THAT CONNECTS OFFSITE POWER TO THE RESERVE STATION TRANSFORMERS HAVE LESS INSULATION THAN SPECIFIED IN INDUSTRY STANDARDS.

The team identified that two of the three cables that connects offsite power to the Reserve Station Transformers (RSSTs) had less insulation than specified in industry standards. A special analysis was performed by an NRC Region II Senior Reactor Analyst (SRA) to determine the effect on risk of the two RSST feeder cables having less than standard insulation thickness based on the 17 years that the cables had been in service. The risk screening analysis performed for the postulated cable failure indicated that there would be a slight increase in the Loss of Offsite Power (LOOP) initiation frequency resulting in a change in the Core Damage Frequency (CDF) of less than 1.0x10-6. The SRA's review concluded that the change in LOOP initiation frequency and the resultant change in CDF was GREEN.

Inspection Report# : 2000007(pdf)

Mitigating Systems Significance: Jun 29, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide appropriate instructions to ensure proper operation of the emergency switchgear room chillers following a loss of instrument air A Non-Cited Violation of Technical Specification 6.4.A was identified due to an inadequate abnormal procedure.

Abnormal procedure (AP) - 40, "Non-Recoverable Loss of Instrument Air," did not contain adequate guidance to ensure continued operation of the emergency switchgear room chillers following a loss of instrument air. The finding was of very low safety significance due to the combination of events that would have to occur for the emergency switchgear room components to be adversely affected by the loss of the chillers. The combination of events included a medium or large break loss of coolant accident coupled with a loss of offsite power during the winter months (cold service water temperatures).

Inspection Report# : 2002002(pdf)

Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Determine the risk significance of the failure to provide proper separation between the 125V DC busses A finding was identified for not providing proper separation between the 125V DC busses. A single failure could affect both redundant DC busses on a unit and encumber normal decay heat removal systems. The finding was of very low safety significance due to plant design features which mitigate the consequences of a fault within the DC system.

Specifically, there are numerous alternative methods of decay heat removal available with simple operator actions.

Inspection Report# : 2002002(pdf)

Inspection Report# : 2002003(pdf)

3Q/2002 Inspection Findings - Surry 1 Page 2 of 7 Significance: N/A Apr 12, 2002 Identified By: NRC Item Type: FIN Finding Supplemental Inspection Results For Unit 1 and 2 White Finding, Emergency Diesel Generator Bearing Failures, and Unit 2 White Performance Indicator, Safety System Unavailability - Emergency AC Power This supplemental inspection was performed by the NRC to assess the licensee's evaluation and corrective actions associated with a low to moderate risk significance (White) finding applicable to Units 1 and 2 and a Unit 2 White performance indicator (PI). The White finding and PI are in the mitigating systems cornerstone in the reactor safety strategic performance area. The White finding is described in NRC Final Significance Determination letter dated December 21, 2001, and was associated with the Emergency Diesel Generator (EDG) 3 wrist pin failures. The PI, Safety System Unavailability - Emergency AC Power, crossed the White threshold in the third quarter of calendar year 2001 and remained there through the current quarter. The White PI resulted mainly from the EDG 3 piston wrist pin failures and from EDG 2 output breaker problems. During this supplemental inspection, which was performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee performed an overall adequate evaluation of performance deficiencies related to the EDG 3 piston wrist pin failure and the EDG 2 output breaker problems. The depth of root cause evaluations was adequate. The corrective actions were appropriately prioritized and consistent with the identified root cause and contributing factors and provided reasonable assurance to prevent recurrence.

Inspection Report# : 2002008(pdf)

Significance: Dec 14, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Procedure For Number 1 Emergency Diesel Generator Standby Operation Alignment Technical Specifications (TS) 6.4.D requires that all procedures specified in TS 6.4.A be followed. TS 6.4.A.1 and 6.4.A.2 requires detailed procedures for operation and testing of systems and components involving nuclear safety of the station be provided. Between September 24 and October 17, 2001, the licensee failed to properly follow Operations Procedure, 1-OP-EG-001, Revision 12. Specifically, the licensee failed to ensure that Number 1 Emergency Diesel Generator (EDG) was properly aligned for standby operation in that the load limit knob was discovered by the licensee to be set to zero, thereby, disabling the Number 1 EDG to start and carry the required loads on a valid start signal. This issue was discovered during surveillance testing on October 17, 2001. This issue has been entered in the licensee's corrective action program as Plant Issue S-2001-2975.

Inspection Report# : 2001007(pdf)

Significance: Sep 25, 2001 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a condition adverse to quality, the failed piston wrist pins and piston carrier bearings in number 3 emergency diesel generator, as required by Criterion XVI.

The licensee failed to promptly investigate the cause of an increasing trend in the Number 3 Emergency Diesel Generator (EDG) lubricating oil silver concentration. As a result, the licensee did not promptly identify and correct failed piston wrist pins and piston carrier bearings in Number 3 EDG. Two apparent violations were identified. The first apparent violation involved the failure to assure that a condition adverse to quality, involving the failure of the Number 3 Emergency Diesel Generator, was promptly identified and corrected as required by 10 CFR 50, Appendix B, Criterion XVI. The second apparent violation involved the Number 3 EDG being inoperable for a period of time greater than that which was allowed by Technical Specification 3.16.B.1.a.3. The finding appears to have substantial safety significance because the failed piston wrist pins and piston carrier bearings would have caused the Number 3 EDG to fail if it had been called upon to operate for a prolonged period to mitigate accident scenarios involving the loss of offsite power. On both units the SDP calculated increase in risk resulted mainly from the time the condition existed on the Number 3 EDG and the "common cause failure to run factor," due to degraded similar components in the Number 1 EDG. In addition, the calculated increase in risk for Unit 2 was greater than Unit 1 because Unit 2 does not have high temperature seals installed on one reactor coolant pump. By letter dated, December 21, 2001, the NRC

3Q/2002 Inspection Findings - Surry 1 Page 3 of 7 informed the licensee of its final significance determination for the proposed apparent violation. The letter identified a violation characterized as a White finding. The text of the violation is: 10 CFR 50, Appendix B, Criterion XVI, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. TS [Technical Specification] 3.16.A.1 requires, in part, that a reactor shall not be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350 degrees Fahrenheit, respectively, without two diesel generators (the unit diesel generator and the shared backup diesel generator) OPERABLE. TS 3.16.B modifies the requirements of TS 3.16.A.1. Specifically, TS 3.16.B.1.a.3 requires, in part, that during power operation, if either unit's dedicated diesel generator or shared backup diesel generator is not returned to an OPERABLE status within 7 days, the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Contrary to the above, from approximately June 2000 until April 28, 2001, the licensee failed to establish measures to assure that a condition adverse to quality was promptly identified and corrected. Specifically, the licensee did not promptly identify and correct abnormal wear and eventual failure of Emergency Diesel Generator (EDG) piston wrist pins and piston carrier bearings, as evidenced by abnormally high bearing material wear products in engine oil samples, which rendered the Number 3 EDG inoperable. As a result, with the Unit 1 and 2 reactors in power operation, the Number 3 EDG was not operable from April 15 until April 28, 2001, and the licensee failed to return the Number 3 EDG to OPERABLE status within 7 days and the Unit 1 and Unit 2 reactors were not brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months as required by TS 3.16.A.1. and 3.16.B.1.a.3. A supplemental inspection of this issue, documented in report 50-280, 281/02-08, determined that the licensee had performed an overall adequate evaluation of performance deficiencies related to the failure, and that the corrective actions were appropriately prioritized and consistent with the identified root cause and contributing factors and provided reasonable assurance to prevent recurrence.

Inspection Report# : 2001006(pdf)

Inspection Report# : 2002002(pdf)

Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a fire brigade drill program as required by Surry License Condition 3.I A non-cited violation of Surry License Condition 3.I on fire protection was identified for failure to implement a fire brigade training program as required by the license condition. The licensee's fire brigade program inappropriately allowed the use of walk through drills, false alarms, and actual fires to satisfy the quarterly requirements for fire brigade drills. Consequently, the fire brigade received less than the minimum required training drills in 1998.

Therefore, it could be less effective at fighting fires. This violation was of more than minor significance because a potentially less effective fire brigade has a credible impact on safety, in that, untimely or ineffective action by the fire brigade could credibly allow a fire to affect the operability or function of a system or train required for safe shutdown.

This issue was determined to have very low safety significance because there was no identified degradation of the other parts of the fire protection defense in depth: fire barriers, fire alarms, and automatic fire suppression.

Inspection Report# : 2001002(pdf)

Significance: Dec 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50 APPENDIX B CRITERION XVI FOR FAILURE TO CORRECT A CONDITION WHICH PREVENTED THE AUXILIARY BUILDING EXHAUST FANS FROM OPERATING AFTER AN AUTOMATIC START The inspectors identified a non-cited violation in which the licensee failed to adequately address a condition adverse to quality that prevented the auxiliary ventilation exhaust filter fans from operating following an automatic start in the minimum safeguards alignment. This matter was originally identified in April 2000. This is a violation of 10 CFR 50, Appendix B, Criterion XVI. This issue was of very low safety significance since operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would be performed.

3Q/2002 Inspection Findings - Surry 1 Page 4 of 7 Inspection Report# : 2000005(pdf)

Significance: Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW A PROCEDURE WHICH RENDERED THE AMSAC SYSTEM INOPERABLE DURING POWER OPERATION The inspectors identified a non-cited violation in which the licensee failed to follow a required procedure which rendered the anticipated transients without scram mitigation system actuation circuit (AMSAC) inoperable while the plant was operating at power. This is a violation of Surry Power Station Technical Specifications, section 6.4.A.2 and is in the licensee's corrective action program as PI S-2000-1186. The risk of having the AMSAC inoperable for less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was considered to be of very low safety significance because operator recovery actions and procedures were available if needed.

Inspection Report# : 2000003(pdf)

Significance: N/A Jun 17, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation MODIFICATION TO AUXILIARY VENT SYSTEM PREVENTED PARALLEL OPERATION OF BOTH FANS IN THE MINIMUM SAFEGUARDS ALIGNMENT The inspectors identified a non-cited violation in which a modification was implemented to the auxiliary ventilation system that prevented parallel operation of both fans in the minimum safeguards alignment which would result in one or both fans tripping following an actuation signal. The post-modification testing did not verify proper operation with both fans operating simultaneously. This is a violation of 10 CFR 50, Appendix B, Criterion III and is in the licensee's corrective action program as PI S-2000-0683. The issue was of very low safety significance since the operators could have manually aligned the system for operation. The plant design allowed sufficient time to manually actuate the system such that the safety functions would performed.

Inspection Report# : 2000003(pdf)

Significance: Jun 17, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE PROCEDURE TO PROVIDE ALTERNATE SHUTDOWN CAPABILITY A non-cited violation was identified for the failure to have an adequate procedure in effect to provide alternative shutdown capability (i.e., to achieve and maintain a safe shutdown condition) in the event of a main control room fire.

This is a violation of 10 CFR 50, Appendix R, Section III.L.3 and is in the licensee's corrective action program as DR S-99-0745. This item is associated with Licensee Event Report 50-280, 281/99-003-00 which has an event date of March 31, 1999. The issue was of very low safety significance due to the very low fire initiating event frequency associated with the violation condition.

Inspection Report# : 2000003(pdf)

Barrier Integrity Significance: N/A Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Follow Refueling Procedure Technical Specifications 6.4.A.8 requires detailed written procedures be provided for Refueling Operations. Technical Specification 6.4.D requires that procedures described in Specification 6.4.A shall be followed. On November [02],

3Q/2002 Inspection Findings - Surry 1 Page 5 of 7 2001, the licensee failed to follow procedure 0-OP-4.8, in that the transfer of a spent fuel assembly was initiated prior to clearing the top of its storage position. This issue has been documented in the licensee's corrective action program as Plant Issue S-2001-3275. (No Color).

Inspection Report# : 2001004(pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Sep 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000.

Technical Specification 6.4.B.3 requires that written procedures shall be established and implemented covering implementation of the Offsite Dose Calculation Manual. Attachments 7 and 9 to Procedure VPAP-2103S, "Offsite Dose Calculation Manual (Surry)," specify the sampling and analysis frequency and the analytical detection capabilities for radiological environmental monitoring program. Analytical frequency and detection capability requirements were not met for several milk samples collected during the 3rd and 4th quarters of CY 2000 due to the vendor laboratory's failure to analyze samples in a timely manner. As a result, the licensee' ability to evaluate the milk-to-man pathway was impaired. This issue was included in the licensee's corrective action program as Plant Issue S-2001-1208.

Inspection Report# : 2001003(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly verify that a receiver's license allows receipt of a quantity of byproduct material prior to shipment as required by 10CFR 30.41(c) 10CFR 30.41(c) requires that, before transferring byproduct material, the licensee transferring byproduct material shall verify that the transferee's (the receiver's) license authorizes the receipt of the type, form, and quantity of byproduct material to be transferred. On May 24, 2000, the licensee failed to properly verify, prior to shipment, that the receiver's license authorized the receipt of the quantity of byproduct material to be transferred. This occurrence was documented in plant issue report No. S-2000-1920.

Inspection Report# : 2000006(pdf)

Significance: Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to survey an exclusive use vehicle before returning it to service as required by 10 CFR 71.5 and 49 CFR 177.843 10 CFR 71.5 requires NRC licensees to comply with all applicable provisions of 49 CFR when transporting or

3Q/2002 Inspection Findings - Surry 1 Page 6 of 7 receiving licensed radioactive materials. 49 CFR 177.843 requires receivers of radioactive materials packages sent exclusive use to survey the vehicle prior to it being returned to service. On September 28, 2000, an exclusive use shipment of radioactive material (surface contaminated object) was received by the licensee and the vehicle returned to service without being surveyed, as described in plant issue S-2000-2126.

Inspection Report# : 2000006(pdf)

Physical Protection Significance: Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Meet Security Plan Requirements A Non-Cited Violation was identified associated with pargraph 4.3.2.a of the Surry Power Station Physical Security Plan Revision 7 While the risk is low in this case, the issue was identified as more than a minor finding because allowing individuals to penentrate the Protected Area perimeter without being detected can have a credible impact on safety and can be viewed as a precursor to a more significant event.

Inspection Report# : 2001008(pdf)

Significance: N/A Oct 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation WILLFUL FAILURE OF AN INDIVIDUAL TO REPORT AN ARREST IN ACCORDANCE WITH PHYSICAL SECURITY PLAN AND VPAP-0105.

The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individual's position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance. Due to the willful nature of the issue, the Significance Determination Process was not used. No color was assigned.

Inspection Report# : 2000008(pdf)

Significance: Jun 17, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH THE REQUIREMENTS OF THE PHYSICAL SECURITY PLAN The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, the officer providing the last access control function, at the Primary Access Control on March 27, 2000, and at the Secondary Access Portal on April 26, 2000, did not remain isolated within a hardened structure in order to satisfy the requirements of the PSP. This is in the licensee's corrective action program as PI S-2000-1024.

Based on the other response and assessment capabilities in place as well as the licensee's previous four-quarter performance in this area, these findings were determined to be of very low risk significance.

Inspection Report# : 2000003(pdf)

Miscellaneous Significance: N/A Dec 14, 2001 Identified By: NRC Item Type: FIN Finding

3Q/2002 Inspection Findings - Surry 1 Page 7 of 7 Problem Identification and Resolution Annual Inspection Results The licensee's threshold for identifying problems and entering them into the corrective action program (CAP) was at an appropriate level. Self-disclosing events, equipment failures and human errors were appropriately evaluated. Formal processes such as audits and self-assessments performed by the licensee's staff and outside organizations were effective in identifying issues. The inspectors concluded that external industry operating experience and NRC generic communications had been evaluated for plant applicability and incorporated into the CAP as appropriate. The inspectors determined that the licensee was effective in prioritizing and evaluating issues commensurate with the safety significance. Licensee reviews adequately addressed the extent of condition, generic implications, common cause failure modes, and previous occurrences. Significant conditions adverse to quality were evaluated and resolved in a timely manner. Plant employees were not reluctant to report safety concerns.

Inspection Report# : 2001007(pdf)

Significance: Dec 14, 2001 Identified By: NRC Item Type: FIN Finding Adequacy Of Emergency Diesel Generator Contingency Plans To Meet Intent Of Nuclear Energy Institute 99-02 Guidance And Report Unavailability Time Accurately A finding was identified when the Revised Oversight Process Working Group determined that the recovery actions in an emergency diesel generator (EDG) surveillance procedure did not meet the guidelines of NEI 99-02, and the corresponding unavailability hours should be counted towards the Safety System Unavailability - Emergency AC Power Performance Indicator (PI) during the testing of the EDG. When the licensee revised the PI data, the PI on Unit 1 changed from green to white for the fourth quarter of 2001 and the first and second quarters of 2002. The finding was of very low safety significance because the added unavailability hours reflect only a small portion of the time required for the PI to exceed the green/white threshold. The majority of the unavailability hours were the result of issues that were previously identified and inspected, and therefore, no additional regulatory response is required.

Inspection Report# : 2001007(pdf)

Inspection Report# : 2002003(pdf)

Significance: N/A Dec 15, 2000 Identified By: NRC Item Type: FIN Finding THE LICENSEE WAS EFFECTIVE AT IDENTIFYING AND RESOLVING PROBLEMS The licensee was effective at identifying problems and entering them into the corrective action program. The threshold for entering problems into the corrective action program was low. Operating experience was appropriately incorporated into plant procedures and activities. Generally, problems entered into the corrective action program were adequately evaluated and appropriate corrective actions were identified. Category 1 and 2 root cause evaluations, performed for the most and next most significant issues, respectively, were thorough and specified corrective actions were appropriate.

Although risk was not formally used in prioritizing issues, corrective actions were usually implemented in a timely manner commensurate with their safety significance. Licensee audit and self-assessment results were consistent with NRC observations and identified deficiencies were entered into the corrective action program. Based on interviews, a safety conscious work environment was present where employees felt free to raise nuclear safety concerns. However, some negative observations were identified involving time to resolve some issues and the adequacy of some evaluations and resolutions. These negative observations involved issues that were of very low safety significance.

Inspection Report# : 2000009(pdf)

Last modified : December 02, 2002

4Q/2002 Inspection Findings - Surry 1 Page 1 of 2 Surry 1 Initiating Events Mitigating Systems Significance: Sep 28, 2002 Identified By: NRC Item Type: FIN Finding Adequacy of emergency diesel generator contingency plans to meet intent of Nuclear Energy Institute (NEI) 99-02 guidance and report unavailability time accurately A finding was identified when the Revised Oversight Process Working Group determined that the recovery actions in an emergency diesel generator (EDG) surveillance procedure did not meet the guidelines of NEI 99-02, and the corresponding unavailability hours should be counted towards the Safety System Unavailability - Emergency AC Power Performance Indicator (PI) during the testing of the EDG. When the licensee revised the PI data, the PI on Unit 1 changed from green to white for the fourth quarter of 2001 and the first and second quarters of 2002. The finding was of very low safety significance because the added unavailability hours reflect only a small portion of the time required for the PI to exceed the green/white threshold. The majority of the unavailability hours were the result of issues that were previously identified and inspected, and therefore, no additional regulatory response is required.

Inspection Report# : 2002003(pdf)

Significance: Jun 29, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide appropriate instructions to ensure proper operation of the emergency switchgear room chillers following a loss of instrument air A Non-Cited Violation of Technical Specification 6.4.A was identified due to an inadequate abnormal procedure. Abnormal procedure (AP) -

40, "Non-Recoverable Loss of Instrument Air," did not contain adequate guidance to ensure continued operation of the emergency switchgear room chillers following a loss of instrument air. The finding was of very low safety significance due to the combination of events that would have to occur for the emergency switchgear room components to be adversely affected by the loss of the chillers. The combination of events included a medium or large break loss of coolant accident coupled with a loss of offsite power during the winter months (cold service water temperatures).

Inspection Report# : 2002002(pdf)

Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Determine the risk significance of the failure to provide proper separation between the 125V DC busses A finding was identified for not providing proper separation between the 125V DC busses. A single failure could affect both redundant DC busses on a unit and encumber normal decay heat removal systems. The finding was of very low safety significance due to plant design features which mitigate the consequences of a fault within the DC system. Specifically, there are numerous alternative methods of decay heat removal available with simple operator actions.

Inspection Report# : 2002002(pdf)

Inspection Report# : 2002003(pdf)

Significance: N/A Apr 12, 2002 Identified By: NRC Item Type: FIN Finding Supplemental Inspection Results For Unit 1 and 2 White Finding, Emergency Diesel Generator Bearing Failures, and Unit 2 White Performance Indicator, Safety System Unavailability - Emergency AC Power This supplemental inspection was performed by the NRC to assess the licensee's evaluation and corrective actions associated with a low to moderate risk significance (White) finding applicable to Units 1 and 2 and a Unit 2 White performance indicator (PI). The White finding and PI are in the mitigating systems cornerstone in the reactor safety strategic performance area. The White finding is described in NRC Final Significance Determination letter dated December 21, 2001, and was associated with the Emergency Diesel Generator (EDG) 3 wrist pin

4Q/2002 Inspection Findings - Surry 1 Page 2 of 2 failures. The PI, Safety System Unavailability - Emergency AC Power, crossed the White threshold in the third quarter of calendar year 2001 and remained there through the current quarter. The White PI resulted mainly from the EDG 3 piston wrist pin failures and from EDG 2 output breaker problems. During this supplemental inspection, which was performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee performed an overall adequate evaluation of performance deficiencies related to the EDG 3 piston wrist pin failure and the EDG 2 output breaker problems. The depth of root cause evaluations was adequate. The corrective actions were appropriately prioritized and consistent with the identified root cause and contributing factors and provided reasonable assurance to prevent recurrence.

Inspection Report# : 2002008(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : March 26, 2003

1Q/2003 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 1Q/2003 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Feb 14, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Test Diesel Driven Fire Pump Automatic Start Features A failure to establish written operating test procedures to demonstrate the functional capability of the diesel-driven fire pump (DDFP) loss-of-power automatic start feature could have resulted in a loss of fire suppression water during a loss-of- offsite power condition. A non-cited violation of 10 CFR 50.48 was identified. This finding is greater than minor because it is associated with fire protection performance and degraded the ability to meet the mitigating systems cornerstone objective. The finding is considered to have very low safety significance because the DDFP successfully started when a loss-of-power test was performed.

Inspection Report# : 2003007(pdf)

Significance: Feb 14, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control of Diesel Driven Fire Pump Fuel Oil Isolation Valve A failure to properly implement and maintain an adequate fire protection program inspection and valve position control process could have resulted in isolation of the fuel oil supply to the diesel-driven fire pump (DDFP). The position of the DDFP fuel oil supply valve was not being controlled by the licensee. A non-cited violation of 10 CFR 50.48 was identified. This finding is greater than minor because it is associated with fire protection performance and degraded the ability to meet the mitigating systems cornerstone objective. The finding is considered to have very low safety significance because the fuel oil supply valve was in its proper position and it had not been mis-positioned in the past.

Inspection Report# : 2003007(pdf)

Significance: Sep 28, 2002 Identified By: NRC Item Type: FIN Finding Adequacy of emergency diesel generator contingency plans to meet intent of Nuclear Energy Institute (NEI) 99-02 guidance and report unavailability time accurately A finding was identified when the Revised Oversight Process Working Group determined that the recovery actions in an emergency diesel generator (EDG) surveillance procedure did not meet the guidelines of NEI 99-02, and the corresponding unavailability hours should be counted towards the Safety System Unavailability - Emergency AC Power Performance Indicator (PI) during the testing of the EDG. When the licensee revised the PI data, the PI on Unit file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/22/2003

1Q/2003 Inspection Findings - Surry 1 Page 2 of 3 1 changed from green to white for the fourth quarter of 2001 and the first and second quarters of 2002. The finding was of very low safety significance because the added unavailability hours reflect only a small portion of the time required for the PI to exceed the green/white threshold. The majority of the unavailability hours were the result of issues that were previously identified and inspected, and therefore, no additional regulatory response is required.

Inspection Report# : 2002003(pdf)

Significance: Jun 29, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide appropriate instructions to ensure proper operation of the emergency switchgear room chillers following a loss of instrument air A Non-Cited Violation of Technical Specification 6.4.A was identified due to an inadequate abnormal procedure.

Abnormal procedure (AP) - 40, "Non-Recoverable Loss of Instrument Air," did not contain adequate guidance to ensure continued operation of the emergency switchgear room chillers following a loss of instrument air. The finding was of very low safety significance due to the combination of events that would have to occur for the emergency switchgear room components to be adversely affected by the loss of the chillers. The combination of events included a medium or large break loss of coolant accident coupled with a loss of offsite power during the winter months (cold service water temperatures).

Inspection Report# : 2002002(pdf)

Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Determine the risk significance of the failure to provide proper separation between the 125V DC busses A finding was identified for not providing proper separation between the 125V DC busses. A single failure could affect both redundant DC busses on a unit and encumber normal decay heat removal systems. The finding was of very low safety significance due to plant design features which mitigate the consequences of a fault within the DC system.

Specifically, there are numerous alternative methods of decay heat removal available with simple operator actions.

Inspection Report# : 2002002(pdf)

Inspection Report# : 2002003(pdf)

Significance: N/A Apr 12, 2002 Identified By: NRC Item Type: FIN Finding Supplemental Inspection Results For Unit 1 and 2 White Finding, Emergency Diesel Generator Bearing Failures, and Unit 2 White Performance Indicator, Safety System Unavailability - Emergency AC Power This supplemental inspection was performed by the NRC to assess the licensee's evaluation and corrective actions associated with a low to moderate risk significance (White) finding applicable to Units 1 and 2 and a Unit 2 White performance indicator (PI). The White finding and PI are in the mitigating systems cornerstone in the reactor safety strategic performance area. The White finding is described in NRC Final Significance Determination letter dated December 21, 2001, and was associated with the Emergency Diesel Generator (EDG) 3 wrist pin failures. The PI, Safety System Unavailability - Emergency AC Power, crossed the White threshold in the third quarter of calendar year 2001 and remained there through the current quarter. The White PI resulted mainly from the EDG 3 piston wrist pin failures and from EDG 2 output breaker problems. During this supplemental inspection, which was performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee performed an overall adequate evaluation of performance deficiencies related to the EDG 3 piston wrist pin failure and the EDG 2 output breaker problems. The depth of root cause evaluations was adequate. The corrective actions were appropriately prioritized and consistent with the identified root cause and contributing factors and provided reasonable assurance to prevent file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/22/2003

1Q/2003 Inspection Findings - Surry 1 Page 3 of 3 recurrence.

Inspection Report# : 2002008(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : May 30, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 07/22/2003

2Q/2003 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 2Q/2003 Plant Inspection Findings Initiating Events Significance: Apr 05, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Evaluate and Approve the Storage of Flammable Materials in the Vicinity of Safety-related Equipment The licensee failed to properly evaluate and approve the storage of flammable materials in the vicinity of safety-related equipment in the Auxiliary Building and the Unit 2 Safeguards area. An NRC-identified non-cited violation of the Technical Specification 6.4.E was identified. This finding is more than minor because the amount of material improperly stored exceeded the quantity specified in the licensee's Combustible Loading Analysis. The finding is of very low safety significance because it did not cause the impairment or degradation of a fire protection feature or defense in depth.

Inspection Report# : 2003002(pdf)

Mitigating Systems Significance: Apr 05, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Take Adequate Corrective Actions to Preclude Additional De-alloying Failures for Valves in the Charging Service Water System The licensee failed to take adequate corrective actions to preclude additional de-alloying failures for valves in the charging service water system after a failure had occurred in August 2001. A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI was identified. This finding is more than minor because of the potential impact on the reliability of the safety injection system. The finding is of very low safety significance because the failure did not actually cause the loss of cooling to the charging pumps.

Inspection Report# : 2003002(pdf)

Significance: Feb 14, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Test Diesel Driven Fire Pump Automatic Start Features A failure to establish written operating test procedures to demonstrate the functional capability of the diesel-driven fire pump (DDFP) loss-of-power automatic start feature could have resulted in a loss of fire suppression water during a loss-of- offsite power condition. A non-cited violation of 10 CFR 50.48 was identified. This finding is greater than minor because it is associated with fire protection performance and degraded the ability to meet the mitigating systems file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 10/08/2003

2Q/2003 Inspection Findings - Surry 1 Page 2 of 3 cornerstone objective. The finding is considered to have very low safety significance because the DDFP successfully started when a loss-of-power test was performed.

Inspection Report# : 2003007(pdf)

Significance: Feb 14, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control of Diesel Driven Fire Pump Fuel Oil Isolation Valve A failure to properly implement and maintain an adequate fire protection program inspection and valve position control process could have resulted in isolation of the fuel oil supply to the diesel-driven fire pump (DDFP). The position of the DDFP fuel oil supply valve was not being controlled by the licensee. A non-cited violation of 10 CFR 50.48 was identified. This finding is greater than minor because it is associated with fire protection performance and degraded the ability to meet the mitigating systems cornerstone objective. The finding is considered to have very low safety significance because the fuel oil supply valve was in its proper position and it had not been mis-positioned in the past.

Inspection Report# : 2003007(pdf)

Significance: Sep 28, 2002 Identified By: NRC Item Type: FIN Finding Adequacy of emergency diesel generator contingency plans to meet intent of Nuclear Energy Institute (NEI) 99-02 guidance and report unavailability time accurately A finding was identified when the Revised Oversight Process Working Group determined that the recovery actions in an emergency diesel generator (EDG) surveillance procedure did not meet the guidelines of NEI 99-02, and the corresponding unavailability hours should be counted towards the Safety System Unavailability - Emergency AC Power Performance Indicator (PI) during the testing of the EDG. When the licensee revised the PI data, the PI on Unit 1 changed from green to white for the fourth quarter of 2001 and the first and second quarters of 2002. The finding was of very low safety significance because the added unavailability hours reflect only a small portion of the time required for the PI to exceed the green/white threshold. The majority of the unavailability hours were the result of issues that were previously identified and inspected, and therefore, no additional regulatory response is required.

Inspection Report# : 2002003(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 10/08/2003

2Q/2003 Inspection Findings - Surry 1 Page 3 of 3 Physical Protection Miscellaneous Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 10/08/2003

3Q/2003 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 3Q/2003 Plant Inspection Findings Initiating Events Significance: Apr 05, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Evaluate and Approve the Storage of Flammable Materials in the Vicinity of Safety-related Equipment The licensee failed to properly evaluate and approve the storage of flammable materials in the vicinity of safety-related equipment in the Auxiliary Building and the Unit 2 Safeguards area.

An NRC-identified non-cited violation of the Technical Specification 6.4.E was identified. This finding is more than minor because the amount of material improperly stored exceeded the quantity specified in the licensee's Combustible Loading Analysis. The finding is of very low safety significance because it did not cause the impairment or degradation of a fire protection feature or defense in depth.

Inspection Report# : 2003002(pdf)

Mitigating Systems Significance: Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator No. 3 Bus-Tie Breaker Control Circuit Design Deficiency The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control because emergency diesel generator (EDG) no. 3 could have been overloaded following a concurrent loss-of-offsite power on Units 1 and 2. The licensee has resolved the problem through a modification of the breaker control circuitry.

This finding is greater than minor because it is associated with EDG performance and affects the mitigating systems cornerstone objective. The finding is of very low safety significance because the inspectors determined that the automatically connected loads are less than the 168-hour rating of the EDG.

Inspection Report# : 2003004(pdf)

Significance: Apr 05, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Take Adequate Corrective Actions to Preclude Additional De-alloying Failures for Valves in the Charging Service Water System The licensee failed to take adequate corrective actions to preclude additional de-alloying failures for valves in the file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 01/12/2004

3Q/2003 Inspection Findings - Surry 1 Page 2 of 3 charging service water system after a failure had occurred in August 2001.

A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI was identified. This finding is more than minor because of the potential impact on the reliability of the safety injection system. The finding is of very low safety significance because the failure did not actually cause the loss of cooling to the charging pumps.

Inspection Report# : 2003002(pdf)

Significance: Feb 14, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Test Diesel Driven Fire Pump Automatic Start Features A failure to establish written operating test procedures to demonstrate the functional capability of the diesel-driven fire pump (DDFP) loss-of-power automatic start feature could have resulted in a loss of fire suppression water during a loss-of- offsite power condition.

A non-cited violation of 10 CFR 50.48 was identified. This finding is greater than minor because it is associated with fire protection performance and degraded the ability to meet the mitigating systems cornerstone objective. The finding is considered to have very low safety significance because the DDFP successfully started when a loss-of-power test was performed.

Inspection Report# : 2003007(pdf)

Significance: Feb 14, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control of Diesel Driven Fire Pump Fuel Oil Isolation Valve A failure to properly implement and maintain an adequate fire protection program inspection and valve position control process could have resulted in isolation of the fuel oil supply to the diesel-driven fire pump (DDFP). The position of the DDFP fuel oil supply valve was not being controlled by the licensee.

A non-cited violation of 10 CFR 50.48 was identified. This finding is greater than minor because it is associated with fire protection performance and degraded the ability to meet the mitigating systems cornerstone objective. The finding is considered to have very low safety significance because the fuel oil supply valve was in its proper position and it had not been mis-positioned in the past.

Inspection Report# : 2003007(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 01/12/2004

3Q/2003 Inspection Findings - Surry 1 Page 3 of 3 Public Radiation Safety Physical Protection Miscellaneous Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 01/12/2004

4Q/2003 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 4Q/2003 Plant Inspection Findings Initiating Events Significance: Apr 05, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Evaluate and Approve the Storage of Flammable Materials in the Vicinity of Safety-related Equipment The licensee failed to properly evaluate and approve the storage of flammable materials in the vicinity of safety-related equipment in the Auxiliary Building and the Unit 2 Safeguards area.

An NRC-identified non-cited violation of the Technical Specification 6.4.E was identified. This finding is more than minor because the amount of material improperly stored exceeded the quantity specified in the licensee's Combustible Loading Analysis. The finding is of very low safety significance because it did not cause the impairment or degradation of a fire protection feature or defense in depth.

Inspection Report# : 2003002(pdf)

Mitigating Systems Significance: Dec 05, 2003 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater Pump Design Basis not Translated into Procedures The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criteria III (Design Control), in that, a design basis requirement for the Unit 1 auxiliary feedwater pump turbine governor oil viscosity was not correctly translated into a March 2001 procedure revision. The procedure revision failed to require the main steam valve house room temperature to be above that required for minimum vendor specified governor oil viscosity. This non-cited violation contributed to the pump's failure to continue to operate after starting in response to a reactor trip on January 25, 2003.

This finding is greater than minor because it affected the reliability of the Unit 1 turbine driven auxiliary feedwater pump. However, the finding was determined to be of very low safety significance since (1) except for January 25, 2003, conditions after the procedure change in March 2001 would not have been expected to lower main steam valve house room temperatures below acceptable temperatures, and (2) on January 25, 2003, the two motor driven auxiliary feedwater pumps were operable and performed as expected. Surry personnel tracked corrective actions for this issue under plant issue S-2003-5822.

Inspection Report# : 2003009(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 04/22/2004

4Q/2003 Inspection Findings - Surry 1 Page 2 of 3 Significance: Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator No. 3 Bus-Tie Breaker Control Circuit Design Deficiency The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control because emergency diesel generator (EDG) no. 3 could have been overloaded following a concurrent loss-of-offsite power on Units 1 and 2. The licensee has resolved the problem through a modification of the breaker control circuitry.

This finding is greater than minor because it is associated with EDG performance and affects the mitigating systems cornerstone objective. The finding is of very low safety significance because the inspectors determined that the automatically connected loads are less than the 168-hour rating of the EDG.

Inspection Report# : 2003004(pdf)

Significance: Apr 05, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Take Adequate Corrective Actions to Preclude Additional De-alloying Failures for Valves in the Charging Service Water System The licensee failed to take adequate corrective actions to preclude additional de-alloying failures for valves in the charging service water system after a failure had occurred in August 2001.

A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI was identified. This finding is more than minor because of the potential impact on the reliability of the safety injection system. The finding is of very low safety significance because the failure did not actually cause the loss of cooling to the charging pumps.

Inspection Report# : 2003002(pdf)

Significance: Feb 14, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Test Diesel Driven Fire Pump Automatic Start Features A failure to establish written operating test procedures to demonstrate the functional capability of the diesel-driven fire pump (DDFP) loss-of-power automatic start feature could have resulted in a loss of fire suppression water during a loss-of- offsite power condition.

A non-cited violation of 10 CFR 50.48 was identified. This finding is greater than minor because it is associated with fire protection performance and degraded the ability to meet the mitigating systems cornerstone objective. The finding is considered to have very low safety significance because the DDFP successfully started when a loss-of-power test was performed.

Inspection Report# : 2003007(pdf)

Significance: Feb 14, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control of Diesel Driven Fire Pump Fuel Oil Isolation Valve A failure to properly implement and maintain an adequate fire protection program inspection and valve position control file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 04/22/2004

4Q/2003 Inspection Findings - Surry 1 Page 3 of 3 process could have resulted in isolation of the fuel oil supply to the diesel-driven fire pump (DDFP). The position of the DDFP fuel oil supply valve was not being controlled by the licensee.

A non-cited violation of 10 CFR 50.48 was identified. This finding is greater than minor because it is associated with fire protection performance and degraded the ability to meet the mitigating systems cornerstone objective. The finding is considered to have very low safety significance because the fuel oil supply valve was in its proper position and it had not been mis-positioned in the past.

Inspection Report# : 2003007(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Significance: N/A Dec 05, 2003 Identified By: NRC Item Type: FIN Finding Biennial Problem Identification and Resolution Report The team concluded that Surry personnel were properly identifying problems and entering them into the corrective action program at a threshold that supported safe plant operation. The team did not identify instances where conditions adverse to quality were handled outside the corrective action process. The team further concluded that evaluations were prioritized and completed in a timely fashion consistent with the safety significance of the issue. Cause evaluations generally were found to address technical issues to a depth necessary to identify likely causes. The team identified one finding regarding a less than adequate procedure change evaluation that impacted the reliability of the Unit 1 turbine driven auxiliary feedwater pump. The team found that corrective actions were adequately tracked and completed in a time frame commensurate with their safety significance.

Inspection Report# : 2003009(pdf)

Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\SUR1\sur1_pim.html 04/22/2004

1Q/2004 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 1Q/2004 Plant Inspection Findings Initiating Events Significance: Apr 05, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Evaluate and Approve the Storage of Flammable Materials in the Vicinity of Safety-related Equipment The licensee failed to properly evaluate and approve the storage of flammable materials in the vicinity of safety-related equipment in the Auxiliary Building and the Unit 2 Safeguards area.

An NRC-identified non-cited violation of the Technical Specification 6.4.E was identified. This finding is more than minor because the amount of material improperly stored exceeded the quantity specified in the licensee's Combustible Loading Analysis. The finding is of very low safety significance because it did not cause the impairment or degradation of a fire protection feature or defense in depth.

Inspection Report# : 2003002(pdf)

Mitigating Systems Significance: TBD Jan 07, 2004 Identified By: NRC Item Type: AV Apparent Violation Alternative Shutdown Capability and Response Procedures Not Adequate to Ensure Safe Shutdown of Unit 1 Preliminary White. An apparent violation of 10 CFR 50, Appendix R, Sections III.L.2.b and III.L.3 was identified, in that, for a severe fire in the Emergency Switchgear and Relay Room Number 1 (Fire Area 3), the licensee's fire response procedures were not effective in assuring a safe shutdown of the Unit 1 reactor. The licensee has revised the affected fire response procedures and is evaluating the need for additional corrective action.

This finding is greater than minor because it was associated with "protection against one of the external factors" attribute. It affected the objective of the Initiating Events cornerstone to limit the likelihood events that challenge critical safety functions as well as affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events.

This degraded condition increased plant risk because, if a severe fire occurred in Fire Area 3, these procedures may not preclude an extended loss of reactor coolant pump seal injection flow and may initiate a reactor coolant pump seal loss of coolant accident which could result in pressurizer level failing to be maintained within the indicating range as required.

Inspection Report# : 2003008(pdf)

Significance: Jan 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation Fire Response Procedures 1-FCA-3.00 And 0-FCA-14.00 Not Adequate To Ensure Safe Shutdown Of Unit 1.

A Green non-cited violation of 10 CFR 50, Appendix R, Sections III.L.2.b and III.L.3, was identified, in that, for a severe fire in the Unit 1 Cable Vault and Tunnel (Fire Area 1), the licensee's alternative shutdown capability may not ensure that the reactor coolant makeup function would be capable of maintaining the reactor coolant level within the level indication of the pressurizer. The licensee has entered this finding into its corrective action program.

This finding is greater than minor because it was associated with "protection against one of the external factors" attribute. It affected the objective of the Initiating Events cornerstone to limit the likelihood events that challenge critical safety functions as well as affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events.

This finding was determined to be of very low safety significance because the likelihood of a severe fire in the service building cable vault (SBCV) or the cable tunnel that could cause a loss of all three Unit 1 charging pumps is very low and a 3-hour rated fire door would prevent a severe fire in the remaining sections of Fire Area 1 from spreading through the cable tunnel to the SBCV.

Inspection Report# : 2003008(pdf) 07/14/2004

1Q/2004 Inspection Findings - Surry 1 Page 2 of 3 Significance: Jan 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation Alternate Shutdown Panel Ventilation System Not Independent From Impacts Of A Main Control Room Fire.

A Green non-cited violation was identified for failure to comply with 10 CFR 50, Appendix R, Sections III.G.3.a and III.L.3. Specifically, the shared ventilation system between the main control room (MCR) and the Unit 1 and Unit 2 emergency switchgear and relay rooms (ESGRs),

did not have adequate separation, isolation, or barriers to preclude smoke and toxic gases from being transported to the ESGRs during a fire in the MCR. The alternative shutdown capability for an MCR fire is located in each unit's ESGR, respectively. Consequently, operators may not have the environmental conditions or visibility to safely man and accomplish a successful shutdown of either Unit 1 or Unit 2 from the Auxiliary Shutdown Panels. The licensee has entered this finding into its corrective action program.

This finding is greater than minor because it was associated with the "protection against external factors" attribute and affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events. This finding was determined to be of very low safety significance because heat from a fire, and the natural buoyancy of smoke, will cause the smoke gas layer to accumulate near the ceiling of the MCR (away from the ESGRs), the likelihood of a severe fire in the MCR is low, and the prompt response and actions of the MCR operators and the fire brigade would prevent any fires that start from becoming severe.

Inspection Report# : 2003008(pdf)

Significance: Dec 05, 2003 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater Pump Design Basis not Translated into Procedures The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criteria III (Design Control), in that, a design basis requirement for the Unit 1 auxiliary feedwater pump turbine governor oil viscosity was not correctly translated into a March 2001 procedure revision. The procedure revision failed to require the main steam valve house room temperature to be above that required for minimum vendor specified governor oil viscosity. This non-cited violation contributed to the pump's failure to continue to operate after starting in response to a reactor trip on January 25, 2003.

This finding is greater than minor because it affected the reliability of the Unit 1 turbine driven auxiliary feedwater pump. However, the finding was determined to be of very low safety significance since (1) except for January 25, 2003, conditions after the procedure change in March 2001 would not have been expected to lower main steam valve house room temperatures below acceptable temperatures, and (2) on January 25, 2003, the two motor driven auxiliary feedwater pumps were operable and performed as expected. Surry personnel tracked corrective actions for this issue under plant issue S-2003-5822.

Inspection Report# : 2003009(pdf)

Significance: Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator No. 3 Bus-Tie Breaker Control Circuit Design Deficiency The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control because emergency diesel generator (EDG) no. 3 could have been overloaded following a concurrent loss-of-offsite power on Units 1 and 2. The licensee has resolved the problem through a modification of the breaker control circuitry.

This finding is greater than minor because it is associated with EDG performance and affects the mitigating systems cornerstone objective. The finding is of very low safety significance because the inspectors determined that the automatically connected loads are less than the 168-hour rating of the EDG.

Inspection Report# : 2003004(pdf)

Significance: Apr 05, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Take Adequate Corrective Actions to Preclude Additional De-alloying Failures for Valves in the Charging Service Water System The licensee failed to take adequate corrective actions to preclude additional de-alloying failures for valves in the charging service water system after a failure had occurred in August 2001.

A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI was identified. This finding is more than minor because of the potential impact on the reliability of the safety injection system. The finding is of very low safety significance because the failure did not actually cause the loss of cooling to the charging pumps.

07/14/2004

1Q/2004 Inspection Findings - Surry 1 Page 3 of 3 Inspection Report# : 2003002(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Significance: N/A Dec 05, 2003 Identified By: NRC Item Type: FIN Finding Biennial Problem Identification and Resolution Report The team concluded that Surry personnel were properly identifying problems and entering them into the corrective action program at a threshold that supported safe plant operation. The team did not identify instances where conditions adverse to quality were handled outside the corrective action process. The team further concluded that evaluations were prioritized and completed in a timely fashion consistent with the safety significance of the issue. Cause evaluations generally were found to address technical issues to a depth necessary to identify likely causes. The team identified one finding regarding a less than adequate procedure change evaluation that impacted the reliability of the Unit 1 turbine driven auxiliary feedwater pump. The team found that corrective actions were adequately tracked and completed in a time frame commensurate with their safety significance.

Inspection Report# : 2003009(pdf)

Last modified : May 05, 2004 07/14/2004

2Q/2004 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 2Q/2004 Plant Inspection Findings Initiating Events Significance: N/A Apr 28, 2004 Identified By: NRC Item Type: FIN Finding Results of Supplemental Inspection for White Performance Indicator This supplemental inspection was conducted to assess the licensee's evaluation associated with a White performance indicator in the initiating events cornerstone. The Unplanned Scrams per 7,000 Critical Hours Performance Indicator crossed the threshold from Green to White in the third quarter of calendar year 2003. Specifically, the licensee experienced two reactor trips during the first quarter of 2003, one reactor trip during the second quarter of 2003, and one reactor trip in the third quarter of 2003. The first reactor trip, which occurred on January 14, 2003, was a manual trip from approximately 100 percent reactor power due to high temperature and shaft vibration alarms on the C reactor coolant pump. The second reactor trip, which occurred on January 25, 2003, was an automatic trip from approximately 27 percent reactor power due to problems associated with manually controlling steam generator water level. The third reactor trip, which occurred on June 13, 2003, was a manual trip from less than one percent reactor power due to a control rod misalignment. The fourth reactor trip, which occurred on September 18, 2003, was a manual reactor trip from approximately 79 percent reactor power due to inclement weather conditions and a loss of the 1G and 2G buses which supplied power to all the circulating water pumps for both units.

The licensee's problem identification, root cause and extent-of-condition evaluations, and corrective actions for the four reactor trips were adequate.

Common cause aspects linking the four reactor trips from a risk perspective were not evident.

Inspection Report# : 2004009(pdf)

Mitigating Systems Significance: TBD Jan 07, 2004 Identified By: NRC Item Type: AV Apparent Violation Alternative Shutdown Capability and Response Procedures Not Adequate to Ensure Safe Shutdown of Unit 1 Preliminary White. An apparent violation of 10 CFR 50, Appendix R, Sections III.L.2.b and III.L.3 was identified, in that, for a severe fire in the Emergency Switchgear and Relay Room Number 1 (Fire Area 3), the licensee's fire response procedures were not effective in assuring a safe shutdown of the Unit 1 reactor. The licensee has revised the affected fire response procedures and is evaluating the need for additional corrective action.

This finding is greater than minor because it was associated with "protection against one of the external factors" attribute. It affected the objective of the Initiating Events cornerstone to limit the likelihood events that challenge critical safety functions as well as affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events. This degraded condition increased plant risk because, if a severe fire occurred in Fire Area 3, these procedures may not preclude an extended loss of reactor coolant pump seal injection flow and may initiate a reactor coolant pump seal loss of coolant accident which could result in pressurizer level failing to be maintained within the indicating range as required.

Inspection Report# : 2003008(pdf)

Significance: Jan 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation Fire Response Procedures 1-FCA-3.00 And 0-FCA-14.00 Not Adequate To Ensure Safe Shutdown Of Unit 1.

A Green non-cited violation of 10 CFR 50, Appendix R, Sections III.L.2.b and III.L.3, was identified, in that, for a severe fire in the Unit 1 Cable Vault and Tunnel (Fire Area 1), the licensee's alternative shutdown capability may not ensure that the reactor coolant makeup function would be capable of maintaining the reactor coolant level within the level indication of the pressurizer. The licensee has entered this finding into its corrective action program.

This finding is greater than minor because it was associated with "protection against one of the external factors" attribute. It affected the objective of the Initiating Events cornerstone to limit the likelihood events that challenge critical safety functions as well as affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events. This finding was determined to be of very low safety significance because the likelihood of a severe fire in the service building cable vault (SBCV) or the cable tunnel that could cause a loss of all three Unit 1 charging pumps is very low and a 3-hour rated fire door would prevent a severe fire in the remaining sections of Fire Area 1 from spreading through the cable tunnel to the SBCV.

Inspection Report# : 2003008(pdf)

2Q/2004 Inspection Findings - Surry 1 Page 2 of 3 Significance: Jan 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation Alternate Shutdown Panel Ventilation System Not Independent From Impacts Of A Main Control Room Fire.

A Green non-cited violation was identified for failure to comply with 10 CFR 50, Appendix R, Sections III.G.3.a and III.L.3. Specifically, the shared ventilation system between the main control room (MCR) and the Unit 1 and Unit 2 emergency switchgear and relay rooms (ESGRs), did not have adequate separation, isolation, or barriers to preclude smoke and toxic gases from being transported to the ESGRs during a fire in the MCR. The alternative shutdown capability for an MCR fire is located in each unit's ESGR, respectively. Consequently, operators may not have the environmental conditions or visibility to safely man and accomplish a successful shutdown of either Unit 1 or Unit 2 from the Auxiliary Shutdown Panels. The licensee has entered this finding into its corrective action program.

This finding is greater than minor because it was associated with the "protection against external factors" attribute and affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events. This finding was determined to be of very low safety significance because heat from a fire, and the natural buoyancy of smoke, will cause the smoke gas layer to accumulate near the ceiling of the MCR (away from the ESGRs), the likelihood of a severe fire in the MCR is low, and the prompt response and actions of the MCR operators and the fire brigade would prevent any fires that start from becoming severe.

Inspection Report# : 2003008(pdf)

Significance: Dec 05, 2003 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater Pump Design Basis not Translated into Procedures The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criteria III (Design Control), in that, a design basis requirement for the Unit 1 auxiliary feedwater pump turbine governor oil viscosity was not correctly translated into a March 2001 procedure revision. The procedure revision failed to require the main steam valve house room temperature to be above that required for minimum vendor specified governor oil viscosity. This non-cited violation contributed to the pump's failure to continue to operate after starting in response to a reactor trip on January 25, 2003.

This finding is greater than minor because it affected the reliability of the Unit 1 turbine driven auxiliary feedwater pump. However, the finding was determined to be of very low safety significance since (1) except for January 25, 2003, conditions after the procedure change in March 2001 would not have been expected to lower main steam valve house room temperatures below acceptable temperatures, and (2) on January 25, 2003, the two motor driven auxiliary feedwater pumps were operable and performed as expected. Surry personnel tracked corrective actions for this issue under plant issue S-2003-5822.

Inspection Report# : 2003009(pdf)

Significance: Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator No. 3 Bus-Tie Breaker Control Circuit Design Deficiency The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control because emergency diesel generator (EDG) no.

3 could have been overloaded following a concurrent loss-of-offsite power on Units 1 and 2. The licensee has resolved the problem through a modification of the breaker control circuitry.

This finding is greater than minor because it is associated with EDG performance and affects the mitigating systems cornerstone objective. The finding is of very low safety significance because the inspectors determined that the automatically connected loads are less than the 168-hour rating of the EDG.

Inspection Report# : 2003004(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety

2Q/2004 Inspection Findings - Surry 1 Page 3 of 3 Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Dec 05, 2003 Identified By: NRC Item Type: FIN Finding Biennial Problem Identification and Resolution Report The team concluded that Surry personnel were properly identifying problems and entering them into the corrective action program at a threshold that supported safe plant operation. The team did not identify instances where conditions adverse to quality were handled outside the corrective action process. The team further concluded that evaluations were prioritized and completed in a timely fashion consistent with the safety significance of the issue. Cause evaluations generally were found to address technical issues to a depth necessary to identify likely causes. The team identified one finding regarding a less than adequate procedure change evaluation that impacted the reliability of the Unit 1 turbine driven auxiliary feedwater pump. The team found that corrective actions were adequately tracked and completed in a time frame commensurate with their safety significance.

Inspection Report# : 2003009(pdf)

Last modified : September 08, 2004

3Q/2004 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 3Q/2004 Plant Inspection Findings Initiating Events Significance: N/A Apr 28, 2004 Identified By: NRC Item Type: FIN Finding Results of Supplemental Inspection for White Performance Indicator This supplemental inspection was conducted to assess the licensee's evaluation associated with a White performance indicator in the initiating events cornerstone. The Unplanned Scrams per 7,000 Critical Hours Performance Indicator crossed the threshold from Green to White in the third quarter of calendar year 2003. Specifically, the licensee experienced two reactor trips during the first quarter of 2003, one reactor trip during the second quarter of 2003, and one reactor trip in the third quarter of 2003. The first reactor trip, which occurred on January 14, 2003, was a manual trip from approximately 100 percent reactor power due to high temperature and shaft vibration alarms on the C reactor coolant pump. The second reactor trip, which occurred on January 25, 2003, was an automatic trip from approximately 27 percent reactor power due to problems associated with manually controlling steam generator water level. The third reactor trip, which occurred on June 13, 2003, was a manual trip from less than one percent reactor power due to a control rod misalignment. The fourth reactor trip, which occurred on September 18, 2003, was a manual reactor trip from approximately 79 percent reactor power due to inclement weather conditions and a loss of the 1G and 2G buses which supplied power to all the circulating water pumps for both units.

The licensee's problem identification, root cause and extent-of-condition evaluations, and corrective actions for the four reactor trips were adequate. Common cause aspects linking the four reactor trips from a risk perspective were not evident.

Inspection Report# : 2004009(pdf)

Mitigating Systems Significance: Jun 30, 2004 Identified By: NRC Item Type: VIO Violation Alternative Shutdown Capability and Response Procedures Not Adequate to Ensure Safe Shutdown of Unit 1 and 2 A violation of 10 CFR 50, Appendix R, Sections III.L.2.b and III.L.3 was identified, in that, for a severe fire in the Emergency Switchgear and Relay Room Number 1 (Fire Area 3), the licensee's fire response procedures were not effective in assuring a safe shutdown of the Unit 1 reactor. The licensee has revised the affected fire response procedures and is evaluating the need for additional corrective action.

This finding is greater than minor because it was associated with "protection against one of the external factors" attribute. It affected the objective of the Initiating Events cornerstone to limit the likelihood events that challenge critical safety functions as well as affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events.

This degraded condition increased plant risk because, if a severe fire occurred in Fire Area 3, these procedures may not preclude an extended loss of reactor coolant pump seal injection flow and may initiate a reactor coolant pump seal loss of coolant accident which could result in pressurizer level failing to be maintained within the indicating range as required.

This violation was dispositioned as a White finding by NRC Inspection Report 05000280/2004008 and 05000281/2004008, dated September 15, 2004.

Inspection Report# : 2004008(pdf)

Significance: Jan 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation Fire Response Procedures 1-FCA-3.00 And 0-FCA-14.00 Not Adequate To Ensure Safe Shutdown Of Unit 1.

A Green non-cited violation of 10 CFR 50, Appendix R, Sections III.L.2.b and III.L.3, was identified, in that, for a severe fire in the Unit 1 Cable Vault and Tunnel (Fire Area 1), the licensee's alternative shutdown capability may not ensure that the reactor coolant makeup function would be capable of maintaining the reactor coolant level within the level indication of the pressurizer. The licensee has entered this finding into its corrective action program.

3Q/2004 Inspection Findings - Surry 1 Page 2 of 3 This finding is greater than minor because it was associated with "protection against one of the external factors" attribute. It affected the objective of the Initiating Events cornerstone to limit the likelihood events that challenge critical safety functions as well as affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events.

This finding was determined to be of very low safety significance because the likelihood of a severe fire in the service building cable vault (SBCV) or the cable tunnel that could cause a loss of all three Unit 1 charging pumps is very low and a 3-hour rated fire door would prevent a severe fire in the remaining sections of Fire Area 1 from spreading through the cable tunnel to the SBCV.

Inspection Report# : 2003008(pdf)

Significance: Jan 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation Alternate Shutdown Panel Ventilation System Not Independent From Impacts Of A Main Control Room Fire.

A Green non-cited violation was identified for failure to comply with 10 CFR 50, Appendix R, Sections III.G.3.a and III.L.3. Specifically, the shared ventilation system between the main control room (MCR) and the Unit 1 and Unit 2 emergency switchgear and relay rooms (ESGRs),

did not have adequate separation, isolation, or barriers to preclude smoke and toxic gases from being transported to the ESGRs during a fire in the MCR. The alternative shutdown capability for an MCR fire is located in each unit's ESGR, respectively. Consequently, operators may not have the environmental conditions or visibility to safely man and accomplish a successful shutdown of either Unit 1 or Unit 2 from the Auxiliary Shutdown Panels. The licensee has entered this finding into its corrective action program.

This finding is greater than minor because it was associated with the "protection against external factors" attribute and affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events. This finding was determined to be of very low safety significance because heat from a fire, and the natural buoyancy of smoke, will cause the smoke gas layer to accumulate near the ceiling of the MCR (away from the ESGRs), the likelihood of a severe fire in the MCR is low, and the prompt response and actions of the MCR operators and the fire brigade would prevent any fires that start from becoming severe.

Inspection Report# : 2003008(pdf)

Significance: Dec 05, 2003 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater Pump Design Basis not Translated into Procedures The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criteria III (Design Control), in that, a design basis requirement for the Unit 1 auxiliary feedwater pump turbine governor oil viscosity was not correctly translated into a March 2001 procedure revision. The procedure revision failed to require the main steam valve house room temperature to be above that required for minimum vendor specified governor oil viscosity. This non-cited violation contributed to the pump's failure to continue to operate after starting in response to a reactor trip on January 25, 2003.

This finding is greater than minor because it affected the reliability of the Unit 1 turbine driven auxiliary feedwater pump. However, the finding was determined to be of very low safety significance since (1) except for January 25, 2003, conditions after the procedure change in March 2001 would not have been expected to lower main steam valve house room temperatures below acceptable temperatures, and (2) on January 25, 2003, the two motor driven auxiliary feedwater pumps were operable and performed as expected. Surry personnel tracked corrective actions for this issue under plant issue S-2003-5822.

Inspection Report# : 2003009(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement and maintain a respiratory protection program that includes written procedures regarding training of respirator

3Q/2004 Inspection Findings - Surry 1 Page 3 of 3 users in the change out of SCBA air cylinders The inspectors identified a violation of 10 CFR 20.1703(c)(4)(ii) which requires the licensee to implement and maintain a respiratory protection program that includes written procedures regarding training of respirator users. In addition, this was related to the emergency planning standards of 10 CFR 50.47(b) (10). Specifically, procedures were not in place to ensure that all Control Room staff had demonstrated proficiency in changing Self Contained Breathing Apparatus (SCBA) air cylinders during emergencies.

This finding is greater than minor because emergency workers who are required to use respiratory protective equipment are not trained to use that equipment. This finding is of very low safety significance because an adequate number of SCBA qualified plant personnel/staff, which were designated emergency responders, would have been available to respond in the event of an actual emergency.

Inspection Report# : 2004004(pdf)

Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Dec 05, 2003 Identified By: NRC Item Type: FIN Finding Biennial Problem Identification and Resolution Report The team concluded that Surry personnel were properly identifying problems and entering them into the corrective action program at a threshold that supported safe plant operation. The team did not identify instances where conditions adverse to quality were handled outside the corrective action process. The team further concluded that evaluations were prioritized and completed in a timely fashion consistent with the safety significance of the issue. Cause evaluations generally were found to address technical issues to a depth necessary to identify likely causes. The team identified one finding regarding a less than adequate procedure change evaluation that impacted the reliability of the Unit 1 turbine driven auxiliary feedwater pump. The team found that corrective actions were adequately tracked and completed in a time frame commensurate with their safety significance.

Inspection Report# : 2003009(pdf)

Last modified : December 29, 2004

4Q/2004 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 4Q/2004 Plant Inspection Findings Initiating Events Significance: N/A Apr 28, 2004 Identified By: NRC Item Type: FIN Finding Results of Supplemental Inspection for White Performance Indicator This supplemental inspection was conducted to assess the licensee's evaluation associated with a White performance indicator in the initiating events cornerstone. The Unplanned Scrams per 7,000 Critical Hours Performance Indicator crossed the threshold from Green to White in the third quarter of calendar year 2003. Specifically, the licensee experienced two reactor trips during the first quarter of 2003, one reactor trip during the second quarter of 2003, and one reactor trip in the third quarter of 2003. The first reactor trip, which occurred on January 14, 2003, was a manual trip from approximately 100 percent reactor power due to high temperature and shaft vibration alarms on the C reactor coolant pump. The second reactor trip, which occurred on January 25, 2003, was an automatic trip from approximately 27 percent reactor power due to problems associated with manually controlling steam generator water level. The third reactor trip, which occurred on June 13, 2003, was a manual trip from less than one percent reactor power due to a control rod misalignment. The fourth reactor trip, which occurred on September 18, 2003, was a manual reactor trip from approximately 79 percent reactor power due to inclement weather conditions and a loss of the 1G and 2G buses which supplied power to all the circulating water pumps for both units.

The licensee's problem identification, root cause and extent-of-condition evaluations, and corrective actions for the four reactor trips were adequate. Common cause aspects linking the four reactor trips from a risk perspective were not evident.

Inspection Report# : 2004009(pdf)

Mitigating Systems Significance: N/A Dec 10, 2004 Identified By: NRC Item Type: FIN Finding 95002 Supplemental Inspection Resulta for Degraded Mitigating Systems Cornerstone This supplemental inspection was performed by the NRC to assess the licensee's problem identification, root cause evaluation, extent of condition determination, and corrective actions associated with a White performance indicator (PI) and a White inspection finding. These two issues, which were in the Mitigating Systems Cornerstone, placed the performance of Surry Units 1 and 2 in the Degraded Cornerstone Column of the NRC's Action Matrix for the first quarter 2004. The PI, Safety System Unavailability - Emergency AC Power, crossed the threshold from Green to White in the fourth quarter 2001 for both units and remained through the first quarter 2004 for Unit 2, and through the third quarter 2004 for Unit 1. The White PI was evaluated in Supplemental Inspection Report 05000280,281/2002008. The White inspection finding involved Surry fire response procedures that were not effective in ensuring safe shutdown for a fire in Emergency Switchgear and Relay Room Numbers 1 or 2, of Surry Power Station Units 1 and 2 respectively. Specifically, the procedures may not have precluded an extended loss of reactor coolant pump (RCP) seal injection flow, resulting in an RCP seal loss of coolant accident. The performance issue associated with this inspection finding was previously characterized as having low to moderate risk significance (White) in NRC "Final Significance Determination" letter dated September 15, 2004.

During this supplemental inspection, which was performed in accordance with Inspection Procedure 95002, the inspectors utilized the results from Supplemental Inspection Report 05000280,281/2002008 to address the White PI, Safety System Unavailability - Emergency AC Power.

The combined assessment of the White PI and the White inspection finding that resulted in the degraded Mitigating Systems cornerstone is summarized below.

As indicated in Supplemental Inspection Report 05000280,281/2002008, the licensee's formal root cause evaluations (RCE) for the White PI, Safety System Unavailability - Emergency AC Power, was acceptable. The licensee implemented adequate corrective actions to prevent recurrence based upon their RCEs.

The licensee performed a Category 1 RCE, S-2003-1490, to address the fire response procedure finding associated with restoration of seal injection flow to the RCPs. This RCE was considered by the inspectors to be independent and consistent with the prescribed charter. However, the inspectors noted that the licensee's extent of condition reviews lacked thoroughness with regard to the RCE findings. Additionally, the licensee performed Common Cause Evaluation (CCE) S-2004-1504 in January 2004 to assess Surry Power Station Units 1 and 2 performance in the NRC's Reactor Oversight Process. The licensee also performed CCE S-2004-3295 in October 2004 to address the degraded Mitigating Systems cornerstone for Surry Units 1 and 2. The inspectors considered that, although CCE S-2004-3295 did not possess the attributes of an extent of condition evaluation, this CCE determined, through review of various corrective action system documents, that there was a common cause for these White issues. During this 95002 supplemental inspection, the licensee performed more comprehensive extent of condition

4Q/2004 Inspection Findings - Surry 1 Page 2 of 3 related actions through additional reviews of external information programs and processes, and reviews of various management committees' charters/procedures for dispositioning technical concerns. These additional extent of condition and extent of cause related reviews, combined with the efforts in CCE S-2004-3295, were considered to be appropriately focused based on the inspectors' independent extent of condition review.

Although corrective actions appeared to be appropriately prioritized and tracked, the inspectors noted that the licensee was still evaluating long-term corrective action options for resolving the White inspection finding related to restoration of RCP seal injection flow. Consequently, the licensee had not identified all of the corrective actions for this finding and a completion date was not available. Overall, corrective actions related to this White inspection finding adequately addressed compliance restoration and the identified root causes and causal factors. While the inspectors considered that the appropriate root causes were identified by the licensee in RCE S-2003-1490, the contributing cause identified in this RCE was not considered to be the most appropriate. Specifically, the licensee identified that the failure to install Westinghouse (W) high temperature O-rings in the RCP seals in a timely manner was a contributing cause to the failure to revise the Surry Fire Contingency Action (FCA) procedures once the difference between the FCAs and the emergency response guidelines (ERG) was identified. The inspectors noted that the RCE did not recommend any corrective actions for this identified contributing cause. However, the inspectors considered that this contributing cause identified in the RCE was not the most appropriate one. The inspectors considered that the more appropriate contributing cause should have been the unclear responsibilities and inaccurate perception of who had ownership of the FCA procedures. This determination was based on the inspectors' review of RCE S-2003-1490, Potential Problem Report (PPR) 2000-004, and the meeting minutes of the Management Problem Review Team (MPRT) related to PPR 2000-004. The inspectors noted that the licensee had implemented corrective actions to address ownership of the FCA procedures by revising Virginia Power Administrative Procedure (VPAP)-0502, Procedure Process Control.

Inspection Report# : 2004011(pdf)

Significance: Feb 02, 2004 Identified By: NRC Item Type: VIO Violation Alternative Shutdown Capability and Response Procedures Not Adequate to Ensure Safe Shutdown of Unit 1 and 2 A violation of 10 CFR 50, Appendix R, Sections III.L.2.b and III.L.3 was identified, in that, for a severe fire in the Emergency Switchgear and Relay Room Number 1 (Fire Area 3), the licensee's fire response procedures were not effective in assuring a safe shutdown of the Unit 1 reactor. The licensee has revised the affected fire response procedures and is evaluating the need for additional corrective action.

This finding is greater than minor because it was associated with "protection against one of the external factors" attribute. It affected the objective of the Initiating Events cornerstone to limit the likelihood events that challenge critical safety functions as well as affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events.

This degraded condition increased plant risk because, if a severe fire occurred in Fire Area 3, these procedures may not preclude an extended loss of reactor coolant pump seal injection flow and may initiate a reactor coolant pump seal loss of coolant accident which could result in pressurizer level failing to be maintained within the indicating range as required.

This violation was dispositioned as a White finding by NRC Inspection Report 05000280/2004008 and 05000281/2004008, dated September 15, 2004.

Inspection Report# : 2004008(pdf)

Significance: Jan 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation Fire Response Procedures 1-FCA-3.00 And 0-FCA-14.00 Not Adequate To Ensure Safe Shutdown Of Unit 1.

A Green non-cited violation of 10 CFR 50, Appendix R, Sections III.L.2.b and III.L.3, was identified, in that, for a severe fire in the Unit 1 Cable Vault and Tunnel (Fire Area 1), the licensee's alternative shutdown capability may not ensure that the reactor coolant makeup function would be capable of maintaining the reactor coolant level within the level indication of the pressurizer. The licensee has entered this finding into its corrective action program.

This finding is greater than minor because it was associated with "protection against one of the external factors" attribute. It affected the objective of the Initiating Events cornerstone to limit the likelihood events that challenge critical safety functions as well as affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events.

This finding was determined to be of very low safety significance because the likelihood of a severe fire in the service building cable vault (SBCV) or the cable tunnel that could cause a loss of all three Unit 1 charging pumps is very low and a 3-hour rated fire door would prevent a severe fire in the remaining sections of Fire Area 1 from spreading through the cable tunnel to the SBCV.

Inspection Report# : 2003008(pdf)

Significance: Jan 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation

4Q/2004 Inspection Findings - Surry 1 Page 3 of 3 Alternate Shutdown Panel Ventilation System Not Independent From Impacts Of A Main Control Room Fire.

A Green non-cited violation was identified for failure to comply with 10 CFR 50, Appendix R, Sections III.G.3.a and III.L.3. Specifically, the shared ventilation system between the main control room (MCR) and the Unit 1 and Unit 2 emergency switchgear and relay rooms (ESGRs),

did not have adequate separation, isolation, or barriers to preclude smoke and toxic gases from being transported to the ESGRs during a fire in the MCR. The alternative shutdown capability for an MCR fire is located in each unit's ESGR, respectively. Consequently, operators may not have the environmental conditions or visibility to safely man and accomplish a successful shutdown of either Unit 1 or Unit 2 from the Auxiliary Shutdown Panels. The licensee has entered this finding into its corrective action program.

This finding is greater than minor because it was associated with the "protection against external factors" attribute and affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events. This finding was determined to be of very low safety significance because heat from a fire, and the natural buoyancy of smoke, will cause the smoke gas layer to accumulate near the ceiling of the MCR (away from the ESGRs), the likelihood of a severe fire in the MCR is low, and the prompt response and actions of the MCR operators and the fire brigade would prevent any fires that start from becoming severe.

Inspection Report# : 2003008(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement and maintain a respiratory protection program that includes written procedures regarding training of respirator users in the change out of SCBA air cylinders The inspectors identified a violation of 10 CFR 20.1703(c)(4)(ii) which requires the licensee to implement and maintain a respiratory protection program that includes written procedures regarding training of respirator users. In addition, this was related to the emergency planning standards of 10 CFR 50.47(b) (10). Specifically, procedures were not in place to ensure that all Control Room staff had demonstrated proficiency in changing Self Contained Breathing Apparatus (SCBA) air cylinders during emergencies.

This finding is greater than minor because emergency workers who are required to use respiratory protective equipment are not trained to use that equipment. This finding is of very low safety significance because an adequate number of SCBA qualified plant personnel/staff, which were designated emergency responders, would have been available to respond in the event of an actual emergency.

Inspection Report# : 2004004(pdf)

Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Last modified : March 09, 2005

1Q/2005 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 1Q/2005 Plant Inspection Findings Initiating Events Significance: N/A Apr 28, 2004 Identified By: NRC Item Type: FIN Finding Results of Supplemental Inspection for White Performance Indicator This supplemental inspection was conducted to assess the licensee's evaluation associated with a White performance indicator in the initiating events cornerstone. The Unplanned Scrams per 7,000 Critical Hours Performance Indicator crossed the threshold from Green to White in the third quarter of calendar year 2003. Specifically, the licensee experienced two reactor trips during the first quarter of 2003, one reactor trip during the second quarter of 2003, and one reactor trip in the third quarter of 2003. The first reactor trip, which occurred on January 14, 2003, was a manual trip from approximately 100 percent reactor power due to high temperature and shaft vibration alarms on the C reactor coolant pump. The second reactor trip, which occurred on January 25, 2003, was an automatic trip from approximately 27 percent reactor power due to problems associated with manually controlling steam generator water level. The third reactor trip, which occurred on June 13, 2003, was a manual trip from less than one percent reactor power due to a control rod misalignment. The fourth reactor trip, which occurred on September 18, 2003, was a manual reactor trip from approximately 79 percent reactor power due to inclement weather conditions and a loss of the 1G and 2G buses which supplied power to all the circulating water pumps for both units.

The licensee's problem identification, root cause and extent-of-condition evaluations, and corrective actions for the four reactor trips were adequate. Common cause aspects linking the four reactor trips from a risk perspective were not evident.

Inspection Report# : 2004009(pdf)

Mitigating Systems Significance: Mar 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Provide a Power Supply for Turbine Building Flood Instrumentation and Circulating Water Condenser Inlet Valve Logic Which Would be Available Following a Loss of offsite power The inspectors identified a finding in that the turbine building flood control system did not provide adequate protection for all licensing basis flooding scenarios. Specifically, portions of the flooding detection and mitigation circuitry, turbine building flood level detection instrumentation, and circulating water (CW) condenser inlet valve closure logic, would not be available for some flooding scenarios involving a loss of offsite power. The licensee's completed corrective actions include installation of a design change which provides redundant, vital bus powered detection and warning of flooding in the turbine building basement which alarms in the control room.

The finding is greater than minor because it affects the design control attribute of the mitigating systems cornerstone objective. A Phase 3 risk analysis determined that this finding was of very low safety significance. This was primarily due to the low frequency of an earthquake of sufficient magnitude to fail offsite power and the circulating water piping connected to the condenser, but of insufficient magnitude to cause catastrophic failure of the turbine building. (Section 4OA5.2)

Inspection Report# : 2005002(pdf)

Significance: N/A Dec 10, 2004 Identified By: NRC Item Type: FIN Finding 95002 Supplemental Inspection Resulta for Degraded Mitigating Systems Cornerstone This supplemental inspection was performed by the NRC to assess the licensee's problem identification, root cause evaluation, extent of condition determination, and corrective actions associated with a White performance indicator (PI) and a White inspection finding. These two issues, which were in the Mitigating Systems Cornerstone, placed the performance of Surry Units 1 and 2 in the Degraded Cornerstone Column of the NRC's Action Matrix for the first quarter 2004. The PI, Safety System Unavailability - Emergency AC Power, crossed the threshold from Green to White in the fourth quarter 2001 for both units and remained through the first quarter 2004 for Unit 2, and through the third quarter 2004 for Unit 1. The White PI was evaluated in Supplemental Inspection Report 05000280,281/2002008. The White inspection finding involved Surry fire response procedures that were not effective in ensuring safe shutdown for a fire in Emergency Switchgear and Relay Room Numbers 1 or 2, of Surry Power Station Units 1 and 2 respectively. Specifically, the procedures may not have precluded an extended loss of reactor coolant pump (RCP) seal injection flow, resulting in an RCP seal loss of coolant accident. The performance issue associated with this inspection finding was previously characterized as having low to moderate risk significance (White) in NRC "Final Significance Determination" letter dated September 15, 2004.

1Q/2005 Inspection Findings - Surry 1 Page 2 of 3 During this supplemental inspection, which was performed in accordance with Inspection Procedure 95002, the inspectors utilized the results from Supplemental Inspection Report 05000280,281/2002008 to address the White PI, Safety System Unavailability - Emergency AC Power.

The combined assessment of the White PI and the White inspection finding that resulted in the degraded Mitigating Systems cornerstone is summarized below.

As indicated in Supplemental Inspection Report 05000280,281/2002008, the licensee's formal root cause evaluations (RCE) for the White PI, Safety System Unavailability - Emergency AC Power, was acceptable. The licensee implemented adequate corrective actions to prevent recurrence based upon their RCEs.

The licensee performed a Category 1 RCE, S-2003-1490, to address the fire response procedure finding associated with restoration of seal injection flow to the RCPs. This RCE was considered by the inspectors to be independent and consistent with the prescribed charter. However, the inspectors noted that the licensee's extent of condition reviews lacked thoroughness with regard to the RCE findings. Additionally, the licensee performed Common Cause Evaluation (CCE) S-2004-1504 in January 2004 to assess Surry Power Station Units 1 and 2 performance in the NRC's Reactor Oversight Process. The licensee also performed CCE S-2004-3295 in October 2004 to address the degraded Mitigating Systems cornerstone for Surry Units 1 and 2. The inspectors considered that, although CCE S-2004-3295 did not possess the attributes of an extent of condition evaluation, this CCE determined, through review of various corrective action system documents, that there was a common cause for these White issues. During this 95002 supplemental inspection, the licensee performed more comprehensive extent of condition related actions through additional reviews of external information programs and processes, and reviews of various management committees' charters/procedures for dispositioning technical concerns. These additional extent of condition and extent of cause related reviews, combined with the efforts in CCE S-2004-3295, were considered to be appropriately focused based on the inspectors' independent extent of condition review.

Although corrective actions appeared to be appropriately prioritized and tracked, the inspectors noted that the licensee was still evaluating long-term corrective action options for resolving the White inspection finding related to restoration of RCP seal injection flow. Consequently, the licensee had not identified all of the corrective actions for this finding and a completion date was not available. Overall, corrective actions related to this White inspection finding adequately addressed compliance restoration and the identified root causes and causal factors. While the inspectors considered that the appropriate root causes were identified by the licensee in RCE S-2003-1490, the contributing cause identified in this RCE was not considered to be the most appropriate. Specifically, the licensee identified that the failure to install Westinghouse (W) high temperature O-rings in the RCP seals in a timely manner was a contributing cause to the failure to revise the Surry Fire Contingency Action (FCA) procedures once the difference between the FCAs and the emergency response guidelines (ERG) was identified. The inspectors noted that the RCE did not recommend any corrective actions for this identified contributing cause. However, the inspectors considered that this contributing cause identified in the RCE was not the most appropriate one. The inspectors considered that the more appropriate contributing cause should have been the unclear responsibilities and inaccurate perception of who had ownership of the FCA procedures. This determination was based on the inspectors' review of RCE S-2003-1490, Potential Problem Report (PPR) 2000-004, and the meeting minutes of the Management Problem Review Team (MPRT) related to PPR 2000-004. The inspectors noted that the licensee had implemented corrective actions to address ownership of the FCA procedures by revising Virginia Power Administrative Procedure (VPAP)-0502, Procedure Process Control.

Inspection Report# : 2004011(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement and maintain a respiratory protection program that includes written procedures regarding training of respirator users in the change out of SCBA air cylinders The inspectors identified a violation of 10 CFR 20.1703(c)(4)(ii) which requires the licensee to implement and maintain a respiratory protection program that includes written procedures regarding training of respirator users. In addition, this was related to the emergency planning standards of 10 CFR 50.47(b) (10). Specifically, procedures were not in place to ensure that all Control Room staff had demonstrated proficiency in changing Self Contained Breathing Apparatus (SCBA) air cylinders during emergencies.

1Q/2005 Inspection Findings - Surry 1 Page 3 of 3 This finding is greater than minor because emergency workers who are required to use respiratory protective equipment are not trained to use that equipment. This finding is of very low safety significance because an adequate number of SCBA qualified plant personnel/staff, which were designated emergency responders, would have been available to respond in the event of an actual emergency.

Inspection Report# : 2004004(pdf)

Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Last modified : June 17, 2005

2Q/2005 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 2Q/2005 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action Resulting in Recurring Thru-wall Leaks on Main Control Room Chillers 4D' and 4E' The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action" for failure to prevent recurrence of a condition adverse to quality. The licensee identified but did not take corrective actions to prevent recurrence of thru-wall leaks in service water related components on main control room chillers 4D' and 4E'. At least 11 thru-wall leaks have occurred between June 1995 and February 2005 without proper corrective actions to address the cause.

The finding was determined to be more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is associated with the equipment performance and design control attributes of the cornerstone. The finding affects the Mitigating Systems Cornerstone function of core decay heat removal and is of low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events. The finding is also related to the cross-cutting area of identification and resolution of problems because corrective actions were not taken to prevent recurrence of the flow accelerated corrosion condition.

Inspection Report# : 2005003(pdf)

Significance: Mar 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Provide a Power Supply for Turbine Building Flood Instrumentation and Circulating Water Condenser Inlet Valve Logic Which Would be Available Following a Loss of offsite power The inspectors identified a finding in that the turbine building flood control system did not provide adequate protection for all licensing basis flooding scenarios. Specifically, portions of the flooding detection and mitigation circuitry, turbine building flood level detection instrumentation, and circulating water (CW) condenser inlet valve closure logic, would not be available for some flooding scenarios involving a loss of offsite power. The licensee's completed corrective actions include installation of a design change which provides redundant, vital bus powered detection and warning of flooding in the turbine building basement which alarms in the control room.

The finding is greater than minor because it affects the design control attribute of the mitigating systems cornerstone objective. A Phase 3 risk analysis determined that this finding was of very low safety significance. This was primarily due to the low frequency of an earthquake of sufficient magnitude to fail offsite power and the circulating water piping connected to the condenser, but of insufficient magnitude to cause catastrophic failure of the turbine building. (Section 4OA5.2)

Inspection Report# : 2005002(pdf)

Significance: N/A Dec 10, 2004 Identified By: NRC Item Type: FIN Finding 95002 Supplemental Inspection Resulta for Degraded Mitigating Systems Cornerstone This supplemental inspection was performed by the NRC to assess the licensee's problem identification, root cause evaluation, extent of condition determination, and corrective actions associated with a White performance indicator (PI) and a White inspection finding. These two issues, which were in the Mitigating Systems Cornerstone, placed the performance of Surry Units 1 and 2 in the Degraded Cornerstone Column of the NRC's Action Matrix for the first quarter 2004. The PI, Safety System Unavailability - Emergency AC Power, crossed the threshold from Green to White in the fourth quarter 2001 for both units and remained through the first quarter 2004 for Unit 2, and through the third quarter 2004 for Unit 1. The White PI was evaluated in Supplemental Inspection Report 05000280,281/2002008. The White inspection finding involved Surry fire response procedures that were not effective in ensuring safe shutdown for a fire in Emergency Switchgear and Relay Room Numbers 1 or 2, of Surry Power Station Units 1 and 2 respectively. Specifically, the procedures may not have precluded an extended loss of reactor coolant pump (RCP) seal injection flow, resulting in an RCP seal loss of coolant accident. The performance issue associated with this inspection finding was previously characterized as having low to moderate risk significance (White) in NRC "Final Significance

2Q/2005 Inspection Findings - Surry 1 Page 2 of 3 Determination" letter dated September 15, 2004.

During this supplemental inspection, which was performed in accordance with Inspection Procedure 95002, the inspectors utilized the results from Supplemental Inspection Report 05000280,281/2002008 to address the White PI, Safety System Unavailability - Emergency AC Power.

The combined assessment of the White PI and the White inspection finding that resulted in the degraded Mitigating Systems cornerstone is summarized below.

As indicated in Supplemental Inspection Report 05000280,281/2002008, the licensee's formal root cause evaluations (RCE) for the White PI, Safety System Unavailability - Emergency AC Power, was acceptable. The licensee implemented adequate corrective actions to prevent recurrence based upon their RCEs.

The licensee performed a Category 1 RCE, S-2003-1490, to address the fire response procedure finding associated with restoration of seal injection flow to the RCPs. This RCE was considered by the inspectors to be independent and consistent with the prescribed charter. However, the inspectors noted that the licensee's extent of condition reviews lacked thoroughness with regard to the RCE findings. Additionally, the licensee performed Common Cause Evaluation (CCE) S-2004-1504 in January 2004 to assess Surry Power Station Units 1 and 2 performance in the NRC's Reactor Oversight Process. The licensee also performed CCE S-2004-3295 in October 2004 to address the degraded Mitigating Systems cornerstone for Surry Units 1 and 2. The inspectors considered that, although CCE S-2004-3295 did not possess the attributes of an extent of condition evaluation, this CCE determined, through review of various corrective action system documents, that there was a common cause for these White issues. During this 95002 supplemental inspection, the licensee performed more comprehensive extent of condition related actions through additional reviews of external information programs and processes, and reviews of various management committees' charters/procedures for dispositioning technical concerns. These additional extent of condition and extent of cause related reviews, combined with the efforts in CCE S-2004-3295, were considered to be appropriately focused based on the inspectors' independent extent of condition review.

Although corrective actions appeared to be appropriately prioritized and tracked, the inspectors noted that the licensee was still evaluating long-term corrective action options for resolving the White inspection finding related to restoration of RCP seal injection flow. Consequently, the licensee had not identified all of the corrective actions for this finding and a completion date was not available. Overall, corrective actions related to this White inspection finding adequately addressed compliance restoration and the identified root causes and causal factors. While the inspectors considered that the appropriate root causes were identified by the licensee in RCE S-2003-1490, the contributing cause identified in this RCE was not considered to be the most appropriate. Specifically, the licensee identified that the failure to install Westinghouse (W) high temperature O-rings in the RCP seals in a timely manner was a contributing cause to the failure to revise the Surry Fire Contingency Action (FCA) procedures once the difference between the FCAs and the emergency response guidelines (ERG) was identified. The inspectors noted that the RCE did not recommend any corrective actions for this identified contributing cause. However, the inspectors considered that this contributing cause identified in the RCE was not the most appropriate one. The inspectors considered that the more appropriate contributing cause should have been the unclear responsibilities and inaccurate perception of who had ownership of the FCA procedures. This determination was based on the inspectors' review of RCE S-2003-1490, Potential Problem Report (PPR) 2000-004, and the meeting minutes of the Management Problem Review Team (MPRT) related to PPR 2000-004. The inspectors noted that the licensee had implemented corrective actions to address ownership of the FCA procedures by revising Virginia Power Administrative Procedure (VPAP)-0502, Procedure Process Control.

Inspection Report# : 2004011(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Sep 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement and maintain a respiratory protection program that includes written procedures regarding training of respirator users in the change out of SCBA air cylinders The inspectors identified a violation of 10 CFR 20.1703(c)(4)(ii) which requires the licensee to implement and maintain a respiratory protection program that includes written procedures regarding training of respirator users. In addition, this was related to the emergency planning standards of 10 CFR 50.47(b) (10). Specifically, procedures were not in place to ensure that all Control Room staff had demonstrated proficiency in changing Self Contained Breathing Apparatus (SCBA) air cylinders during emergencies.

2Q/2005 Inspection Findings - Surry 1 Page 3 of 3 This finding is greater than minor because emergency workers who are required to use respiratory protective equipment are not trained to use that equipment. This finding is of very low safety significance because an adequate number of SCBA qualified plant personnel/staff, which were designated emergency responders, would have been available to respond in the event of an actual emergency.

Inspection Report# : 2004004(pdf)

Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Last modified : August 24, 2005

3Q/2005 Inspection Findings - Surry 1 Page 1 of 4 Surry 1 3Q/2005 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Translate a Design Change into Design Specifications A self-revealing non-cited violation of 10CFR50, Appendix B, Criterion III, Design Control, was identified for failure to correctly translate design changes into design specifications. The licensee developed a design change for the Unit 1 and Unit 2 charging pump lube oil cooler heat exchangers to prevent corrosion related tube failure. The licensee failed to transfer this design change into an applicable procurement specification to procure lube oil cooler heat exchangers with coated tubes.

The finding is determined to be more than minor because it affects the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment and human performance attributes of the cornerstone. The finding was evaluated using Manual Chapter 0609 and determined to be of low safety significance. The finding affects the Mitigating Systems Cornerstone for short term decay heat removal and is of low safety significance because it did not result in the actual loss of a safety system and is not risk significant in response to external events.

Inspection Report# : 2005004(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Degraded Flow Condition on an Emergency Service Water Pump The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly correct a condition adverse to quality. The licensee identified, but did not promptly correct, a degraded flow condition on the A' Emergency Service Water Pump.

The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment performance and human performance attributes of the cornerstone. The finding affects the Mitigating Systems Cornerstone function of core decay heat removal and is of very low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events (seismic, flood, and severe weather). The finding is also related to the cross-cutting area of problem identification and resolution because the cause of the degraded flow condition was not promptly corrected by the licensee.

Inspection Report# : 2005006(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Lubricating Oil Dilution Condition on an Emergency Service Water Pump The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly identify and correct a condition adverse to quality. The licensee identified, but did not promptly correct, a degrading trend in the lubricating oil associated with the B' Emergency Service Water Pump.

The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment performance and human performance attributes of the cornerstone. NRC Inspection Manual Chapter 0609, Appendix A was used to evaluate this finding. Phase 2 Significance Determination Process analyses determined that this finding is of very low safety significance (Green) because only one of the three trains of emergency service water was affected and only one train is required to mitigate the consequences of an accident. The finding is also related to the cross-cutting area of problem identification and resolution because the lubricating oil degradation condition was not promptly identified and corrected by the licensee.

3Q/2005 Inspection Findings - Surry 1 Page 2 of 4 Inspection Report# : 2005006(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct High Moisture Content of the EDG Air Start System The team identified a non-cited violation for the failure to comply with 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to take timely corrective actions from a previous event in which corrosion products from the carbon steel air start system prevented the #2 Emergency Diesel Generator to start.

The event was determined to be more than minor because it affected the Mitigation System Cornerstone and affected the reliability of the emergency power system. The item was determined to be of very low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events (seismic, flood, and severe weather). The finding is also related to the cross-cutting area of problem identification and resolution because the air dryer installation was not implemented in a timely manner Inspection Report# : 2005006(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action Resulting in Recurring Thru-wall Leaks on Main Control Room Chillers 4D' and 4E' The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action" for failure to prevent recurrence of a condition adverse to quality. The licensee identified but did not take corrective actions to prevent recurrence of thru-wall leaks in service water related components on main control room chillers 4D' and 4E'. At least 11 thru-wall leaks have occurred between June 1995 and February 2005 without proper corrective actions to address the cause.

The finding was determined to be more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is associated with the equipment performance and design control attributes of the cornerstone. The finding affects the Mitigating Systems Cornerstone function of core decay heat removal and is of low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events. The finding is also related to the cross-cutting area of identification and resolution of problems because corrective actions were not taken to prevent recurrence of the flow accelerated corrosion condition.

Inspection Report# : 2005003(pdf)

Significance: Mar 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Provide a Power Supply for Turbine Building Flood Instrumentation and Circulating Water Condenser Inlet Valve Logic Which Would be Available Following a Loss of offsite power The inspectors identified a finding in that the turbine building flood control system did not provide adequate protection for all licensing basis flooding scenarios. Specifically, portions of the flooding detection and mitigation circuitry, turbine building flood level detection instrumentation, and circulating water (CW) condenser inlet valve closure logic, would not be available for some flooding scenarios involving a loss of offsite power. The licensee's completed corrective actions include installation of a design change which provides redundant, vital bus powered detection and warning of flooding in the turbine building basement which alarms in the control room.

The finding is greater than minor because it affects the design control attribute of the mitigating systems cornerstone objective. A Phase 3 risk analysis determined that this finding was of very low safety significance. This was primarily due to the low frequency of an earthquake of sufficient magnitude to fail offsite power and the circulating water piping connected to the condenser, but of insufficient magnitude to cause catastrophic failure of the turbine building. (Section 4OA5.2)

Inspection Report# : 2005002(pdf)

Significance: N/A Dec 10, 2004 Identified By: NRC Item Type: FIN Finding 95002 Supplemental Inspection Resulta for Degraded Mitigating Systems Cornerstone This supplemental inspection was performed by the NRC to assess the licensee's problem identification, root cause evaluation, extent of condition determination, and corrective actions associated with a White performance indicator (PI) and a White inspection finding. These two issues, which were in the Mitigating Systems Cornerstone, placed the performance of Surry Units 1 and 2 in the Degraded Cornerstone Column of the NRC's Action Matrix for the first quarter 2004. The PI, Safety System Unavailability - Emergency AC Power, crossed the threshold from Green to White in the fourth quarter 2001 for both units and remained through the first quarter 2004 for Unit 2, and through the third quarter 2004 for Unit 1. The White PI was evaluated in Supplemental Inspection Report 05000280,281/2002008. The White inspection finding involved Surry fire response procedures that were not effective in ensuring safe shutdown for a fire in Emergency Switchgear and Relay Room

3Q/2005 Inspection Findings - Surry 1 Page 3 of 4 Numbers 1 or 2, of Surry Power Station Units 1 and 2 respectively. Specifically, the procedures may not have precluded an extended loss of reactor coolant pump (RCP) seal injection flow, resulting in an RCP seal loss of coolant accident. The performance issue associated with this inspection finding was previously characterized as having low to moderate risk significance (White) in NRC "Final Significance Determination" letter dated September 15, 2004.

During this supplemental inspection, which was performed in accordance with Inspection Procedure 95002, the inspectors utilized the results from Supplemental Inspection Report 05000280,281/2002008 to address the White PI, Safety System Unavailability - Emergency AC Power.

The combined assessment of the White PI and the White inspection finding that resulted in the degraded Mitigating Systems cornerstone is summarized below.

As indicated in Supplemental Inspection Report 05000280,281/2002008, the licensee's formal root cause evaluations (RCE) for the White PI, Safety System Unavailability - Emergency AC Power, was acceptable. The licensee implemented adequate corrective actions to prevent recurrence based upon their RCEs.

The licensee performed a Category 1 RCE, S-2003-1490, to address the fire response procedure finding associated with restoration of seal injection flow to the RCPs. This RCE was considered by the inspectors to be independent and consistent with the prescribed charter. However, the inspectors noted that the licensee's extent of condition reviews lacked thoroughness with regard to the RCE findings. Additionally, the licensee performed Common Cause Evaluation (CCE) S-2004-1504 in January 2004 to assess Surry Power Station Units 1 and 2 performance in the NRC's Reactor Oversight Process. The licensee also performed CCE S-2004-3295 in October 2004 to address the degraded Mitigating Systems cornerstone for Surry Units 1 and 2. The inspectors considered that, although CCE S-2004-3295 did not possess the attributes of an extent of condition evaluation, this CCE determined, through review of various corrective action system documents, that there was a common cause for these White issues. During this 95002 supplemental inspection, the licensee performed more comprehensive extent of condition related actions through additional reviews of external information programs and processes, and reviews of various management committees' charters/procedures for dispositioning technical concerns. These additional extent of condition and extent of cause related reviews, combined with the efforts in CCE S-2004-3295, were considered to be appropriately focused based on the inspectors' independent extent of condition review.

Although corrective actions appeared to be appropriately prioritized and tracked, the inspectors noted that the licensee was still evaluating long-term corrective action options for resolving the White inspection finding related to restoration of RCP seal injection flow. Consequently, the licensee had not identified all of the corrective actions for this finding and a completion date was not available. Overall, corrective actions related to this White inspection finding adequately addressed compliance restoration and the identified root causes and causal factors. While the inspectors considered that the appropriate root causes were identified by the licensee in RCE S-2003-1490, the contributing cause identified in this RCE was not considered to be the most appropriate. Specifically, the licensee identified that the failure to install Westinghouse (W) high temperature O-rings in the RCP seals in a timely manner was a contributing cause to the failure to revise the Surry Fire Contingency Action (FCA) procedures once the difference between the FCAs and the emergency response guidelines (ERG) was identified. The inspectors noted that the RCE did not recommend any corrective actions for this identified contributing cause. However, the inspectors considered that this contributing cause identified in the RCE was not the most appropriate one. The inspectors considered that the more appropriate contributing cause should have been the unclear responsibilities and inaccurate perception of who had ownership of the FCA procedures. This determination was based on the inspectors' review of RCE S-2003-1490, Potential Problem Report (PPR) 2000-004, and the meeting minutes of the Management Problem Review Team (MPRT) related to PPR 2000-004. The inspectors noted that the licensee had implemented corrective actions to address ownership of the FCA procedures by revising Virginia Power Administrative Procedure (VPAP)-0502, Procedure Process Control.

Inspection Report# : 2004011(pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety

3Q/2005 Inspection Findings - Surry 1 Page 4 of 4 Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Aug 26, 2005 Identified By: NRC Item Type: FIN Finding Results of Problem Identification and Resolution Inspection The team concluded that, in general, problems were properly identified, evaluated, and corrected. The licensee was effective at identifying problems, issues were prioritized, evaluated appropriately, and dispositioned in a timely fashion. Evaluations of significant problems were in general, of sufficient depth to determine the likely root or apparent causes, as well as, address the potential extent of the circumstances contributing to the problem and provide a clear basis to establish corrective actions. Corrective actions that addressed the causes of problems were generally identified and implemented. A recent licensee self-assessment identified several areas of improvement. Numerous corrective actions have been implemented as well as planned to address issues raised during the recent self-assessment. Significant changes to address issues, such as extent of condition review, ensure corrective actions match what was expected, and manage number of action items stemming from Plant Issues (PIs), are underway or planned. The team observed the corrective action review board as well as the Plant Issues Review Team (PIRT) and noted improvement in the quality of the resolution of PIs. Reviews of sampled operating experience information were comprehensive. Previous noncompliance issues documented as non-cited violations were properly tracked and resolved via the corrective action program. Based on discussions with plant personnel and the low threshold for items entered in the corrective action program database, the team concluded that workers at the site were free to raise safety concerns to their management.

Inspection Report# : 2005006(pdf)

Last modified : November 30, 2005

4Q/2005 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 4Q/2005 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Translate a Design Change into Design Specifications A self-revealing non-cited violation of 10CFR50, Appendix B, Criterion III, Design Control, was identified for failure to correctly translate design changes into design specifications. The licensee developed a design change for the Unit 1 and Unit 2 charging pump lube oil cooler heat exchangers to prevent corrosion related tube failure. The licensee failed to transfer this design change into an applicable procurement specification to procure lube oil cooler heat exchangers with coated tubes.

The finding is determined to be more than minor because it affects the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment and human performance attributes of the cornerstone. The finding was evaluated using Manual Chapter 0609 and determined to be of low safety significance. The finding affects the Mitigating Systems Cornerstone for short term decay heat removal and is of low safety significance because it did not result in the actual loss of a safety system and is not risk significant in response to external events.

Inspection Report# : 2005004(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Degraded Flow Condition on an Emergency Service Water Pump The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly correct a condition adverse to quality. The licensee identified, but did not promptly correct, a degraded flow condition on the A' Emergency Service Water Pump.

The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment performance and human performance attributes of the cornerstone. The finding affects the Mitigating Systems Cornerstone function of core decay heat removal and is of very low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events (seismic, flood, and severe weather). The finding is also related to the cross-cutting area of problem identification and resolution because the cause of the degraded flow condition was not promptly corrected by the licensee.

Inspection Report# : 2005006(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Lubricating Oil Dilution Condition on an Emergency Service Water Pump The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly identify and correct a condition adverse to quality. The licensee identified, but did not promptly correct, a degrading trend in the lubricating oil associated with the B' Emergency Service Water Pump.

The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment performance and human performance attributes of the cornerstone. NRC Inspection Manual Chapter 0609, Appendix A was used to evaluate this finding. Phase 2 Significance Determination Process analyses determined that this finding is of very low safety significance (Green) because only one of the three trains of emergency service water was affected and only one train is required to mitigate the consequences of an accident. The finding is also related to the cross-cutting area of problem identification and resolution because the lubricating oil degradation condition was not promptly identified and corrected by the licensee.

4Q/2005 Inspection Findings - Surry 1 Page 2 of 3 Inspection Report# : 2005006(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct High Moisture Content of the EDG Air Start System The team identified a non-cited violation for the failure to comply with 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to take timely corrective actions from a previous event in which corrosion products from the carbon steel air start system prevented the #2 Emergency Diesel Generator to start.

The event was determined to be more than minor because it affected the Mitigation System Cornerstone and affected the reliability of the emergency power system. The item was determined to be of very low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events (seismic, flood, and severe weather). The finding is also related to the cross-cutting area of problem identification and resolution because the air dryer installation was not implemented in a timely manner Inspection Report# : 2005006(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action Resulting in Recurring Thru-wall Leaks on Main Control Room Chillers 4D' and 4E' The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action" for failure to prevent recurrence of a condition adverse to quality. The licensee identified but did not take corrective actions to prevent recurrence of thru-wall leaks in service water related components on main control room chillers 4D' and 4E'. At least 11 thru-wall leaks have occurred between June 1995 and February 2005 without proper corrective actions to address the cause.

The finding was determined to be more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is associated with the equipment performance and design control attributes of the cornerstone. The finding affects the Mitigating Systems Cornerstone function of core decay heat removal and is of low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events. The finding is also related to the cross-cutting area of identification and resolution of problems because corrective actions were not taken to prevent recurrence of the flow accelerated corrosion condition.

Inspection Report# : 2005003(pdf)

Significance: Mar 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Provide a Power Supply for Turbine Building Flood Instrumentation and Circulating Water Condenser Inlet Valve Logic Which Would be Available Following a Loss of offsite power The inspectors identified a finding in that the turbine building flood control system did not provide adequate protection for all licensing basis flooding scenarios. Specifically, portions of the flooding detection and mitigation circuitry, turbine building flood level detection instrumentation, and circulating water (CW) condenser inlet valve closure logic, would not be available for some flooding scenarios involving a loss of offsite power. The licensee's completed corrective actions include installation of a design change which provides redundant, vital bus powered detection and warning of flooding in the turbine building basement which alarms in the control room.

The finding is greater than minor because it affects the design control attribute of the mitigating systems cornerstone objective. A Phase 3 risk analysis determined that this finding was of very low safety significance. This was primarily due to the low frequency of an earthquake of sufficient magnitude to fail offsite power and the circulating water piping connected to the condenser, but of insufficient magnitude to cause catastrophic failure of the turbine building. (Section 4OA5.2)

Inspection Report# : 2005002(pdf)

Barrier Integrity Emergency Preparedness

4Q/2005 Inspection Findings - Surry 1 Page 3 of 3 Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Aug 26, 2005 Identified By: NRC Item Type: FIN Finding Results of Problem Identification and Resolution Inspection The team concluded that, in general, problems were properly identified, evaluated, and corrected. The licensee was effective at identifying problems, issues were prioritized, evaluated appropriately, and dispositioned in a timely fashion. Evaluations of significant problems were in general, of sufficient depth to determine the likely root or apparent causes, as well as, address the potential extent of the circumstances contributing to the problem and provide a clear basis to establish corrective actions. Corrective actions that addressed the causes of problems were generally identified and implemented. A recent licensee self-assessment identified several areas of improvement. Numerous corrective actions have been implemented as well as planned to address issues raised during the recent self-assessment. Significant changes to address issues, such as extent of condition review, ensure corrective actions match what was expected, and manage number of action items stemming from Plant Issues (PIs), are underway or planned. The team observed the corrective action review board as well as the Plant Issues Review Team (PIRT) and noted improvement in the quality of the resolution of PIs. Reviews of sampled operating experience information were comprehensive. Previous noncompliance issues documented as non-cited violations were properly tracked and resolved via the corrective action program. Based on discussions with plant personnel and the low threshold for items entered in the corrective action program database, the team concluded that workers at the site were free to raise safety concerns to their management.

Inspection Report# : 2005006(pdf)

Last modified : March 03, 2006

1Q/2006 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 1Q/2006 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Feb 10, 2006 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative ECA Procedure Setpoint for Operator Action to Secure LHSI and HHSI Pumps on Low RWST Level The team identified a Green, non-cited violation (NCV) of Technical Specification 6.4.A.3, Unit Operating Procedures and Programs, for a non-conservative emergency contingency action (ECA) procedure setpoint regarding low Refueling Water Storage Tank (RWST) level.

Specifically, the licensee failed to adequately address the potential for vortexing at low RWST levels into the determination of the RWST level for operator action to secure low head safety injection and high head safety injection pumps drawing suction from the RWST, in Procedures 1,2-ECA-1.1, Loss of Emergency Coolant Recirculation, Rev. 23. When the NRC notified the licensee of this condition, the licensee entered it into the corrective action program, and proceeded to revise the ECA setpoint in the affected procedures.

This finding is greater than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring reliable, available, and capable systems that respond to initiating events to prevent undesirable consequences. This finding is of very low safety significance because no loss of safety function occurred and operators have been trained to identify loss of pump suction. This finding has been entered into the licensees corrective action program as PIlant Issue S-2006-0334.

Inspection Report# : 2006006(pdf)

Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Translate a Design Change into Design Specifications A self-revealing non-cited violation of 10CFR50, Appendix B, Criterion III, Design Control, was identified for failure to correctly translate design changes into design specifications. The licensee developed a design change for the Unit 1 and Unit 2 charging pump lube oil cooler heat exchangers to prevent corrosion related tube failure. The licensee failed to transfer this design change into an applicable procurement specification to procure lube oil cooler heat exchangers with coated tubes.

The finding is determined to be more than minor because it affects the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment and human performance attributes of the cornerstone. The finding was evaluated using Manual Chapter 0609 and determined to be of low safety significance. The finding affects the Mitigating Systems Cornerstone for short term decay heat removal and is of low safety significance because it did not result in the actual loss of a safety system and is not risk significant in response to external events.

Inspection Report# : 2005004(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Degraded Flow Condition on an Emergency Service Water Pump The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly correct a condition adverse to quality. The licensee identified, but did not promptly correct, a degraded flow condition on the A' Emergency Service Water Pump.

The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment performance and human performance attributes of the cornerstone. The finding affects the Mitigating Systems Cornerstone function of core decay heat removal and is of very low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events (seismic, flood, and severe weather). The finding is also related to the cross-cutting area of problem identification and resolution because the cause of the degraded flow condition was not promptly corrected by the licensee.

1Q/2006 Inspection Findings - Surry 1 Page 2 of 3 Inspection Report# : 2005006(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Lubricating Oil Dilution Condition on an Emergency Service Water Pump The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly identify and correct a condition adverse to quality. The licensee identified, but did not promptly correct, a degrading trend in the lubricating oil associated with the B' Emergency Service Water Pump.

The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment performance and human performance attributes of the cornerstone. NRC Inspection Manual Chapter 0609, Appendix A was used to evaluate this finding. Phase 2 Significance Determination Process analyses determined that this finding is of very low safety significance (Green) because only one of the three trains of emergency service water was affected and only one train is required to mitigate the consequences of an accident. The finding is also related to the cross-cutting area of problem identification and resolution because the lubricating oil degradation condition was not promptly identified and corrected by the licensee.

Inspection Report# : 2005006(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct High Moisture Content of the EDG Air Start System The team identified a non-cited violation for the failure to comply with 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to take timely corrective actions from a previous event in which corrosion products from the carbon steel air start system prevented the #2 Emergency Diesel Generator to start.

The event was determined to be more than minor because it affected the Mitigation System Cornerstone and affected the reliability of the emergency power system. The item was determined to be of very low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events (seismic, flood, and severe weather). The finding is also related to the cross-cutting area of problem identification and resolution because the air dryer installation was not implemented in a timely manner Inspection Report# : 2005006(pdf)

Significance: Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action Resulting in Recurring Thru-wall Leaks on Main Control Room Chillers 4D' and 4E' The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action" for failure to prevent recurrence of a condition adverse to quality. The licensee identified but did not take corrective actions to prevent recurrence of thru-wall leaks in service water related components on main control room chillers 4D' and 4E'. At least 11 thru-wall leaks have occurred between June 1995 and February 2005 without proper corrective actions to address the cause.

The finding was determined to be more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is associated with the equipment performance and design control attributes of the cornerstone. The finding affects the Mitigating Systems Cornerstone function of core decay heat removal and is of low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events. The finding is also related to the cross-cutting area of identification and resolution of problems because corrective actions were not taken to prevent recurrence of the flow accelerated corrosion condition.

Inspection Report# : 2005003(pdf)

Barrier Integrity Emergency Preparedness

1Q/2006 Inspection Findings - Surry 1 Page 3 of 3 Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Aug 26, 2005 Identified By: NRC Item Type: FIN Finding Results of Problem Identification and Resolution Inspection The team concluded that, in general, problems were properly identified, evaluated, and corrected. The licensee was effective at identifying problems, issues were prioritized, evaluated appropriately, and dispositioned in a timely fashion. Evaluations of significant problems were in general, of sufficient depth to determine the likely root or apparent causes, as well as, address the potential extent of the circumstances contributing to the problem and provide a clear basis to establish corrective actions. Corrective actions that addressed the causes of problems were generally identified and implemented. A recent licensee self-assessment identified several areas of improvement. Numerous corrective actions have been implemented as well as planned to address issues raised during the recent self-assessment. Significant changes to address issues, such as extent of condition review, ensure corrective actions match what was expected, and manage number of action items stemming from Plant Issues (PIs), are underway or planned. The team observed the corrective action review board as well as the Plant Issues Review Team (PIRT) and noted improvement in the quality of the resolution of PIs. Reviews of sampled operating experience information were comprehensive. Previous noncompliance issues documented as non-cited violations were properly tracked and resolved via the corrective action program. Based on discussions with plant personnel and the low threshold for items entered in the corrective action program database, the team concluded that workers at the site were free to raise safety concerns to their management.

Inspection Report# : 2005006(pdf)

Last modified : May 25, 2006

2Q/2006 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 2Q/2006 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Feb 10, 2006 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative ECA Procedure Setpoint for Operator Action to Secure LHSI and HHSI Pumps on Low RWST Level The team identified a Green, non-cited violation (NCV) of Technical Specification 6.4.A.3, Unit Operating Procedures and Programs, for a non-conservative emergency contingency action (ECA) procedure setpoint regarding low Refueling Water Storage Tank (RWST) level. Specifically, the licensee failed to adequately address the potential for vortexing at low RWST levels into the determination of the RWST level for operator action to secure low head safety injection and high head safety injection pumps drawing suction from the RWST, in Procedures 1,2-ECA-1.1, Loss of Emergency Coolant Recirculation, Rev. 23. When the NRC notified the licensee of this condition, the licensee entered it into the corrective action program, and proceeded to revise the ECA setpoint in the affected procedures.

This finding is greater than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring reliable, available, and capable systems that respond to initiating events to prevent undesirable consequences.

This finding is of very low safety significance because no loss of safety function occurred and operators have been trained to identify loss of pump suction. This finding has been entered into the licensees corrective action program as PIlant Issue S-2006-0334.

Inspection Report# : 2006006(pdf)

Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Translate a Design Change into Design Specifications A self-revealing non-cited violation of 10CFR50, Appendix B, Criterion III, Design Control, was identified for failure to correctly translate design changes into design specifications. The licensee developed a design change for the Unit 1 and Unit 2 charging pump lube oil cooler heat exchangers to prevent corrosion related tube failure. The licensee failed to transfer this design change into an applicable procurement specification to procure lube oil cooler heat exchangers with coated tubes.

The finding is determined to be more than minor because it affects the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment and human performance attributes of the cornerstone. The finding was evaluated using Manual Chapter 0609 and determined to be of low safety significance. The finding affects the Mitigating Systems Cornerstone for short term decay heat removal and is of low safety significance because it did not result in the actual loss of a safety system and is not risk significant in response to external events.

Inspection Report# : 2005004(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Degraded Flow Condition on an Emergency Service Water Pump The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly correct a condition adverse to quality. The licensee identified, but did not promptly correct, a degraded flow condition on the A' Emergency Service Water Pump.

The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment performance and human performance attributes of the cornerstone. The finding affects the Mitigating Systems Cornerstone function of core decay heat removal and is of very low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events (seismic, flood, and severe weather). The finding is also related to the cross-cutting area of problem identification and resolution because the cause of the degraded flow condition was not promptly corrected by the licensee.

Inspection Report# : 2005006(pdf)

2Q/2006 Inspection Findings - Surry 1 Page 2 of 3 Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Lubricating Oil Dilution Condition on an Emergency Service Water Pump The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly identify and correct a condition adverse to quality. The licensee identified, but did not promptly correct, a degrading trend in the lubricating oil associated with the B' Emergency Service Water Pump.

The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was associated with the equipment performance and human performance attributes of the cornerstone. NRC Inspection Manual Chapter 0609, Appendix A was used to evaluate this finding. Phase 2 Significance Determination Process analyses determined that this finding is of very low safety significance (Green) because only one of the three trains of emergency service water was affected and only one train is required to mitigate the consequences of an accident. The finding is also related to the cross-cutting area of problem identification and resolution because the lubricating oil degradation condition was not promptly identified and corrected by the licensee.

Inspection Report# : 2005006(pdf)

Significance: Aug 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct High Moisture Content of the EDG Air Start System The team identified a non-cited violation for the failure to comply with 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to take timely corrective actions from a previous event in which corrosion products from the carbon steel air start system prevented the #2 Emergency Diesel Generator to start.

The event was determined to be more than minor because it affected the Mitigation System Cornerstone and affected the reliability of the emergency power system. The item was determined to be of very low safety significance (Green) because it did not result in the loss of a safety function of a single train for greater than the Technical Specification allowed outage time and is not risk significant in response to external events (seismic, flood, and severe weather). The finding is also related to the cross-cutting area of problem identification and resolution because the air dryer installation was not implemented in a timely manner Inspection Report# : 2005006(pdf)

Barrier Integrity Emergency Preparedness Significance: Mar 29, 2006 Identified By: NRC Item Type: VIO Violation Failure of Exercise Critique to identify a RSPS weakness as a DEP PI opportunity Failure The NRC identified an apparent violation (AV) for failure of the licensees exercise critique process to properly identify a weakness associated with a risk-significant planning standard (RSPS) that was determined to be a Drill/Exercise Performance (DEP) Performance Indicator (PI) opportunity failure during a full-scale exercise. The AV is associated with emergency preparedness planning standards 10 CFR 50.47(b)(14) and 10 CFR 50.47 (b)(4), and the requirements of 10 CFR 50, Appendix E, IV.F.2.g. This finding was not entered into the licensees corrective action program.

The failure of the licensees exercise critique process was a performance deficiency. This finding was greater than minor because it was associated with the Emergency Preparedness Cornerstone. The finding affects the associated cornerstone objective to ensure that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was an identified weakness that demonstrated a level of performance that could preclude effective implementation of the Emergency Plan in an actual emergency. This finding was also determined to potentially have greater significance because the licensees exercise critique process failed to properly identify a weakness associated with a RSPS that was determined to be a DEP PI opportunity failure during a full-scale exercise.

NRC inspection report 05000280, 281/2006010, issued July 25, 2006, closed the apparent violation to a violation with a final significance of White for both Units 1 and 2. The violation, designated as 05000280, 281/2006010-01, is listed below.

10 CFR 50.47(b)(4) requires, in part, that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

2Q/2006 Inspection Findings - Surry 1 Page 3 of 3 10 CFR 50.47(b)(14) requires, in part, that periodic exercises be conducted to evaluate major portions of emergency response capabilities and deficiencies identified as a result of exercises be corrected.

10 CFR Part 50, Appendix E, Section IV.F.2.g, requires that all training, including exercises, shall provide for formal critiques in order to identify weak or deficient areas that need correction. Any weaknesses or deficiencies that are identified shall be corrected.

Contrary to the above, the licensees formal critique of an emergency preparedness exercise conducted on February 7, 2006, failed to identify weak or deficient areas. Specifically, the exercise critique failed to identify that the Station Emergency Managers Site Area Emergency event classification was an inaccurate classification.

Inspection Report# : 2006008(pdf)

Inspection Report# : 2006010(pdf)

Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Significance: N/A Aug 26, 2005 Identified By: NRC Item Type: FIN Finding Results of Problem Identification and Resolution Inspection The team concluded that, in general, problems were properly identified, evaluated, and corrected. The licensee was effective at identifying problems, issues were prioritized, evaluated appropriately, and dispositioned in a timely fashion. Evaluations of significant problems were in general, of sufficient depth to determine the likely root or apparent causes, as well as, address the potential extent of the circumstances contributing to the problem and provide a clear basis to establish corrective actions. Corrective actions that addressed the causes of problems were generally identified and implemented. A recent licensee self-assessment identified several areas of improvement. Numerous corrective actions have been implemented as well as planned to address issues raised during the recent self-assessment. Significant changes to address issues, such as extent of condition review, ensure corrective actions match what was expected, and manage number of action items stemming from Plant Issues (PIs), are underway or planned. The team observed the corrective action review board as well as the Plant Issues Review Team (PIRT) and noted improvement in the quality of the resolution of PIs. Reviews of sampled operating experience information were comprehensive. Previous noncompliance issues documented as non-cited violations were properly tracked and resolved via the corrective action program. Based on discussions with plant personnel and the low threshold for items entered in the corrective action program database, the team concluded that workers at the site were free to raise safety concerns to their management.

Inspection Report# : 2005006(pdf)

Last modified : August 25, 2006

3Q/2006 Inspection Findings - Surry 1 Page 1 of 3 Surry 1 3Q/2006 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Removal of Damper Motor Operators From CO2 System in Normal Switchgear Rooms The team identified a non-cited violation of Operating License Condition 3.I for removing the automatic feature of ventilation dampers which degraded the fixed gaseous suppression system in the normal switchgear room at both units by allowing carbon dioxide to flow out should the manual operated dampers be in the open position.

The finding is more than minor because it was associated with the reactor safety, mitigating systems cornerstone attribute of protection against external factors, i.e. fire, and it affected the objective of ensuring reliability and capability of systems that respond to initiating events. The finding is of very low safety significance because the frequency of fires potentially challenging mitigating systems was relatively low and multiple trains of shutdown equipment would be available.

Inspection Report# : 2006009(pdf)

Significance: Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Post-Fire Safe Shutdown During a Fire in Mechanical Equipment Room 3 The team identified a non-cited violation of Technical Specification 6.4.E for failure to provide an adequate post-fire safe shutdown procedure. Procedure 0-FCA-7.00, Rev. 10, failed to ensure that a source of water would be aligned to the suction of the charging pump service water pumps during a severe fire in Mechanical Equipment Room 3. Consequently, all charging pumps of both units could have no service water cooling resulting in pump overheating and failure.

The finding is greater than minor because it affected the objective of the mitigating system cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events. Since the procedure had been in place for less than one month and during that time a source of water could have been aligned, this finding is of very low safety significance.

Inspection Report# : 2006009(pdf)

Significance: Feb 10, 2006 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative ECA Procedure Setpoint for Operator Action to Secure LHSI and HHSI Pumps on Low RWST Level The team identified a Green, non-cited violation (NCV) of Technical Specification 6.4.A.3, Unit Operating Procedures and Programs, for a non-conservative emergency contingency action (ECA) procedure setpoint regarding low Refueling Water Storage Tank (RWST) level. Specifically, the licensee failed to adequately address the potential for vortexing at low RWST levels into the determination of the RWST level for operator action to secure low head safety injection and high head safety injection pumps drawing suction from the RWST, in Procedures 1,2-ECA-1.1, Loss of Emergency Coolant Recirculation, Rev. 23. When the NRC notified the licensee of this condition, the licensee entered it into the corrective action program, and proceeded to revise the ECA setpoint in the affected procedures.

3Q/2006 Inspection Findings - Surry 1 Page 2 of 3 This finding is greater than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring reliable, available, and capable systems that respond to initiating events to prevent undesirable consequences. This finding is of very low safety significance because no loss of safety function occurred and operators have been trained to identify loss of pump suction. This finding has been entered into the licensees corrective action program as PIlant Issue S-2006-0334.

Inspection Report# : 2006006(pdf)

Barrier Integrity Emergency Preparedness Significance: Mar 29, 2006 Identified By: NRC Item Type: VIO Violation Failure of Exercise Critique to identify a RSPS weakness as a DEP PI opportunity Failure The NRC identified an apparent violation (AV) for failure of the licensees exercise critique process to properly identify a weakness associated with a risk-significant planning standard (RSPS) that was determined to be a Drill/Exercise Performance (DEP) Performance Indicator (PI) opportunity failure during a full-scale exercise. The AV is associated with emergency preparedness planning standards 10 CFR 50.47(b)(14) and 10 CFR 50.47(b)(4), and the requirements of 10 CFR 50, Appendix E, IV.F.2.g. This finding was not entered into the licensees corrective action program.

The failure of the licensees exercise critique process was a performance deficiency. This finding was greater than minor because it was associated with the Emergency Preparedness Cornerstone. The finding affects the associated cornerstone objective to ensure that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was an identified weakness that demonstrated a level of performance that could preclude effective implementation of the Emergency Plan in an actual emergency. This finding was also determined to potentially have greater significance because the licensees exercise critique process failed to properly identify a weakness associated with a RSPS that was determined to be a DEP PI opportunity failure during a full-scale exercise.

NRC inspection report 05000280, 281/2006010, issued July 25, 2006, closed the apparent violation to a violation with a final significance of White for both Units 1 and 2. The violation, designated as 05000280, 281/2006010-01, is listed below.

10 CFR 50.47(b)(4) requires, in part, that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

10 CFR 50.47(b)(14) requires, in part, that periodic exercises be conducted to evaluate major portions of emergency response capabilities and deficiencies identified as a result of exercises be corrected.

10 CFR Part 50, Appendix E, Section IV.F.2.g, requires that all training, including exercises, shall provide for formal critiques in order to identify weak or deficient areas that need correction. Any weaknesses or deficiencies that are identified shall be corrected.

Contrary to the above, the licensees formal critique of an emergency preparedness exercise conducted on February 7, 2006, failed to identify weak or deficient areas. Specifically, the exercise critique failed to identify that the Station Emergency Managers Site Area Emergency event classification was an inaccurate classification.

Inspection Report# : 2006008(pdf)

3Q/2006 Inspection Findings - Surry 1 Page 3 of 3 Inspection Report# : 2006010(pdf)

Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous Last modified : December 21, 2006

4Q/2006 Inspection Findings - Surry 1 Page 1 of 4 Surry 1 4Q/2006 Plant Inspection Findings Initiating Events Mitigating Systems Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Proceduralized Departures from TS The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments.

Specifically, the licensee implemented proceduralized departures from the approved station technical specifications (TS) without the required NRC approval in procedures AP-13.0, Turbine Building Flooding, revision 13, and FCA 6.01, Uncontrollable Turbine Building Flooding, revision 2.

This finding was evaluated using traditional enforcement since it impacted or impeded the regulatory process in that the licensee improperly used the 10 CFR 50.59, Changes, Tests, and Experiments, process to incorporate operator actions inconsistent with the TS. This finding was of more than minor safety significance because the procedure changes improperly bypassed the required NRC review and approval prior to implementation. The unapproved procedural actions would only be involved at the end of a very rare accident sequence. Given the time during the accident sequence in which these actions were to be accomplished, the actions were not a determent to core damage. Therefore, the violation was of very low safety significance. The finding is identified as Severity Level IV because the noncompliance is not considered to be of more than very low significance based on risk.

Inspection Report# : 2006005 (pdf)

Significance: Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Removal of Damper Motor Operators From CO2 System in Normal Switchgear Rooms The team identified a non-cited violation of Operating License Condition 3.I for removing the automatic feature of ventilation dampers which degraded the fixed gaseous suppression system in the normal switchgear room at both units by allowing carbon dioxide to flow out should the manual operated dampers be in the open position.

The finding is more than minor because it was associated with the reactor safety, mitigating systems cornerstone attribute of protection against external factors, i.e. fire, and it affected the objective of ensuring reliability and capability of systems that respond to initiating events. The finding is of very low safety significance because the frequency of fires potentially challenging mitigating systems was relatively low and multiple trains of shutdown equipment would be available.

Inspection Report# : 2006009 (pdf)

Significance: Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Post-Fire Safe Shutdown During a Fire in Mechanical Equipment Room 3 The team identified a non-cited violation of Technical Specification 6.4.E for failure to provide an adequate post-fire safe shutdown procedure. Procedure 0-FCA-7.00, Rev. 10, failed to ensure that a source of water would be aligned to the suction of the charging pump service water pumps during a severe fire in Mechanical Equipment Room 3. Consequently, all charging pumps of both units could have no service water cooling resulting in pump overheating and failure.

4Q/2006 Inspection Findings - Surry 1 Page 2 of 4 The finding is greater than minor because it affected the objective of the mitigating system cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events. Since the procedure had been in place for less than one month and during that time a source of water could have been aligned, this finding is of very low safety significance.

Inspection Report# : 2006009 (pdf)

Significance: Feb 10, 2006 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative ECA Procedure Setpoint for Operator Action to Secure LHSI and HHSI Pumps on Low RWST Level The team identified a Green, non-cited violation (NCV) of Technical Specification 6.4.A.3, Unit Operating Procedures and Programs, for a non-conservative emergency contingency action (ECA) procedure setpoint regarding low Refueling Water Storage Tank (RWST) level. Specifically, the licensee failed to adequately address the potential for vortexing at low RWST levels into the determination of the RWST level for operator action to secure low head safety injection and high head safety injection pumps drawing suction from the RWST, in Procedures 1,2-ECA-1.1, Loss of Emergency Coolant Recirculation, Rev. 23. When the NRC notified the licensee of this condition, the licensee entered it into the corrective action program, and proceeded to revise the ECA setpoint in the affected procedures.

This finding is greater than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring reliable, available, and capable systems that respond to initiating events to prevent undesirable consequences. This finding is of very low safety significance because no loss of safety function occurred and operators have been trained to identify loss of pump suction. This finding has been entered into the licensees corrective action program as PIlant Issue S-2006-0334.

Inspection Report# : 2006006 (pdf)

Barrier Integrity Emergency Preparedness Significance: Oct 27, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Activate ERDS Within One Hour of an Alert Declaration A green self-revealing non-cited violation of 10 CFR 50.72(a)(4) was identified. During the October 7, 2006, partial loss of offsite power event, the licensee failed to activate the Emergency Response Data System (ERDS) within one hour of an Alert declaration. The ERDS was not made operable until approximately five and one-half hours after the Alert declaration due to an upgrade to the telephone exchange that had been done seven days prior to the event.

The finding is more than minor due to its impact on the Emergency Preparedness cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency and the related attribute of Emergency Response Organization response. The finding is of very low safety significance (Green) because it involved a failure to implement (in distinction to a failure to meet) an NRC emergency planning standard. The cause of the finding is related to the cross-cutting area of human performance, in that, the licensee failed to reprogram the telephone exchange following a telephone system change which occurred prior to the event. Upon discovery, the licensee immediately reprogrammed the telephone exchange and entered the problem into their corrective action program as condition report CR 002183.

Inspection Report# : 2006011 (pdf)

4Q/2006 Inspection Findings - Surry 1 Page 3 of 4 Significance: Mar 29, 2006 Identified By: NRC Item Type: VIO Violation Failure of Exercise Critique to identify a RSPS weakness as a DEP PI opportunity Failure The NRC identified an apparent violation (AV) for failure of the licensees exercise critique process to properly identify a weakness associated with a risk-significant planning standard (RSPS) that was determined to be a Drill/Exercise Performance (DEP) Performance Indicator (PI) opportunity failure during a full-scale exercise. The AV is associated with emergency preparedness planning standards 10 CFR 50.47(b)(14) and 10 CFR 50.47(b)(4), and the requirements of 10 CFR 50, Appendix E, IV.F.2.g. This finding was not entered into the licensees corrective action program.

The failure of the licensees exercise critique process was a performance deficiency. This finding was greater than minor because it was associated with the Emergency Preparedness Cornerstone. The finding affects the associated cornerstone objective to ensure that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was an identified weakness that demonstrated a level of performance that could preclude effective implementation of the Emergency Plan in an actual emergency. This finding was also determined to potentially have greater significance because the licensees exercise critique process failed to properly identify a weakness associated with a RSPS that was determined to be a DEP PI opportunity failure during a full-scale exercise.

NRC inspection report 05000280, 281/2006010, issued July 25, 2006, closed the apparent violation to a violation with a final significance of White for both Units 1 and 2. The violation, designated as 05000280, 281/2006010-01, is listed below.

10 CFR 50.47(b)(4) requires, in part, that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

10 CFR 50.47(b)(14) requires, in part, that periodic exercises be conducted to evaluate major portions of emergency response capabilities and deficiencies identified as a result of exercises be corrected.

10 CFR Part 50, Appendix E, Section IV.F.2.g, requires that all training, including exercises, shall provide for formal critiques in order to identify weak or deficient areas that need correction. Any weaknesses or deficiencies that are identified shall be corrected.

Contrary to the above, the licensees formal critique of an emergency preparedness exercise conducted on February 7, 2006, failed to identify weak or deficient areas. Specifically, the exercise critique failed to identify that the Station Emergency Managers Site Area Emergency event classification was an inaccurate classification.

Inspection Report# : 2006008 (pdf)

Inspection Report# : 2006010 (pdf)

Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

4Q/2006 Inspection Findings - Surry 1 Page 4 of 4 Miscellaneous Last modified : March 01, 2007

Surry 1 1Q/2007 Plant Inspection Findings Initiating Events Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform an Adequate Extent of Condition Review for Unit 1 June 29, 2006, Turbine Building Damage Event A Green self-revealing finding was identified for not performing an adequate extent of condition review in accordance with the licensees established procedures. The potential consequences of having siding torn from the Turbine Building if Unit 2 experienced steam relief valve actuations, as had occurred on Unit 1 on June 29, 2006, was not recognized. Consequently on October 7, 2006, the actuation of Unit 2 steam relief valves resulted in the loss of normal power to two emergency buses on Unit 1 and one emergency bus on Unit 2 due to flying debris impacting electrical conductors. The finding was entered into the licensees corrective action program as Condition Report (CR) 003598. Sections of the Turbine Building near the discharge of the steam relief valves were temporarily strengthened while additional long term corrective actions were being evaluated.

The finding is more than minor due to its impact on the Initiating Events objective to limit the likelihood of those events that upset plant stability and the related attribute of human performance. This finding was of very low safety significance (Green) because the increase in risk was limited by the duration of the condition. The cause of the finding was directly related to the appropriate and timely corrective actions aspect of the problem identification and resolution cross-cutting area because sufficient information was available for the licensee to have identified potential damage to plant equipment and taken actions to limit it when the Unit 2 steam relief valves actuated.

Inspection Report# : 2007002 (pdf)

Mitigating Systems Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures During Maintenance Resulting in a Stuck Control Rod A self-revealing, non-cited violation of Technical Specification 6.4.D, Unit Operating Procedures and Programs, was identified for failure to follow procedure. Specifically, foreign material was left inside a Unit 1 control rod guide tube which prevented the full insertion of control rod K-14 during a manual reactor trip. The procedure in use during maintenance specifically required an inspection for and removal of all foreign material from the control rod guide tube. The licensee entered this violation in their corrective action program as CR002285 for resolution, performed a root cause analysis, and determined corrective actions.

The finding is more than minor because it affects the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences to the reactor core and resulted in a degraded rod control system. The inspectors determined that the finding is of very low safety significance (Green) since only one control rod was affected. The cause of the finding was directly related to the procedural compliance aspect of the human performance cross-cutting area because personnel failed to adequately perform a search for foreign material as required by procedures.

Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007

Identified By: Self-Revealing Item Type: NCV NonCited Violation Breaker Control Circuit Design Deficiency Results in Failure to Supply Emergency Bus 1J A self-revealing, non-cited violation of 10 CFR 50.63 was identified regarding a breaker control circuit design deficiency which prevented the licensee from supplying the 1J emergency bus on Unit 1 from the alternate AC diesel generator. The problem occurred during a transient on October 7, 2006, in which the actuation of Unit 2 steam relief valves resulted in the loss of normal power to two emergency buses on Unit 1 and one emergency bus on Unit 2 due to flying debris impacting electrical conductors. The licensee installed a modification to correct the breaker circuit design deficiency.

The finding is more than minor because it impacted the Mitigating Systems objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) and the related attribute of design control. This finding was of very low safety significance (Green) because the recovery of the alternate AC diesel generators ability to energize a safety bus after four hours was credible.

Inspection Report# : 2007002 (pdf)

Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Proceduralized Departures from TS The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments.

Specifically, the licensee implemented proceduralized departures from the approved station technical specifications (TS) without the required NRC approval in procedures AP-13.0, Turbine Building Flooding, revision 13, and FCA 6.01, Uncontrollable Turbine Building Flooding, revision 2.

This finding was evaluated using traditional enforcement since it impacted or impeded the regulatory process in that the licensee improperly used the 10 CFR 50.59, Changes, Tests, and Experiments, process to incorporate operator actions inconsistent with the TS. This finding was of more than minor safety significance because the procedure changes improperly bypassed the required NRC review and approval prior to implementation. The unapproved procedural actions would only be involved at the end of a very rare accident sequence. Given the time during the accident sequence in which these actions were to be accomplished, the actions were not a determent to core damage. Therefore, the violation was of very low safety significance. The finding is identified as Severity Level IV because the noncompliance is not considered to be of more than very low significance based on risk.

Inspection Report# : 2006005 (pdf)

Significance: Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Removal of Damper Motor Operators From CO2 System in Normal Switchgear Rooms The team identified a non-cited violation of Operating License Condition 3.I for removing the automatic feature of ventilation dampers which degraded the fixed gaseous suppression system in the normal switchgear room at both units by allowing carbon dioxide to flow out should the manual operated dampers be in the open position.

The finding is more than minor because it was associated with the reactor safety, mitigating systems cornerstone attribute of protection against external factors, i.e. fire, and it affected the objective of ensuring reliability and capability of systems that respond to initiating events. The finding is of very low safety significance because the frequency of fires potentially challenging mitigating systems was relatively low and multiple trains of shutdown equipment would be available.

Inspection Report# : 2006009 (pdf)

Significance: Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Post-Fire Safe Shutdown During a Fire in Mechanical Equipment Room 3 The team identified a non-cited violation of Technical Specification 6.4.E for failure to provide an adequate post-fire safe shutdown procedure. Procedure 0-FCA-7.00, Rev. 10, failed to ensure that a source of water would be aligned to the suction of the charging pump service water pumps during a severe fire in Mechanical Equipment Room 3. Consequently,

all charging pumps of both units could have no service water cooling resulting in pump overheating and failure.

The finding is greater than minor because it affected the objective of the mitigating system cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events. Since the procedure had been in place for less than one month and during that time a source of water could have been aligned, this finding is of very low safety significance.

Inspection Report# : 2006009 (pdf)

Barrier Integrity Emergency Preparedness Significance: Oct 27, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Activate ERDS Within One Hour of an Alert Declaration A green self-revealing non-cited violation of 10 CFR 50.72(a)(4) was identified. During the October 7, 2006, partial loss of offsite power event, the licensee failed to activate the Emergency Response Data System (ERDS) within one hour of an Alert declaration. The ERDS was not made operable until approximately five and one-half hours after the Alert declaration due to an upgrade to the telephone exchange that had been done seven days prior to the event.

The finding is more than minor due to its impact on the Emergency Preparedness cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency and the related attribute of Emergency Response Organization response. The finding is of very low safety significance (Green) because it involved a failure to implement (in distinction to a failure to meet) an NRC emergency planning standard. The cause of the finding is related to the cross-cutting area of human performance, in that, the licensee failed to reprogram the telephone exchange following a telephone system change which occurred prior to the event. Upon discovery, the licensee immediately reprogrammed the telephone exchange and entered the problem into their corrective action program as condition report CR 002183.

Inspection Report# : 2006011 (pdf)

Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.

Miscellaneous

Last modified : June 01, 2007 Surry 1 2Q/2007 Plant Inspection Findings Initiating Events Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an adequte Risk Assessment for Unit 2 Cross-Under Relief Valve Event The inspectors identified a non-cited violation of 10 CFR 50.65 (a)(4), which requires that the licensee assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, in assessing the increase in risk of planned maintenance activities, the licensee failed to adequately assess planned risk. The licensee entered this issue in their corrective action program as CR-003611 for resolution.

The finding was considered to be more than minor because the licensees risk assessment had known errors or incorrect assumptions that had the potential to change the outcome of the assessment. The inspectors determined that the finding is of very low safety significance (Green) since the incremental core damage probability deficit was less than 1E-6. The inspectors determined that the cause of the finding was related to the proper work planning aspect of the human performance cross-cutting area.

Inspection Report# : 2007003 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform an Adequate Extent of Condition Review for Unit 1 June 29, 2006, Turbine Building Damage Event A Green self-revealing finding was identified for not performing an adequate extent of condition review in accordance with the licensees established procedures. The potential consequences of having siding torn from the Turbine Building if Unit 2 experienced steam relief valve actuations, as had occurred on Unit 1 on June 29, 2006, was not recognized. Consequently on October 7, 2006, the actuation of Unit 2 steam relief valves resulted in the loss of normal power to two emergency buses on Unit 1 and one emergency bus on Unit 2 due to flying debris impacting electrical conductors. The finding was entered into the licensees corrective action program as Condition Report (CR) 003598.

Sections of the Turbine Building near the discharge of the steam relief valves were temporarily strengthened while additional long term corrective actions were being evaluated.

The finding is more than minor due to its impact on the Initiating Events objective to limit the likelihood of those events that upset plant stability and the related attribute of human performance. This finding was of very low safety significance (Green) because the increase in risk was limited by the duration of the condition. The cause of the finding was directly related to the appropriate and timely corrective actions aspect of the problem identification and resolution cross-cutting area because sufficient information was available for the licensee to have identified potential damage to plant equipment and taken actions to limit it when the Unit 2 steam relief valves actuated.

Inspection Report# : 2007002 (pdf)

Mitigating Systems Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure the Suitability of Application of Equipment Essential to Safety-Related Functions

The NRC identified a non-cited violation (NCV) for the failure to ensure the suitability of application of equipment essential to the safety-related functions of structures, systems, and components (SSCs) through their commercial dedication process as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control. The licensee entered each of the two examples identified by the team into their corrective actions program as CR-013984, including an action to review their overall commercial dedication program.

The examples involve Agastat 7000 relays used in supporting the emergency diesel generator (EDG) start sequence and pressure control valves (PCVs) for use in the safety-related air supply supporting design operation of the power-operated relief valves (PORVs). In the first example, the licensees commercial grade dedication did not verify the adequacy of seismic qualification. In the second, the licensee utilized a non-conservative test pressure as part of their dedication to critical characteristics. Both examples of the finding are more than minor because they are associated with the Design Control attribute affecting the Reactor Safety Mitigating Systems Cornerstone objective. The examples to the finding were evaluated using the SDP for Reactor Inspection Findings for At-Power Situations. The SDP Phase 1 analysis demonstrates the finding to be of very low safety significance (Green) as the licensee confirmed operability in accordance with plant procedures for both examples. The cause of the first example is related to the cross cutting aspect of human performance.

Inspection Report# : 2007003 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures During Maintenance Resulting in a Stuck Control Rod A self-revealing, non-cited violation of Technical Specification 6.4.D, Unit Operating Procedures and Programs, was identified for failure to follow procedure. Specifically, foreign material was left inside a Unit 1 control rod guide tube which prevented the full insertion of control rod K-14 during a manual reactor trip. The procedure in use during maintenance specifically required an inspection for and removal of all foreign material from the control rod guide tube. The licensee entered this violation in their corrective action program as CR002285 for resolution, performed a root cause analysis, and determined corrective actions.

The finding is more than minor because it affects the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences to the reactor core and resulted in a degraded rod control system. The inspectors determined that the finding is of very low safety significance (Green) since only one control rod was affected. The cause of the finding was directly related to the procedural compliance aspect of the human performance cross-cutting area because personnel failed to adequately perform a search for foreign material as required by procedures.

Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Breaker Control Circuit Design Deficiency Results in Failure to Supply Emergency Bus 1J A self-revealing, non-cited violation of 10 CFR 50.63 was identified regarding a breaker control circuit design deficiency which prevented the licensee from supplying the 1J emergency bus on Unit 1 from the alternate AC diesel generator. The problem occurred during a transient on October 7, 2006, in which the actuation of Unit 2 steam relief valves resulted in the loss of normal power to two emergency buses on Unit 1 and one emergency bus on Unit 2 due to flying debris impacting electrical conductors. The licensee installed a modification to correct the breaker circuit design deficiency.

The finding is more than minor because it impacted the Mitigating Systems objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) and the related attribute of design control. This finding was of very low safety significance (Green) because the recovery of the alternate AC diesel generators ability to energize a safety bus after four hours was credible.

Inspection Report# : 2007002 (pdf)

Significance: SL-IV Dec 31, 2006

Identified By: NRC Item Type: NCV NonCited Violation Proceduralized Departures from TS The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments. Specifically, the licensee implemented proceduralized departures from the approved station technical specifications (TS) without the required NRC approval in procedures AP-13.0, Turbine Building Flooding, revision 13, and FCA 6.01, Uncontrollable Turbine Building Flooding, revision 2.

This finding was evaluated using traditional enforcement since it impacted or impeded the regulatory process in that the licensee improperly used the 10 CFR 50.59, Changes, Tests, and Experiments, process to incorporate operator actions inconsistent with the TS. This finding was of more than minor safety significance because the procedure changes improperly bypassed the required NRC review and approval prior to implementation. The unapproved procedural actions would only be involved at the end of a very rare accident sequence. Given the time during the accident sequence in which these actions were to be accomplished, the actions were not a determent to core damage.

Therefore, the violation was of very low safety significance. The finding is identified as Severity Level IV because the noncompliance is not considered to be of more than very low significance based on risk.

Inspection Report# : 2006005 (pdf)

Barrier Integrity Emergency Preparedness Significance: Oct 27, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Activate ERDS Within One Hour of an Alert Declaration A green self-revealing non-cited violation of 10 CFR 50.72(a)(4) was identified. During the October 7, 2006, partial loss of offsite power event, the licensee failed to activate the Emergency Response Data System (ERDS) within one hour of an Alert declaration. The ERDS was not made operable until approximately five and one-half hours after the Alert declaration due to an upgrade to the telephone exchange that had been done seven days prior to the event.

The finding is more than minor due to its impact on the Emergency Preparedness cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency and the related attribute of Emergency Response Organization response. The finding is of very low safety significance (Green) because it involved a failure to implement (in distinction to a failure to meet) an NRC emergency planning standard. The cause of the finding is related to the cross-cutting area of human performance, in that, the licensee failed to reprogram the telephone exchange following a telephone system change which occurred prior to the event. Upon discovery, the licensee immediately reprogrammed the telephone exchange and entered the problem into their corrective action program as condition report CR 002183.

Inspection Report# : 2006011 (pdf)

Occupational Radiation Safety Public Radiation Safety

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : August 24, 2007

Surry 1 3Q/2007 Plant Inspection Findings Initiating Events Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an adequte Risk Assessment for Unit 2 Cross-Under Relief Valve Event The inspectors identified a non-cited violation of 10 CFR 50.65 (a)(4), which requires that the licensee assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, in assessing the increase in risk of planned maintenance activities, the licensee failed to adequately assess planned risk. The licensee entered this issue in their corrective action program as CR-003611 for resolution.

The finding was considered to be more than minor because the licensees risk assessment had known errors or incorrect assumptions that had the potential to change the outcome of the assessment. The inspectors determined that the finding is of very low safety significance (Green) since the incremental core damage probability deficit was less than 1E-6. The inspectors determined that the cause of the finding was related to the proper work planning aspect of the human performance cross-cutting area.

Inspection Report# : 2007003 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform an Adequate Extent of Condition Review for Unit 1 June 29, 2006, Turbine Building Damage Event A Green self-revealing finding was identified for not performing an adequate extent of condition review in accordance with the licensees established procedures. The potential consequences of having siding torn from the Turbine Building if Unit 2 experienced steam relief valve actuations, as had occurred on Unit 1 on June 29, 2006, was not recognized. Consequently on October 7, 2006, the actuation of Unit 2 steam relief valves resulted in the loss of normal power to two emergency buses on Unit 1 and one emergency bus on Unit 2 due to flying debris impacting electrical conductors. The finding was entered into the licensees corrective action program as Condition Report (CR) 003598.

Sections of the Turbine Building near the discharge of the steam relief valves were temporarily strengthened while additional long term corrective actions were being evaluated.

The finding is more than minor due to its impact on the Initiating Events objective to limit the likelihood of those events that upset plant stability and the related attribute of human performance. This finding was of very low safety significance (Green) because the increase in risk was limited by the duration of the condition. The cause of the finding was directly related to the appropriate and timely corrective actions aspect of the problem identification and resolution cross-cutting area because sufficient information was available for the licensee to have identified potential damage to plant equipment and taken actions to limit it when the Unit 2 steam relief valves actuated.

Inspection Report# : 2007002 (pdf)

Mitigating Systems Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure the Suitability of Application of Equipment Essential to Safety-Related Functions

The NRC identified a non-cited violation (NCV) for the failure to ensure the suitability of application of equipment essential to the safety-related functions of structures, systems, and components (SSCs) through their commercial dedication process as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control. The licensee entered each of the two examples identified by the team into their corrective actions program as CR-013984, including an action to review their overall commercial dedication program.

The examples involve Agastat 7000 relays used in supporting the emergency diesel generator (EDG) start sequence and pressure control valves (PCVs) for use in the safety-related air supply supporting design operation of the power-operated relief valves (PORVs). In the first example, the licensees commercial grade dedication did not verify the adequacy of seismic qualification. In the second, the licensee utilized a non-conservative test pressure as part of their dedication to critical characteristics. Both examples of the finding are more than minor because they are associated with the Design Control attribute affecting the Reactor Safety Mitigating Systems Cornerstone objective. The examples to the finding were evaluated using the SDP for Reactor Inspection Findings for At-Power Situations. The SDP Phase 1 analysis demonstrates the finding to be of very low safety significance (Green) as the licensee confirmed operability in accordance with plant procedures for both examples. The cause of the first example is related to the cross cutting aspect of human performance.

Inspection Report# : 2007003 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures During Maintenance Resulting in a Stuck Control Rod A self-revealing, non-cited violation of Technical Specification 6.4.D, Unit Operating Procedures and Programs, was identified for failure to follow procedure. Specifically, foreign material was left inside a Unit 1 control rod guide tube which prevented the full insertion of control rod K-14 during a manual reactor trip. The procedure in use during maintenance specifically required an inspection for and removal of all foreign material from the control rod guide tube. The licensee entered this violation in their corrective action program as CR002285 for resolution, performed a root cause analysis, and determined corrective actions.

The finding is more than minor because it affects the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences to the reactor core and resulted in a degraded rod control system. The inspectors determined that the finding is of very low safety significance (Green) since only one control rod was affected. The cause of the finding was directly related to the procedural compliance aspect of the human performance cross-cutting area because personnel failed to adequately perform a search for foreign material as required by procedures.

Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Breaker Control Circuit Design Deficiency Results in Failure to Supply Emergency Bus 1J A self-revealing, non-cited violation of 10 CFR 50.63 was identified regarding a breaker control circuit design deficiency which prevented the licensee from supplying the 1J emergency bus on Unit 1 from the alternate AC diesel generator. The problem occurred during a transient on October 7, 2006, in which the actuation of Unit 2 steam relief valves resulted in the loss of normal power to two emergency buses on Unit 1 and one emergency bus on Unit 2 due to flying debris impacting electrical conductors. The licensee installed a modification to correct the breaker circuit design deficiency.

The finding is more than minor because it impacted the Mitigating Systems objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) and the related attribute of design control. This finding was of very low safety significance (Green) because the recovery of the alternate AC diesel generators ability to energize a safety bus after four hours was credible.

Inspection Report# : 2007002 (pdf)

Significance: SL-IV Dec 31, 2006

Identified By: NRC Item Type: NCV NonCited Violation Proceduralized Departures from TS The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments. Specifically, the licensee implemented proceduralized departures from the approved station technical specifications (TS) without the required NRC approval in procedures AP-13.0, Turbine Building Flooding, revision 13, and FCA 6.01, Uncontrollable Turbine Building Flooding, revision 2.

This finding was evaluated using traditional enforcement since it impacted or impeded the regulatory process in that the licensee improperly used the 10 CFR 50.59, Changes, Tests, and Experiments, process to incorporate operator actions inconsistent with the TS. This finding was of more than minor safety significance because the procedure changes improperly bypassed the required NRC review and approval prior to implementation. The unapproved procedural actions would only be involved at the end of a very rare accident sequence. Given the time during the accident sequence in which these actions were to be accomplished, the actions were not a determent to core damage.

Therefore, the violation was of very low safety significance. The finding is identified as Severity Level IV because the noncompliance is not considered to be of more than very low significance based on risk.

Inspection Report# : 2006005 (pdf)

Barrier Integrity Emergency Preparedness Significance: Oct 27, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Activate ERDS Within One Hour of an Alert Declaration A green self-revealing non-cited violation of 10 CFR 50.72(a)(4) was identified. During the October 7, 2006, partial loss of offsite power event, the licensee failed to activate the Emergency Response Data System (ERDS) within one hour of an Alert declaration. The ERDS was not made operable until approximately five and one-half hours after the Alert declaration due to an upgrade to the telephone exchange that had been done seven days prior to the event.

The finding is more than minor due to its impact on the Emergency Preparedness cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency and the related attribute of Emergency Response Organization response. The finding is of very low safety significance (Green) because it involved a failure to implement (in distinction to a failure to meet) an NRC emergency planning standard. The cause of the finding is related to the cross-cutting area of human performance, in that, the licensee failed to reprogram the telephone exchange following a telephone system change which occurred prior to the event. Upon discovery, the licensee immediately reprogrammed the telephone exchange and entered the problem into their corrective action program as condition report CR 002183.

Inspection Report# : 2006011 (pdf)

Occupational Radiation Safety Public Radiation Safety

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : December 07, 2007

Surry 1 4Q/2007 Plant Inspection Findings Initiating Events Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an adequte Risk Assessment for Unit 2 Cross-Under Relief Valve Event The inspectors identified a non-cited violation of 10 CFR 50.65 (a)(4), which requires that the licensee assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, in assessing the increase in risk of planned maintenance activities, the licensee failed to adequately assess planned risk. The licensee entered this issue in their corrective action program as CR-003611 for resolution.

The finding was considered to be more than minor because the licensees risk assessment had known errors or incorrect assumptions that had the potential to change the outcome of the assessment. The inspectors determined that the finding is of very low safety significance (Green) since the incremental core damage probability deficit was less than 1E-6. The inspectors determined that the cause of the finding was related to the proper work planning aspect of the human performance cross-cutting area.

Inspection Report# : 2007003 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform an Adequate Extent of Condition Review for Unit 1 June 29, 2006, Turbine Building Damage Event A Green self-revealing finding was identified for not performing an adequate extent of condition review in accordance with the licensees established procedures. The potential consequences of having siding torn from the Turbine Building if Unit 2 experienced steam relief valve actuations, as had occurred on Unit 1 on June 29, 2006, was not recognized. Consequently on October 7, 2006, the actuation of Unit 2 steam relief valves resulted in the loss of normal power to two emergency buses on Unit 1 and one emergency bus on Unit 2 due to flying debris impacting electrical conductors. The finding was entered into the licensees corrective action program as Condition Report (CR) 003598.

Sections of the Turbine Building near the discharge of the steam relief valves were temporarily strengthened while additional long term corrective actions were being evaluated.

The finding is more than minor due to its impact on the Initiating Events objective to limit the likelihood of those events that upset plant stability and the related attribute of human performance. This finding was of very low safety significance (Green) because the increase in risk was limited by the duration of the condition. The cause of the finding was directly related to the appropriate and timely corrective actions aspect of the problem identification and resolution cross-cutting area because sufficient information was available for the licensee to have identified potential damage to plant equipment and taken actions to limit it when the Unit 2 steam relief valves actuated.

Inspection Report# : 2007002 (pdf)

Mitigating Systems Significance: Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Emergency Service Water Pump 1-SW-P-1B Inoperable Due to an Inadequate Maintenance Instruction for

Reassembly of the ESW strainer A Green self-revealing non-cited violation of Technical Specification 6.4, Unit Operating Procedures and Programs, was identified for failure to have an adequate maintenance procedure for the emergency service water (ESW) pump strainer. This resulted in the emergency service water pump 1-SW-P-1B being declared inoperable. This procedure failed to provide adequate instructions for the reassembly of ESW strainer 1-SW-STR-4B. The finding was documented in the licensees corrective action program as condition report CR023818. Corrective action was taken to restore pump operability and to correct the procedure and post-maintenance test error.

The finding is more than minor, because it is associated with the operability, availability, reliability, or function of a system or train in a mitigating system. This finding was evaluated using the Significance Determination Process and was determined to be of very low safety significance because it did not result in a loss of safety function or the loss of a single train of ESW for more than the allowed Technical Specification outage time. This finding has a cross-cutting aspect in the area of human performance work practices (H.4.a), because personnel proceeded in the face of uncertainty when they continued to re-assemble the strainer operating mechanism without the requisite work instructions.

Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Temporary Fire Suppression Capacity Not Equivalent to Unit 1 Containment Fire Hose Stations A Green NRC-identified non-cited violation of paragraph (a)(1) of 10 CFR 50.48, "Fire Protection," was identified for failure to maintain the fire suppression capability for the Unit 1 containment building as specified by the approved fire protection plan. On October 27, 2007, the licensee failed to provide equivalent fire suppression capacity when the Unit 1 containment fire hose stations were removed from service for repair. This finding was entered into the licensees corrective action program as condition report CR025073. Planned corrective actions included developing equivalent fire suppression capacity determinations for other hose stations.

This finding is more than minor because it was associated with a degradation of a fire protection feature. The finding is of very low safety significance because it involved low degradation of a fixed fire protection system. A significant cause of this finding involved the Decision-Making component of the cross-cutting area of Human Performance and the aspect of making safety-significant or risk-significant decisions using a systematic process, in that, a formal evaluation was not used to determine equivalent capacity (H.1.a).

Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Carbon Dioxide Suppression System Degraded in Two Fire Areas in Unit 1 and Three Fire Areas in Unit 2 The NRC identified a Green NCV of Unit 1 and Unit 2 Operating License Condition 3.I because the installed carbon dioxide (CO2) fire suppression systems could not be shown to deliver the design basis gas concentration. This finding applied to the Unit 1 and Unit 2 normal switchgear rooms, the Unit 2 cable tunnel, and the Unit 1 and Unit 2 cable vaults. The licensee had implemented or initiated system modifications to address this violation.

The finding is more than minor because it affects the Mitigating Systems cornerstone objective of ensuring reliability and capability of systems that respond to initiating events and the cornerstone attribute of protection against external factors, i.e. fire. The finding was determined to be of very low safety significance in a Significance Determination Process Phase 3 analysis. For the cable vault areas, the analysis showed that fires could initiate scenarios which could challenge the mitigating systems. However, the risk of these scenarios was calculated to be in the very low significance band. Analysis with respect to the normal switchgear rooms led to the conclusion that it was of very low safety significance primarily due to the frequency of fires potentially challenging mitigating systems being relatively low and the availability of unaffected safety-related shutdown systems. The finding for the Unit 2 cable tunnel was also of very low safety significance because it did not have any significant fixed ignition sources (cables were thermoset type) and the probability for transient combustible fires or hot work initiated fires damaging important cables was judged to be low.

Inspection Report# : 2007005 (pdf)

Significance: Oct 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Catqagorize a Maintenance Preventable Functional Failure The inspectors identified an NCV of 10 CFR 50.65 (a)(2) after Surry Power Station failed to categorize the failure of the Unit 2 Charging Pump Component Cooling System as a maintenance preventable functional failure and accordingly, failed to monitor the component as required by 10 CFR 50.65 (a)(1). Condition Report 021045.

The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of the Charging System. Example 7.b in MC 0612, App. E, states that violations of Paragraph 10 CFR 50.65 (a)(2), failure to demonstrate effective control of performance or condition and not putting the affected (SSCs) in (a)(1), are not minor because they necessarily involve degraded SSC performance or condition. The finding is of very low safety significance because the failure to place the system in (a)(1) status did not lead to any further instances of system unreliability or unavailability. The cause of this finding was directly related to the aspect of Training of personnel in the cross-cutting area of human performance (resources component) because the engineer conducting the maintenance rule evaluation, the Maintenance Rule Program Coordinator, and Engineering Supervisor reviewing the evaluation, did not fully understand when to apply the functional failure exemptions. (IMC 0305, H.2.b)

Inspection Report# : 2007008 (pdf)

Significance: Oct 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct Procedures Related to the Operation of the Auxiliary Feedwater System NRC identified non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to revise procedure ECA 0.0, Loss of all AC Power, as corrective action for a condition identified by the licensee that could cause a loss of Net Positive Suction Head (NPSH) to the Turbine Driven Auxiliary Feedwater Pump (TDAFW), and potential damage to the only available feedwater pump during a loss of all AC power event.

Other procedures where single AFW pump operation could cause inadequate NPSH had been revised.

The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because no actual safety function was lost. The cause of the finding is related to the cross-cutting area of problem identification and resolution and the aspect of lack of thoroughness of evaluation such that the resolution addresses the causes and extent of conditions. (IMC 0305, P.1.c) [Section 4OA2.a (3)(ii)]

Inspection Report# : 2007008 (pdf)

Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure the Suitability of Application of Equipment Essential to Safety-Related Functions The NRC identified a non-cited violation (NCV) for the failure to ensure the suitability of application of equipment essential to the safety-related functions of structures, systems, and components (SSCs) through their commercial dedication process as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control. The licensee entered each of the two examples identified by the team into their corrective actions program as CR-013984, including an action to review their overall commercial dedication program.

The examples involve Agastat 7000 relays used in supporting the emergency diesel generator (EDG) start sequence and pressure control valves (PCVs) for use in the safety-related air supply supporting design operation of the power-operated relief valves (PORVs). In the first example, the licensees commercial grade dedication did not verify the

adequacy of seismic qualification. In the second, the licensee utilized a non-conservative test pressure as part of their dedication to critical characteristics. Both examples of the finding are more than minor because they are associated with the Design Control attribute affecting the Reactor Safety Mitigating Systems Cornerstone objective. The examples to the finding were evaluated using the SDP for Reactor Inspection Findings for At-Power Situations. The SDP Phase 1 analysis demonstrates the finding to be of very low safety significance (Green) as the licensee confirmed operability in accordance with plant procedures for both examples. The cause of the first example is related to the cross cutting aspect of human performance.

Inspection Report# : 2007003 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures During Maintenance Resulting in a Stuck Control Rod A self-revealing, non-cited violation of Technical Specification 6.4.D, Unit Operating Procedures and Programs, was identified for failure to follow procedure. Specifically, foreign material was left inside a Unit 1 control rod guide tube which prevented the full insertion of control rod K-14 during a manual reactor trip. The procedure in use during maintenance specifically required an inspection for and removal of all foreign material from the control rod guide tube. The licensee entered this violation in their corrective action program as CR002285 for resolution, performed a root cause analysis, and determined corrective actions.

The finding is more than minor because it affects the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences to the reactor core and resulted in a degraded rod control system. The inspectors determined that the finding is of very low safety significance (Green) since only one control rod was affected. The cause of the finding was directly related to the procedural compliance aspect of the human performance cross-cutting area because personnel failed to adequately perform a search for foreign material as required by procedures.

Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Breaker Control Circuit Design Deficiency Results in Failure to Supply Emergency Bus 1J A self-revealing, non-cited violation of 10 CFR 50.63 was identified regarding a breaker control circuit design deficiency which prevented the licensee from supplying the 1J emergency bus on Unit 1 from the alternate AC diesel generator. The problem occurred during a transient on October 7, 2006, in which the actuation of Unit 2 steam relief valves resulted in the loss of normal power to two emergency buses on Unit 1 and one emergency bus on Unit 2 due to flying debris impacting electrical conductors. The licensee installed a modification to correct the breaker circuit design deficiency.

The finding is more than minor because it impacted the Mitigating Systems objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) and the related attribute of design control. This finding was of very low safety significance (Green) because the recovery of the alternate AC diesel generators ability to energize a safety bus after four hours was credible.

Inspection Report# : 2007002 (pdf)

Barrier Integrity Emergency Preparedness

Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : February 04, 2008

Surry 1 1Q/2008 Plant Inspection Findings Initiating Events Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an adequte Risk Assessment for Unit 2 Cross-Under Relief Valve Event The inspectors identified a non-cited violation of 10 CFR 50.65 (a)(4), which requires that the licensee assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, in assessing the increase in risk of planned maintenance activities, the licensee failed to adequately assess planned risk. The licensee entered this issue in their corrective action program as CR-003611 for resolution.

The finding was considered to be more than minor because the licensees risk assessment had known errors or incorrect assumptions that had the potential to change the outcome of the assessment. The inspectors determined that the finding is of very low safety significance (Green) since the incremental core damage probability deficit was less than 1E-6. The inspectors determined that the cause of the finding was related to the proper work planning aspect of the human performance cross-cutting area.

Inspection Report# : 2007003 (pdf)

Mitigating Systems Significance: Mar 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Start-up Procedure which Resulted in Leaving Loose Fibrous Insulation in Containment An NRC-identified, non-cited violation (NCV) of very low safety significance was identified for the failure to follow start-up procedure 1-GOP-1.7, revision 2, Unit Startup, RCS Heat Up from Ambient to HSD, which resulted in leaving loose fibrous insulation in containment (Unit 1).

This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using the IMC 0609, "Significance Determination Process," Phase 1 Worksheets, the finding is determined to have very low safety significance (Green) since it only affected the mitigating systems cornerstone and did not represent a loss of system safety function. The cause of this finding had cross-cutting aspects associated with work practices of the Human Performance area in that the licensee did not provide the appropriate oversight of contractors conducting the containment walk downs (H.4.c).

The finding was entered into the corrective action program as Condition Report 02564. Corrective actions to remove the fibrous material from containment prior to startup and to establish the extent of condition and potential impact on Unit-2 were adequate.

Inspection Report# : 2008002 (pdf)

Significance: Feb 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Use Limiting Case 4160 VAC Bus Frequency and Voltage in Design Calculations The inspectors identified two examples of a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to evaluate variations of emergency diesel generator (EDG) output frequency in electrical

design loading calculations, and failure to consider worst case 4160 VAC bus voltage in safety related motor starting calculations. This finding was entered into the licensees corrective action program as condition reports (CR) 091493 and 091494. Planned corrective actions included revision of the EDG loading calculations to incorporate the most limiting voltages and frequencies.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the EDGs to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiencies did not result in any EDG being inoperable based upon additional analysis that showed that the EDGs had sufficient margin to accommodate the increased loading due to worst case acceptably high EDG output frequency; and all safety related motor loads remained operable since they were still capable of starting with the revised worst case low voltage values.

Inspection Report# : 2008006 (pdf)

Significance: Feb 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Appropriate Acceptance Criteria for Testing Battery Voltage at the One Minute Mark The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for incorrect acceptance criteria in test procedure 1-EPT-0106-01, Main Station Battery 1A Service Test. This finding was entered into the licensees corrective action program as condition report 091906. Planned corrective actions included revision of the main station battery test procedures to include the correct voltage at the one minute mark.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the station batteries to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. maintenance and testing procedures. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in station batteries being inoperable based upon a recent review of station battery discharge test results.

The inspectors determined that the lack of a thorough evaluation of condition report 022112, which addressed deficiencies in station battery test procedures such that resolutions addressed causes, was a significant cause of this performance deficiency. Failure to thoroughly evaluate problems such that resolutions address causes is directly related to the Corrective Action Program component of the cross-cutting area of Problem Identification and Resolution and the aspect of thorough evaluation of problems (P.1(c)).

Inspection Report# : 2008006 (pdf)

Significance: Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Emergency Service Water Pump 1-SW-P-1B Inoperable Due to an Inadequate Maintenance Instruction for Reassembly of the ESW strainer A Green self-revealing non-cited violation of Technical Specification 6.4, Unit Operating Procedures and Programs, was identified for failure to have an adequate maintenance procedure for the emergency service water (ESW) pump strainer. This resulted in the emergency service water pump 1-SW-P-1B being declared inoperable. This procedure failed to provide adequate instructions for the reassembly of ESW strainer 1-SW-STR-4B. The finding was documented in the licensees corrective action program as condition report CR023818. Corrective action was taken to restore pump operability and to correct the procedure and post-maintenance test error.

The finding is more than minor, because it is associated with the operability, availability, reliability, or function of a system or train in a mitigating system. This finding was evaluated using the Significance Determination Process and was determined to be of very low safety significance because it did not result in a loss of safety function or the loss of a single train of ESW for more than the allowed Technical Specification outage time. This finding has a cross-cutting aspect in the area of human performance work practices (H.4.a), because personnel proceeded in the face of uncertainty when they continued to re-assemble the strainer operating mechanism without the requisite work

instructions.

Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Temporary Fire Suppression Capacity Not Equivalent to Unit 1 Containment Fire Hose Stations A Green NRC-identified non-cited violation of paragraph (a)(1) of 10 CFR 50.48, "Fire Protection," was identified for failure to maintain the fire suppression capability for the Unit 1 containment building as specified by the approved fire protection plan. On October 27, 2007, the licensee failed to provide equivalent fire suppression capacity when the Unit 1 containment fire hose stations were removed from service for repair. This finding was entered into the licensees corrective action program as condition report CR025073. Planned corrective actions included developing equivalent fire suppression capacity determinations for other hose stations.

This finding is more than minor because it was associated with a degradation of a fire protection feature. The finding is of very low safety significance because it involved low degradation of a fixed fire protection system. A significant cause of this finding involved the Decision-Making component of the cross-cutting area of Human Performance and the aspect of making safety-significant or risk-significant decisions using a systematic process, in that, a formal evaluation was not used to determine equivalent capacity (H.1.a).

Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Carbon Dioxide Suppression System Degraded in Two Fire Areas in Unit 1 and Three Fire Areas in Unit 2 The NRC identified a Green NCV of Unit 1 and Unit 2 Operating License Condition 3.I because the installed carbon dioxide (CO2) fire suppression systems could not be shown to deliver the design basis gas concentration. This finding applied to the Unit 1 and Unit 2 normal switchgear rooms, the Unit 2 cable tunnel, and the Unit 1 and Unit 2 cable vaults. The licensee had implemented or initiated system modifications to address this violation.

The finding is more than minor because it affects the Mitigating Systems cornerstone objective of ensuring reliability and capability of systems that respond to initiating events and the cornerstone attribute of protection against external factors, i.e. fire. The finding was determined to be of very low safety significance in a Significance Determination Process Phase 3 analysis. For the cable vault areas, the analysis showed that fires could initiate scenarios which could challenge the mitigating systems. However, the risk of these scenarios was calculated to be in the very low significance band. Analysis with respect to the normal switchgear rooms led to the conclusion that it was of very low safety significance primarily due to the frequency of fires potentially challenging mitigating systems being relatively low and the availability of unaffected safety-related shutdown systems. The finding for the Unit 2 cable tunnel was also of very low safety significance because it did not have any significant fixed ignition sources (cables were thermoset type) and the probability for transient combustible fires or hot work initiated fires damaging important cables was judged to be low.

Inspection Report# : 2007005 (pdf)

Significance: Oct 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Catqagorize a Maintenance Preventable Functional Failure The inspectors identified an NCV of 10 CFR 50.65 (a)(2) after Surry Power Station failed to categorize the failure of the Unit 2 Charging Pump Component Cooling System as a maintenance preventable functional failure and accordingly, failed to monitor the component as required by 10 CFR 50.65 (a)(1). Condition Report 021045.

The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of the Charging System. Example 7.b in MC 0612, App. E, states that violations of Paragraph 10 CFR 50.65 (a)(2), failure to demonstrate effective control of performance or condition and not putting the affected (SSCs) in (a)(1), are not minor because they necessarily involve degraded SSC performance or condition. The finding is of very low safety

significance because the failure to place the system in (a)(1) status did not lead to any further instances of system unreliability or unavailability. The cause of this finding was directly related to the aspect of Training of personnel in the cross-cutting area of human performance (resources component) because the engineer conducting the maintenance rule evaluation, the Maintenance Rule Program Coordinator, and Engineering Supervisor reviewing the evaluation, did not fully understand when to apply the functional failure exemptions. (IMC 0305, H.2.b)

Inspection Report# : 2007008 (pdf)

Significance: Oct 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct Procedures Related to the Operation of the Auxiliary Feedwater System NRC identified non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to revise procedure ECA 0.0, Loss of all AC Power, as corrective action for a condition identified by the licensee that could cause a loss of Net Positive Suction Head (NPSH) to the Turbine Driven Auxiliary Feedwater Pump (TDAFW), and potential damage to the only available feedwater pump during a loss of all AC power event.

Other procedures where single AFW pump operation could cause inadequate NPSH had been revised.

The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because no actual safety function was lost. The cause of the finding is related to the cross-cutting area of problem identification and resolution and the aspect of lack of thoroughness of evaluation such that the resolution addresses the causes and extent of conditions. (IMC 0305, P.1.c) [Section 4OA2.a (3)(ii)]

Inspection Report# : 2007008 (pdf)

Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure the Suitability of Application of Equipment Essential to Safety-Related Functions The NRC identified a non-cited violation (NCV) for the failure to ensure the suitability of application of equipment essential to the safety-related functions of structures, systems, and components (SSCs) through their commercial dedication process as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control. The licensee entered each of the two examples identified by the team into their corrective actions program as CR-013984, including an action to review their overall commercial dedication program.

The examples involve Agastat 7000 relays used in supporting the emergency diesel generator (EDG) start sequence and pressure control valves (PCVs) for use in the safety-related air supply supporting design operation of the power-operated relief valves (PORVs). In the first example, the licensees commercial grade dedication did not verify the adequacy of seismic qualification. In the second, the licensee utilized a non-conservative test pressure as part of their dedication to critical characteristics. Both examples of the finding are more than minor because they are associated with the Design Control attribute affecting the Reactor Safety Mitigating Systems Cornerstone objective. The examples to the finding were evaluated using the SDP for Reactor Inspection Findings for At-Power Situations. The SDP Phase 1 analysis demonstrates the finding to be of very low safety significance (Green) as the licensee confirmed operability in accordance with plant procedures for both examples. The cause of the first example is related to the cross cutting aspect of human performance.

Inspection Report# : 2007003 (pdf)

Barrier Integrity

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : June 05, 2008

Surry 1 2Q/2008 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Mar 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Start-up Procedure which Resulted in Leaving Loose Fibrous Insulation in Containment An NRC-identified, non-cited violation (NCV) of very low safety significance was identified for the failure to follow start-up procedure 1-GOP-1.7, revision 2, Unit Startup, RCS Heat Up from Ambient to HSD, which resulted in leaving loose fibrous insulation in containment (Unit 1).

This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using the IMC 0609, "Significance Determination Process," Phase 1 Worksheets, the finding is determined to have very low safety significance (Green) since it only affected the mitigating systems cornerstone and did not represent a loss of system safety function. The cause of this finding had cross-cutting aspects associated with work practices of the Human Performance area in that the licensee did not provide the appropriate oversight of contractors conducting the containment walk downs (H.4.c).

The finding was entered into the corrective action program as Condition Report 02564. Corrective actions to remove the fibrous material from containment prior to startup and to establish the extent of condition and potential impact on Unit-2 were adequate.

Inspection Report# : 2008002 (pdf)

Significance: Feb 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Use Limiting Case 4160 VAC Bus Frequency and Voltage in Design Calculations The inspectors identified two examples of a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to evaluate variations of emergency diesel generator (EDG) output frequency in electrical design loading calculations, and failure to consider worst case 4160 VAC bus voltage in safety related motor starting calculations. This finding was entered into the licensees corrective action program as condition reports (CR) 091493 and 091494. Planned corrective actions included revision of the EDG loading calculations to incorporate the most limiting voltages and frequencies.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the EDGs to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiencies did not result in any EDG being inoperable based upon additional analysis that showed that the EDGs had sufficient margin to accommodate the increased loading due to worst case acceptably high EDG output frequency; and all safety related motor loads remained operable since they were still capable of starting with the revised worst case low voltage values.

Inspection Report# : 2008006 (pdf)

Significance: Feb 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Appropriate Acceptance Criteria for Testing Battery Voltage at the One Minute Mark The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for incorrect acceptance criteria in test procedure 1-EPT-0106-01, Main Station Battery 1A Service Test. This finding was entered into the licensees corrective action program as condition report 091906. Planned corrective actions included revision of the main station battery test procedures to include the correct voltage at the one minute mark.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the station batteries to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. maintenance and testing procedures. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in station batteries being inoperable based upon a recent review of station battery discharge test results.

The inspectors determined that the lack of a thorough evaluation of condition report 022112, which addressed deficiencies in station battery test procedures such that resolutions addressed causes, was a significant cause of this performance deficiency. Failure to thoroughly evaluate problems such that resolutions address causes is directly related to the Corrective Action Program component of the cross-cutting area of Problem Identification and Resolution and the aspect of thorough evaluation of problems (P.1(c)).

Inspection Report# : 2008006 (pdf)

Significance: Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Emergency Service Water Pump 1-SW-P-1B Inoperable Due to an Inadequate Maintenance Instruction for Reassembly of the ESW strainer A Green self-revealing non-cited violation of Technical Specification 6.4, Unit Operating Procedures and Programs, was identified for failure to have an adequate maintenance procedure for the emergency service water (ESW) pump strainer. This resulted in the emergency service water pump 1-SW-P-1B being declared inoperable. This procedure failed to provide adequate instructions for the reassembly of ESW strainer 1-SW-STR-4B. The finding was documented in the licensees corrective action program as condition report CR023818. Corrective action was taken to restore pump operability and to correct the procedure and post-maintenance test error.

The finding is more than minor, because it is associated with the operability, availability, reliability, or function of a system or train in a mitigating system. This finding was evaluated using the Significance Determination Process and was determined to be of very low safety significance because it did not result in a loss of safety function or the loss of a single train of ESW for more than the allowed Technical Specification outage time. This finding has a cross-cutting aspect in the area of human performance work practices (H.4.a), because personnel proceeded in the face of uncertainty when they continued to re-assemble the strainer operating mechanism without the requisite work instructions.

Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Temporary Fire Suppression Capacity Not Equivalent to Unit 1 Containment Fire Hose Stations A Green NRC-identified non-cited violation of paragraph (a)(1) of 10 CFR 50.48, "Fire Protection," was identified for failure to maintain the fire suppression capability for the Unit 1 containment building as specified by the approved fire protection plan. On October 27, 2007, the licensee failed to provide equivalent fire suppression capacity when the Unit 1 containment fire hose stations were removed from service for repair. This finding was entered into the licensees corrective action program as condition report CR025073. Planned corrective actions included developing equivalent fire suppression capacity determinations for other hose stations.

This finding is more than minor because it was associated with a degradation of a fire protection feature. The finding

is of very low safety significance because it involved low degradation of a fixed fire protection system. A significant cause of this finding involved the Decision-Making component of the cross-cutting area of Human Performance and the aspect of making safety-significant or risk-significant decisions using a systematic process, in that, a formal evaluation was not used to determine equivalent capacity (H.1.a).

Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Carbon Dioxide Suppression System Degraded in Two Fire Areas in Unit 1 and Three Fire Areas in Unit 2 The NRC identified a Green NCV of Unit 1 and Unit 2 Operating License Condition 3.I because the installed carbon dioxide (CO2) fire suppression systems could not be shown to deliver the design basis gas concentration. This finding applied to the Unit 1 and Unit 2 normal switchgear rooms, the Unit 2 cable tunnel, and the Unit 1 and Unit 2 cable vaults. The licensee had implemented or initiated system modifications to address this violation.

The finding is more than minor because it affects the Mitigating Systems cornerstone objective of ensuring reliability and capability of systems that respond to initiating events and the cornerstone attribute of protection against external factors, i.e. fire. The finding was determined to be of very low safety significance in a Significance Determination Process Phase 3 analysis. For the cable vault areas, the analysis showed that fires could initiate scenarios which could challenge the mitigating systems. However, the risk of these scenarios was calculated to be in the very low significance band. Analysis with respect to the normal switchgear rooms led to the conclusion that it was of very low safety significance primarily due to the frequency of fires potentially challenging mitigating systems being relatively low and the availability of unaffected safety-related shutdown systems. The finding for the Unit 2 cable tunnel was also of very low safety significance because it did not have any significant fixed ignition sources (cables were thermoset type) and the probability for transient combustible fires or hot work initiated fires damaging important cables was judged to be low.

Inspection Report# : 2007005 (pdf)

Significance: Oct 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Catqagorize a Maintenance Preventable Functional Failure The inspectors identified an NCV of 10 CFR 50.65 (a)(2) after Surry Power Station failed to categorize the failure of the Unit 2 Charging Pump Component Cooling System as a maintenance preventable functional failure and accordingly, failed to monitor the component as required by 10 CFR 50.65 (a)(1). Condition Report 021045.

The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of the Charging System. Example 7.b in MC 0612, App. E, states that violations of Paragraph 10 CFR 50.65 (a)(2), failure to demonstrate effective control of performance or condition and not putting the affected (SSCs) in (a)(1), are not minor because they necessarily involve degraded SSC performance or condition. The finding is of very low safety significance because the failure to place the system in (a)(1) status did not lead to any further instances of system unreliability or unavailability. The cause of this finding was directly related to the aspect of Training of personnel in the cross-cutting area of human performance (resources component) because the engineer conducting the maintenance rule evaluation, the Maintenance Rule Program Coordinator, and Engineering Supervisor reviewing the evaluation, did not fully understand when to apply the functional failure exemptions. (IMC 0305, H.2.b)

Inspection Report# : 2007008 (pdf)

Significance: Oct 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct Procedures Related to the Operation of the Auxiliary Feedwater System NRC identified non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to revise procedure ECA 0.0, Loss of all AC Power, as corrective action for a condition identified by the

licensee that could cause a loss of Net Positive Suction Head (NPSH) to the Turbine Driven Auxiliary Feedwater Pump (TDAFW), and potential damage to the only available feedwater pump during a loss of all AC power event.

Other procedures where single AFW pump operation could cause inadequate NPSH had been revised.

The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because no actual safety function was lost. The cause of the finding is related to the cross-cutting area of problem identification and resolution and the aspect of lack of thoroughness of evaluation such that the resolution addresses the causes and extent of conditions. (IMC 0305, P.1.c) [Section 4OA2.a (3)(ii)]

Inspection Report# : 2007008 (pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : August 29, 2008

Surry 1 3Q/2008 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control for the EDG Ambient Air Temperature Limit The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion III, Design Control, for a change in the EDG ambient air temperature operating limits, from 100ºF to 120ºF, that was made without an adequate design analysis. The licensee entered the issue into their corrective action program (CAP) for resolution using condition report (CR) 102488.

The inspectors concluded that the licensees failure to perform the necessary analysis to support the increase of the EDG ambient air temperature operating limit from 100ºF to 120ºF was a performance deficiency. The finding, more than minor in accordance with MC 0612, Appendix E, examples 3j and k, is associated with the design control attribute of the Mitigating System Cornerstone. The cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences was adversely affected. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4 the inspectors concluded that the finding is of very low safety significance (Green) because the condition did not represent an actual loss of safety function due to the ambient temperature exceeding 100ºF but not exceeding 105ºF. The finding also was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. A cross-cutting aspect was not assigned to the issue because it is not indicative of recent performance.

Inspection Report# : 2008004 (pdf)

Significance: Mar 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Start-up Procedure which Resulted in Leaving Loose Fibrous Insulation in Containment An NRC-identified, non-cited violation (NCV) of very low safety significance was identified for the failure to follow start-up procedure 1-GOP-1.7, revision 2, Unit Startup, RCS Heat Up from Ambient to HSD, which resulted in leaving loose fibrous insulation in containment (Unit 1).

This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using the IMC 0609, "Significance Determination Process," Phase 1 Worksheets, the finding is determined to have very low safety significance (Green) since it only affected the mitigating systems cornerstone and did not represent a loss of system safety function.

The cause of this finding had cross-cutting aspects associated with work practices of the Human Performance area in that the licensee did not provide the appropriate oversight of contractors conducting the containment walk downs (H.4.c). The finding was entered into the corrective action program as Condition Report 02564. Corrective actions to remove the fibrous material from containment prior to startup and to establish the extent of condition and potential impact on Unit-2 were adequate.

Inspection Report# : 2008002 (pdf)

Significance: Feb 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Use Limiting Case 4160 VAC Bus Frequency and Voltage in Design Calculations The inspectors identified two examples of a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to evaluate variations of emergency diesel generator (EDG) output frequency in electrical design loading calculations, and failure to consider worst case 4160 VAC bus voltage in safety related motor starting calculations. This finding was entered into the licensees corrective action program as condition reports (CR) 091493 and 091494. Planned corrective actions included revision of the EDG loading calculations to incorporate the most limiting voltages and frequencies.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the EDGs to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiencies did not result in any EDG being inoperable based upon additional analysis that showed that the EDGs had sufficient margin to accommodate the increased loading due to worst case acceptably high EDG output frequency; and all safety related motor loads remained operable since they were still capable of starting with the revised worst case low voltage values.

Inspection Report# : 2008006 (pdf)

Significance: Feb 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Appropriate Acceptance Criteria for Testing Battery Voltage at the One Minute Mark The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for incorrect acceptance criteria in test procedure 1-EPT-0106-01, Main Station Battery 1A Service Test. This finding was entered into the licensees corrective action program as condition report 091906. Planned corrective actions included revision of the main station battery test procedures to include the correct voltage at the one minute mark.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the station batteries to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. maintenance and testing procedures. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in station batteries being inoperable based upon a recent review of station battery discharge test results.

The inspectors determined that the lack of a thorough evaluation of condition report 022112, which addressed deficiencies in station battery test procedures such that resolutions addressed causes, was a significant cause of this performance deficiency. Failure to thoroughly evaluate problems such that resolutions address causes is directly related to the Corrective Action Program component of the cross-cutting area of Problem Identification and Resolution and the aspect of thorough evaluation of problems (P.1(c)).

Inspection Report# : 2008006 (pdf)

Significance: Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Emergency Service Water Pump 1-SW-P-1B Inoperable Due to an Inadequate Maintenance Instruction for Reassembly of the ESW strainer A Green self-revealing non-cited violation of Technical Specification 6.4, Unit Operating Procedures and Programs, was identified for failure to have an adequate maintenance procedure for the emergency service water (ESW) pump strainer. This resulted in the emergency service water pump 1-SW-P-1B being declared inoperable. This procedure failed to provide adequate instructions for the reassembly of ESW strainer 1-SW-STR-4B. The finding was documented in the licensees corrective action program as condition report CR023818. Corrective action was taken to restore pump operability and to correct the procedure and post-maintenance test error.

The finding is more than minor, because it is associated with the operability, availability, reliability, or function of a system or train in a mitigating system. This finding was evaluated using the Significance Determination Process and was determined to be of very low safety significance because it did not result in a loss of safety function or the loss of a single train of ESW for more than the allowed Technical Specification outage time. This finding has a cross-cutting aspect in the area of human performance work practices (H.4.a), because personnel proceeded in the face of uncertainty when they continued to re-assemble the strainer operating mechanism without the requisite work instructions.

Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Temporary Fire Suppression Capacity Not Equivalent to Unit 1 Containment Fire Hose Stations A Green NRC-identified non-cited violation of paragraph (a)(1) of 10 CFR 50.48, "Fire Protection," was identified for failure to maintain the fire suppression capability for the Unit 1 containment building as specified by the approved fire protection plan. On October 27, 2007, the licensee failed to provide equivalent fire suppression capacity when the Unit 1 containment fire hose stations were removed from service for repair. This finding was entered into the licensees corrective action program as condition report CR025073. Planned corrective actions included developing equivalent fire suppression capacity determinations for other hose stations.

This finding is more than minor because it was associated with a degradation of a fire protection feature. The finding is of very low safety significance because it involved low degradation of a fixed fire protection system. A significant cause of this finding involved the Decision-Making component of the cross-cutting area of Human Performance and the aspect of making safety-significant or risk-significant decisions

using a systematic process, in that, a formal evaluation was not used to determine equivalent capacity (H.1.a).

Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Carbon Dioxide Suppression System Degraded in Two Fire Areas in Unit 1 and Three Fire Areas in Unit 2 The NRC identified a Green NCV of Unit 1 and Unit 2 Operating License Condition 3.I because the installed carbon dioxide (CO2) fire suppression systems could not be shown to deliver the design basis gas concentration. This finding applied to the Unit 1 and Unit 2 normal switchgear rooms, the Unit 2 cable tunnel, and the Unit 1 and Unit 2 cable vaults. The licensee had implemented or initiated system modifications to address this violation.

The finding is more than minor because it affects the Mitigating Systems cornerstone objective of ensuring reliability and capability of systems that respond to initiating events and the cornerstone attribute of protection against external factors, i.e. fire. The finding was determined to be of very low safety significance in a Significance Determination Process Phase 3 analysis. For the cable vault areas, the analysis showed that fires could initiate scenarios which could challenge the mitigating systems. However, the risk of these scenarios was calculated to be in the very low significance band. Analysis with respect to the normal switchgear rooms led to the conclusion that it was of very low safety significance primarily due to the frequency of fires potentially challenging mitigating systems being relatively low and the availability of unaffected safety-related shutdown systems. The finding for the Unit 2 cable tunnel was also of very low safety significance because it did not have any significant fixed ignition sources (cables were thermoset type) and the probability for transient combustible fires or hot work initiated fires damaging important cables was judged to be low.

Inspection Report# : 2007005 (pdf)

Significance: Oct 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Catqagorize a Maintenance Preventable Functional Failure The inspectors identified an NCV of 10 CFR 50.65 (a)(2) after Surry Power Station failed to categorize the failure of the Unit 2 Charging Pump Component Cooling System as a maintenance preventable functional failure and accordingly, failed to monitor the component as required by 10 CFR 50.65 (a)(1). Condition Report 021045.

The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of the Charging System. Example 7.b in MC 0612, App. E, states that violations of Paragraph 10 CFR 50.65 (a)(2), failure to demonstrate effective control of performance or condition and not putting the affected (SSCs) in (a)(1), are not minor because they necessarily involve degraded SSC performance or condition. The finding is of very low safety significance because the failure to place the system in (a)(1) status did not lead to any further instances of system unreliability or unavailability. The cause of this finding was directly related to the aspect of Training of personnel in the cross-cutting area of human performance (resources component) because the engineer conducting the maintenance rule evaluation, the Maintenance Rule Program Coordinator, and Engineering Supervisor reviewing the evaluation, did not fully understand when to apply the functional failure exemptions. (IMC 0305, H.2.b)

Inspection Report# : 2007008 (pdf)

Significance: Oct 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct Procedures Related to the Operation of the Auxiliary Feedwater System NRC identified non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to revise procedure ECA 0.0, Loss of all AC Power, as corrective action for a condition identified by the licensee that could cause a loss of Net Positive Suction Head (NPSH) to the Turbine Driven Auxiliary Feedwater Pump (TDAFW), and potential damage to the only available feedwater pump during a loss of all AC power event. Other procedures where single AFW pump operation could cause inadequate NPSH had been revised.

The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because no actual safety function was lost. The cause of the finding is related to the cross-cutting area of problem identification and resolution and the aspect of lack of thoroughness of evaluation such that the resolution addresses the causes and extent of conditions. (IMC 0305, P.1.c) [Section 4OA2.a(3)(ii)]

Inspection Report# : 2007008 (pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : November 26, 2008

Surry 1 4Q/2008 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Result in Actuation of Unit 1 Safety Injection Train B.

A Green self-revealing non-cited violation (NCV) of Technical Specification 6.4.A.7 was identified for failure to provide adequate work instructions for corrective maintenance on the safety injection (SI) system. The inadequate work instructions led to an inadvertent actuation of the Unit 1 B train of safety injection on October 29, 2008. Control room operators terminated the invalid actuation within two minutes. The licensee entered the deficiency into the corrective action program for resolution (CR 116664). The proposed corrective actions to establish a response procedure for an inadvertent SI actuation and to provide guidance/restrictions in the work planning process to assure appropriate reviews are obtained, commensurate with the safety significance of the work, appear appropriate.

The finding is greater than minor because it had an actual impact by causing a SI and if left uncorrected could lead to a more significant safety issue. The finding is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance (Green) based on a Phase 3 SDP analyses performed by a regional Senior Reactor Analyst. The analysis was accomplished by increasing the initiating event frequency for a stuck open power operated relief valve (PORV) by one order of magnitude and solving the applicable accident sequences, two of which were Green.

This finding had a cross-cutting aspect in the area of human performance, decision making, because the decision to continue with the planned work was made without a complete understanding of either the effects of the job steps or the worst case possible unintended consequences (H.1(b)). (Section 4OA3)

Inspection Report# : 2008005 (pdf)

Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Review of Vendor Information Led to Unit 1 Manual Reactor Trip.

A Green self-revealing Finding was identified for failure to perform an adequate review of the vendors recommended balance moves for the Unit 1 main turbine. As a result, an improper balance move was made to the Unit 1 main turbine during the April forced outage which caused high turbine vibrations that required the insertion of a manual turbine and reactor trip during the startup on April 20, 2008. A violation of regulatory requirements was not identified.

The licensee entered the deficiency with the work instructions into the corrective action program for resolution (CR 096233). The corrective actions to correct the balance move, implement peer review requirements, and procedural changes that require specifying the detailed location and weight for balance moves with an independent verification appear adequate.

The finding is greater than minor because it had an actual impact, it led to a plant trip, and is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety

functions during shutdown as well as power operations. The finding, evaluated per Attachment 4 of the SDP, screened to very low safety significance (Green) because it did not contribute to both an initiating event and the likelihood of a loss of mitigating equipment or functions.

The cause of the finding is related to the cross-cutting element of human performance work practices. Human error prevention techniques such as peer checks were not invoked by the licensee (H.4(a)). (Section 4OA3)

Inspection Report# : 2008005 (pdf)

Mitigating Systems Significance: Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control for the EDG Ambient Air Temperature Limit.

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion III, Design Control, for a change in the EDG ambient air temperature operating limits, from 100ºF to 120ºF, that was made without an adequate design analysis. The licensee entered the issue into their corrective action program (CAP) for resolution using condition report (CR) 102488.

The inspectors concluded that the licensees failure to perform the necessary analysis to support the increase of the EDG ambient air temperature operating limit from 100ºF to 120ºF was a performance deficiency. The finding, more than minor in accordance with MC 0612, Appendix E, examples 3j and k, is associated with the design control attribute of the Mitigating System Cornerstone. The cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences was adversely affected. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4 the inspectors concluded that the finding is of very low safety significance (Green) because the condition did not represent an actual loss of safety function due to the ambient temperature exceeding 100ºF but not exceeding 105ºF. The finding also was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. A cross-cutting aspect was not assigned to the issue because it is not indicative of recent Inspection Report# : 2008004 (pdf)

Significance: Mar 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Start-up Procedure which Resulted in Leaving Loose Fibrous Insulation in Containment An NRC-identified, non-cited violation (NCV) of very low safety significance was identified for the failure to follow start-up procedure 1-GOP-1.7, revision 2, Unit Startup, RCS Heat Up from Ambient to HSD, which resulted in leaving loose fibrous insulation in containment (Unit 1).

This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using the IMC 0609, "Significance Determination Process," Phase 1 Worksheets, the finding is determined to have very low safety significance (Green) since it only affected the mitigating systems cornerstone and did not represent a loss of system safety function. The cause of this finding had cross-cutting aspects associated with work practices of the Human Performance area in that the licensee did not provide the appropriate oversight of contractors conducting the containment walk downs (H.4.c).

The finding was entered into the corrective action program as Condition Report 02564. Corrective actions to remove the fibrous material from containment prior to startup and to establish the extent of condition and potential impact on Unit-2 were adequate.

Inspection Report# : 2008002 (pdf)

Significance: Feb 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Use Limiting Case 4160 VAC Bus Frequency and Voltage in Design Calculations The inspectors identified two examples of a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to evaluate variations of emergency diesel generator (EDG) output frequency in electrical design loading calculations, and failure to consider worst case 4160 VAC bus voltage in safety related motor starting calculations. This finding was entered into the licensees corrective action program as condition reports (CR) 091493 and 091494. Planned corrective actions included revision of the EDG loading calculations to incorporate the most limiting voltages and frequencies.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the EDGs to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiencies did not result in any EDG being inoperable based upon additional analysis that showed that the EDGs had sufficient margin to accommodate the increased loading due to worst case acceptably high EDG output frequency; and all safety related motor loads remained operable since they were still capable of starting with the revised worst case low voltage values.

Inspection Report# : 2008006 (pdf)

Significance: Feb 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Appropriate Acceptance Criteria for Testing Battery Voltage at the One Minute Mark The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for incorrect acceptance criteria in test procedure 1-EPT-0106-01, Main Station Battery 1A Service Test. This finding was entered into the licensees corrective action program as condition report 091906. Planned corrective actions included revision of the main station battery test procedures to include the correct voltage at the one minute mark.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the station batteries to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. maintenance and testing procedures. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in station batteries being inoperable based upon a recent review of station battery discharge test results.

The inspectors determined that the lack of a thorough evaluation of condition report 022112, which addressed deficiencies in station battery test procedures such that resolutions addressed causes, was a significant cause of this performance deficiency. Failure to thoroughly evaluate problems such that resolutions address causes is directly related to the Corrective Action Program component of the cross-cutting area of Problem Identification and Resolution and the aspect of thorough evaluation of problems (P.1(c)).

Inspection Report# : 2008006 (pdf)

Barrier Integrity

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : April 07, 2009

Surry 1 1Q/2009 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Result in Actuation of Unit 1 Safety Injection Train B.

A Green self-revealing non-cited violation (NCV) of Technical Specification 6.4.A.7 was identified for failure to provide adequate work instructions for corrective maintenance on the safety injection (SI) system. The inadequate work instructions led to an inadvertent actuation of the Unit 1 B train of safety injection on October 29, 2008. Control room operators terminated the invalid actuation within two minutes. The licensee entered the deficiency into the corrective action program for resolution (CR 116664). The proposed corrective actions to establish a response procedure for an inadvertent SI actuation and to provide guidance/restrictions in the work planning process to assure appropriate reviews are obtained, commensurate with the safety significance of the work, appear appropriate.

The finding is greater than minor because it had an actual impact by causing a SI and if left uncorrected could lead to a more significant safety issue. The finding is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance (Green) based on a Phase 3 SDP analyses performed by a regional Senior Reactor Analyst. The analysis was accomplished by increasing the initiating event frequency for a stuck open power operated relief valve (PORV) by one order of magnitude and solving the applicable accident sequences, two of which were Green.

This finding had a cross-cutting aspect in the area of human performance, decision making, because the decision to continue with the planned work was made without a complete understanding of either the effects of the job steps or the worst case possible unintended consequences (H.1(b)). (Section 4OA3)

Inspection Report# : 2008005 (pdf)

Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Review of Vendor Information Led to Unit 1 Manual Reactor Trip.

A Green self-revealing Finding was identified for failure to perform an adequate review of the vendors recommended balance moves for the Unit 1 main turbine. As a result, an improper balance move was made to the Unit 1 main turbine during the April forced outage which caused high turbine vibrations that required the insertion of a manual turbine and reactor trip during the startup on April 20, 2008. A violation of regulatory requirements was not identified.

The licensee entered the deficiency with the work instructions into the corrective action program for resolution (CR 096233). The corrective actions to correct the balance move, implement peer review requirements, and procedural changes that require specifying the detailed location and weight for balance moves with an independent verification appear adequate.

The finding is greater than minor because it had an actual impact, it led to a plant trip, and is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety

functions during shutdown as well as power operations. The finding, evaluated per Attachment 4 of the SDP, screened to very low safety significance (Green) because it did not contribute to both an initiating event and the likelihood of a loss of mitigating equipment or functions.

The cause of the finding is related to the cross-cutting element of human performance work practices. Human error prevention techniques such as peer checks were not invoked by the licensee (H.4(a)). (Section 4OA3)

Inspection Report# : 2008005 (pdf)

Mitigating Systems Significance: Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control for the EDG Ambient Air Temperature Limit.

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion III, Design Control, for a change in the EDG ambient air temperature operating limits, from 100ºF to 120ºF, that was made without an adequate design analysis. The licensee entered the issue into their corrective action program (CAP) for resolution using condition report (CR) 102488.

The inspectors concluded that the licensees failure to perform the necessary analysis to support the increase of the EDG ambient air temperature operating limit from 100ºF to 120ºF was a performance deficiency. The finding, more than minor in accordance with MC 0612, Appendix E, examples 3j and k, is associated with the design control attribute of the Mitigating System Cornerstone. The cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences was adversely affected. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4 the inspectors concluded that the finding is of very low safety significance (Green) because the condition did not represent an actual loss of safety function due to the ambient temperature exceeding 100ºF but not exceeding 105ºF. The finding also was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. A cross-cutting aspect was not assigned to the issue because it is not indicative of recent Inspection Report# : 2008004 (pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : May 28, 2009

Surry 1 2Q/2009 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Result in Actuation of Unit 1 Safety Injection Train B.

A Green self-revealing non-cited violation (NCV) of Technical Specification 6.4.A.7 was identified for failure to provide adequate work instructions for corrective maintenance on the safety injection (SI) system. The inadequate work instructions led to an inadvertent actuation of the Unit 1 B train of safety injection on October 29, 2008. Control room operators terminated the invalid actuation within two minutes. The licensee entered the deficiency into the corrective action program for resolution (CR 116664). The proposed corrective actions to establish a response procedure for an inadvertent SI actuation and to provide guidance/restrictions in the work planning process to assure appropriate reviews are obtained, commensurate with the safety significance of the work, appear appropriate.

The finding is greater than minor because it had an actual impact by causing a SI and if left uncorrected could lead to a more significant safety issue. The finding is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance (Green) based on a Phase 3 SDP analyses performed by a regional Senior Reactor Analyst. The analysis was accomplished by increasing the initiating event frequency for a stuck open power operated relief valve (PORV) by one order of magnitude and solving the applicable accident sequences, two of which were Green.

This finding had a cross-cutting aspect in the area of human performance, decision making, because the decision to continue with the planned work was made without a complete understanding of either the effects of the job steps or the worst case possible unintended consequences (H.1(b)). (Section 4OA3)

Inspection Report# : 2008005 (pdf)

Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Review of Vendor Information Led to Unit 1 Manual Reactor Trip.

A Green self-revealing Finding was identified for failure to perform an adequate review of the vendors recommended balance moves for the Unit 1 main turbine. As a result, an improper balance move was made to the Unit 1 main turbine during the April forced outage which caused high turbine vibrations that required the insertion of a manual turbine and reactor trip during the startup on April 20, 2008. A violation of regulatory requirements was not identified.

The licensee entered the deficiency with the work instructions into the corrective action program for resolution (CR 096233). The corrective actions to correct the balance move, implement peer review requirements, and procedural changes that require specifying the detailed location and weight for balance moves with an independent verification appear adequate.

The finding is greater than minor because it had an actual impact, it led to a plant trip, and is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety

functions during shutdown as well as power operations. The finding, evaluated per Attachment 4 of the SDP, screened to very low safety significance (Green) because it did not contribute to both an initiating event and the likelihood of a loss of mitigating equipment or functions.

The cause of the finding is related to the cross-cutting element of human performance work practices. Human error prevention techniques such as peer checks were not invoked by the licensee (H.4(a)). (Section 4OA3)

Inspection Report# : 2008005 (pdf)

Mitigating Systems Significance: Jun 30, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions for Installation of a Design Change A self-revealing Green non-cited violation of TS 6.4 "Unit Operating Procedures and Programs" was identified for the failure to provide adequate work instructions for installation of design change SU-08-0001, for engine-driven emergency service water pump 1-SW-P-1A. Corrective action to remove the modification from the A pump was completed and reasonable compensatory measures established for all 3 pumps pending removal/alteration of the exhaust piping modificaiton. the licensee entered this issue into the CA program as CR 3337337 The finding associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone, is more than minor because it adversely affected the cornerstone objectivie to ensure the availability, reliability, and operability of 1-SW-P-1A to perform its safety function during a design basis event. Evaluated using a Phase II SDP risk analysis per Appendix A of MC-0609, the finding ws determined to be of very low safety significance (Green) due to availability of the two remainign ESWPs which provided full mitigation capability for the safety functions required.

A cross cutting aspect in the area of human performance work control was assigned to the finding (H.3.a)

Inspection Report# : 2009003 (pdf)

Significance: Jun 26, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Maintenance for Backup Battery for the Halon 1301 System in ESGRs The team identified a performance deficiency and Green NCV for failing to implement a maintenance program for the backup batteries for the Halon 1301 system for the emergency switchgear rooms to ensure on a continuing basis that 24-hour backup power was available as required by the fire protection program (FPP) and Units 1 & 2 Operating License Condition 3.I, Fire Protection. The licensee entered this finding into their corrective action program, and demonstrated that the backup battery had sufficient capacity in the short term until the long term corrective actions can be implemented.

The licensees failure to implement a maintenance program to help ensure that the backup battery for the Halon 1301 system continued to meet its licensing basis requirement of providing backup power for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a performance deficiency. The finding is more than minor because the backup battery actually degraded on several occasions in the past, and the finding is associated with the reactor safety, mitigating systems, cornerstone attribute of protection against external factors, and affected the objective of ensuring reliability and capability of systems that respond to initiating events. The finding was determined to be of very low safety significance because it represented a low degradation of the fixed fire suppression systems. A cross-cutting aspect was not identified in relation to this finding since the cause was not representative of current license performance.

Inspection Report# : 2009007 (pdf)

Significance: Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control for the EDG Ambient Air Temperature Limit.

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion III, Design Control, for a change in the EDG ambient air temperature operating limits, from 100ºF to 120ºF, that was made without an adequate design analysis. The licensee entered the issue into their corrective action program (CAP) for resolution using condition report (CR) 102488.

The inspectors concluded that the licensees failure to perform the necessary analysis to support the increase of the EDG ambient air temperature operating limit from 100ºF to 120ºF was a performance deficiency. The finding, more than minor in accordance with MC 0612, Appendix E, examples 3j and k, is associated with the design control attribute of the Mitigating System Cornerstone. The cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences was adversely affected. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4 the inspectors concluded that the finding is of very low safety significance (Green) because the condition did not represent an actual loss of safety function due to the ambient temperature exceeding 100ºF but not exceeding 105ºF. The finding also was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. A cross-cutting aspect was not assigned to the issue because it is not indicative of recent Inspection Report# : 2008004 (pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous

Last modified : August 31, 2009 Surry 1 3Q/2009 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Result in Actuation of Unit 1 Safety Injection Train B.

A Green self-revealing non-cited violation (NCV) of Technical Specification 6.4.A.7 was identified for failure to provide adequate work instructions for corrective maintenance on the safety injection (SI) system. The inadequate work instructions led to an inadvertent actuation of the Unit 1 B train of safety injection on October 29, 2008. Control room operators terminated the invalid actuation within two minutes. The licensee entered the deficiency into the corrective action program for resolution (CR 116664). The proposed corrective actions to establish a response procedure for an inadvertent SI actuation and to provide guidance/restrictions in the work planning process to assure appropriate reviews are obtained, commensurate with the safety significance of the work, appear appropriate.

The finding is greater than minor because it had an actual impact by causing a SI and if left uncorrected could lead to a more significant safety issue. The finding is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance (Green) based on a Phase 3 SDP analyses performed by a regional Senior Reactor Analyst. The analysis was accomplished by increasing the initiating event frequency for a stuck open power operated relief valve (PORV) by one order of magnitude and solving the applicable accident sequences, two of which were Green.

This finding had a cross-cutting aspect in the area of human performance, decision making, because the decision to continue with the planned work was made without a complete understanding of either the effects of the job steps or the worst case possible unintended consequences (H.1(b)). (Section 4OA3)

Inspection Report# : 2008005 (pdf)

Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Review of Vendor Information Led to Unit 1 Manual Reactor Trip.

A Green self-revealing Finding was identified for failure to perform an adequate review of the vendors recommended balance moves for the Unit 1 main turbine. As a result, an improper balance move was made to the Unit 1 main turbine during the April forced outage which caused high turbine vibrations that required the insertion of a manual turbine and reactor trip during the startup on April 20, 2008. A violation of regulatory requirements was not identified.

The licensee entered the deficiency with the work instructions into the corrective action program for resolution (CR 096233). The corrective actions to correct the balance move, implement peer review requirements, and procedural changes that require specifying the detailed location and weight for balance moves with an independent verification appear adequate.

The finding is greater than minor because it had an actual impact, it led to a plant trip, and is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding, evaluated per Attachment 4 of the SDP, screened

to very low safety significance (Green) because it did not contribute to both an initiating event and the likelihood of a loss of mitigating equipment or functions.

The cause of the finding is related to the cross-cutting element of human performance work practices. Human error prevention techniques such as peer checks were not invoked by the licensee (H.4(a)). (Section 4OA3)

Inspection Report# : 2008005 (pdf)

Mitigating Systems Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Compensatory Measures for the Impairment of Fire Detection Systems The inspectors identified a Green NCV of the Surry operating license, section 3.1 "Fire Protection," for an inadequate procedure that resulted in compensatory continuous fire watches in MERs #3 and #4 being inadequate (CR342078).

Correctie action, revising the requirements for a continuous fire watch, has been implemented. The finding is greater than minor because it is associated with the reactor safety mitigating systems cornerstone attribute to provide protection against external events and adversely affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used MC 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the condition had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for fixed fire protection system outages or degradations. A low degradation rating was assigned to this finding as the provision affected by this finding is expected to display nearly the same level of effectiveness and reliability. Using MC 0609, Appendix F, this finding was determined to be of very low safety significance (Green). A cross-cutting aspect was not assigned to this finding because the performance deficiency for the inadequate procedure occurred long ago and is not a reflection of current performance (Section 1R05)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: FIN Finding Failure to Provide an Adequate Basis for Operability of ESW Pump !-SW-P-1B.

The inspectors identified a Green finding for the incorrect operability determination for emergency service water pump 1-SW-P-1B on August 1, 2009,after vibrations had increased 391% in the vertical plane (CR 343396). A violation of regulatory requirements was not identified. The pump, declared inoperable on August 2, was replaced within the Technical Specificaiton allowed outage time.

The finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated with the equipment performance attribute of the itigation systems cornerstone and adversely affected the cornerstone objective to ensure the availabillity, reliability, and capabilityy of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-609, Attachement 4, "Phase 1-Initial Screening and Characterization of Findings," was determined to be of very low safety significance (Green) because it did not result in a losss of safety function or the loss of a single train of ESW for greater than the allowed outage time. This finding has a cross-cutting aspect in the are of human performance, decision making, because the licensee failed to use conservative assumptions in their operability decision for !-SW-P-1B (H.1.b). (Section 1R15)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC

Item Type: NCV NonCited Violation Inadequate Tornado Protection for Engine Driven Emergency Service Water Pumps 1-SW-P-1A/B/C The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, "Design Control". The design change for the emergency service water pumps (DC-SU-08-0001) was not adequate to protect the diesel-driven emergency service water pumps from damage resulting from a tornado missile as required by the UFSAR (CRs 337720, 337337, 341557). Pending resolution, interim compenatory measues have been established to provide assurance the pumps will be capable of performing their safety function.

The finding, associated with the design control attribute of the mitigation systems cornerstone, is more than minor because it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that evaluated per MC-0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," was determinted to be of very low safety significance (Green) because of the extremely low initiationg event frequency for a tornado. A phase III risk analysis was performed because the finding screened potentially risk significant for a severe weathr initiating event. This finding has a cross-cutting aspect in the area of human performance resources, because the licensee's design documentation for DC SU-08-0001 and ET-S-08-0032 was not complete and accurate which led to the installation of inadequate modifications on ESWPs 1-SW-P-1A/1B/1C (H.2.c)(Section 1R18)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Work Instructions Lead to Packing Failure of ESW Pump 1-SW-P-1B.

A self-revealing Green NCV of Technical Specification 6.4, "Unit Operating Procedures and Programs," was identified for the failure to provide adequate work instructions for maintenance on 1-SW-P-1B, a safety-related component, which led to failure of the pump's packing gland on August 26, 2009, and required the pump be removed from service and repacked (CR 346268).

The finding is associated with the equipment performance attribute of the mitigation systems cornerstone and is more than minor because it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequenses. The finding, evaluated per MC-0609, , "Phase 1- Initial Screening and Characterization of Findings" was determined to be of very low safety significance (green) because it did not result in a loss of safety function or loss of a single train of ESW for more than its allowed outage time. Tthis finding has a cross-cutting aspect in the area of human performance, resources, in that a complete and accurate procedure was not available to assure nuclear safety during replacement of 1-Sw-P-1B (H.2.c)

(Section 1R19)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct a Ground on Safety Bus 1H A green NCV of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," was identified by the inspectors for failure to promptly identify and correct a condition adverse to quality related to a ground on emergency safety bus 1H.

This resulted in the degraded condition being allowed to exist for 72 days prior to de-energizing the containment recircution fan and correcting the adverse condition (CR 336041).

This finding is more than minor because it adversely impacted the equipment performance attribute of the reactor safety mitigation system cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," was determined to be of very low safety significance (Green). The finding screened to a phase II assessment on the assumption that a second ground would result in a complete loss of the safety bus and its safety function. The phase II analysis was performed for the core damage sequence "Loss of a 4.16Kv Bus (1J or 1H)" utilizing an increased initiating event likelihood (IEL) value of 1 due to the degraded conditon of the 1H bus. The duration of the degraded conditon was 72 days. The

finding was not greater than Green because full mitigation capability of the opposite train remained available. This finding has a cross cutting aspect in human performance, decision making, in that the licensee did not use conservative assumption in their decision making process and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than to demonstrate that it is unsafe in order to disapprove the action of continuing to operate with a ground on the 1H emergency bus (H.1.b) (Section 4OA2).

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Action for ELU Performance Deficiencies The inspectors identified a Green NCV of Surry Operating Licenses, Section 3.1 "Fire Protection," for failure to promptly identify and correct a condition adverse to fire protection in regard to Appendix R emergency lighting unit performance failures due to inadequate configuration control of the emergency light's defeat switch. Failure to reposition the switch following maintenance and or inadvertent switch manipulation has over time led to numerous Appendix R emergency lights being discovered non-functional. Corrective action to address the failure to restore the switch following maintenance has been taken and actions to prevent inadvertent manipulation are being evaluated (CR 352214).

The finding is more than minor becaue it adversely affected the external factors attribute (fire) of the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability and availability of the emergency lighting units (ELUs) was affected. The finding, evaluated per MC-0609, Appendix F, "Fire Protection Significance Determination Process," was determined to be of very low safety significance (Green). The finding affected post fire safe shutdown and was assigned a low degradation rating because the issue did not have a significant impact on safe shutdown operations because there was not a simultaneous wide spread failure of the ELUs. This finding has a cross-cutting aspect in the area of problem identification and resolution, because the licensee did not take adequate corrective action in a timely mannr to address an adverse trend in ELU functionality (P.1.d). (Section 4OA2)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Remove Blocking Device From Piping Supports The inspectors identified a Green NCV of Technical Specification 6.4, "Unit Operating Procedures," associated with blocking devices not being removed from piping supports following maintenance due to procedure issues related to procedure adequacy and adherence. The blocking devices were removed upon discovery and appropriate corrective action established to address the issue (ACE017736).

the finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated wit the procedure quality attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding evaluated per MC-0609, , "Phase 1-Initial Screening and Characterization of Findings," was determined to be of very low safety significance (Green) because operability of a safety system, though challenged, was never lost. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee's corrective actions were not effective in identifying additional blocked spring hangers on safety-related systems or preventing further configuration control issues assocatied with spring hanger blocking devices (P.1.d). (Section 4OA2)

Inspection Report# : 2009004 (pdf)

Significance: Jun 30, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions for Installation of a Design Change

A self-revealing Green non-cited violation of TS 6.4 "Unit Operating Procedures and Programs" was identified for the failure to provide adequate work instructions for installation of design change SU-08-0001, for engine-driven emergency service water pump 1-SW-P-1A. Corrective action to remove the modification from the A pump was completed and reasonable compensatory measures established for all 3 pumps pending removal/alteration of the exhaust piping modificaiton. the licensee entered this issue into the CA program as CR 3337337 The finding associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone, is more than minor because it adversely affected the cornerstone objectivie to ensure the availability, reliability, and operability of 1-SW-P-1A to perform its safety function during a design basis event. Evaluated using a Phase II SDP risk analysis per Appendix A of MC-0609, the finding was determined to be of very low safety significance (Green) due to availability of the two remainign ESWPs which provided full mitigation capability for the safety functions required.

A cross cutting aspect in the area of human performance work control was assigned to the finding (H.3.a)

Inspection Report# : 2009003 (pdf)

Significance: Jun 26, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Maintenance for Backup Battery for the Halon 1301 System in ESGRs The team identified a performance deficiency and Green NCV for failing to implement a maintenance program for the backup batteries for the Halon 1301 system for the emergency switchgear rooms to ensure on a continuing basis that 24-hour backup power was available as required by the fire protection program (FPP) and Units 1 & 2 Operating License Condition 3.I, Fire Protection. The licensee entered this finding into their corrective action program, and demonstrated that the backup battery had sufficient capacity in the short term until the long term corrective actions can be implemented.

The licensees failure to implement a maintenance program to help ensure that the backup battery for the Halon 1301 system continued to meet its licensing basis requirement of providing backup power for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a performance deficiency. The finding is more than minor because the backup battery actually degraded on several occasions in the past, and the finding is associated with the reactor safety, mitigating systems, cornerstone attribute of protection against external factors, and affected the objective of ensuring reliability and capability of systems that respond to initiating events. The finding was determined to be of very low safety significance because it represented a low degradation of the fixed fire suppression systems. A cross-cutting aspect was not identified in relation to this finding since the cause was not representative of current license performance.

Inspection Report# : 2009007 (pdf)

Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : December 10, 2009

Surry 1 4Q/2009 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Compensatory Measures for the Impairment of Fire Detection Systems The inspectors identified a Green NCV of the Surry operating license, section 3.1 "Fire Protection," for an inadequate procedure that resulted in compensatory continuous fire watches in MERs #3 and #4 being inadequate (CR342078).

Correctie action, revising the requirements for a continuous fire watch, has been implemented. The finding is greater than minor because it is associated with the reactor safety mitigating systems cornerstone attribute to provide protection against external events and adversely affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used MC 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the condition had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for fixed fire protection system outages or degradations. A low degradation rating was assigned to this finding as the provision affected by this finding is expected to display nearly the same level of effectiveness and reliability. Using MC 0609, Appendix F, this finding was determined to be of very low safety significance (Green). A cross-cutting aspect was not assigned to this finding because the performance deficiency for the inadequate procedure occurred long ago and is not a reflection of current performance (Section 1R05)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: FIN Finding Failure to Provide an Adequate Basis for Operability of ESW Pump !-SW-P-1B.

The inspectors identified a Green finding for the incorrect operability determination for emergency service water pump 1-SW-P-1B on August 1, 2009,after vibrations had increased 391% in the vertical plane (CR 343396). A violation of regulatory requirements was not identified. The pump, declared inoperable on August 2, was replaced within the Technical Specificaiton allowed outage time.

The finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated with the equipment performance attribute of the itigation systems cornerstone and adversely affected the cornerstone objective to ensure the availabillity, reliability, and capabilityy of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-609, Attachement 4, "Phase 1-Initial Screening and Characterization of Findings," was determined to be of very low safety significance (Green) because it did not result in a losss of safety function or the loss of a single train of ESW for greater than the allowed outage time. This finding has a cross-cutting aspect in the are of human performance, decision making, because the licensee failed to use conservative assumptions in their operability decision for !-SW-P-1B (H.1.b). (Section 1R15)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC

Item Type: NCV NonCited Violation Inadequate Tornado Protection for Engine Driven Emergency Service Water Pumps 1-SW-P-1A/B/C The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, "Design Control". The design change for the emergency service water pumps (DC-SU-08-0001) was not adequate to protect the diesel-driven emergency service water pumps from damage resulting from a tornado missile as required by the UFSAR (CRs 337720, 337337, 341557). Pending resolution, interim compenatory measues have been established to provide assurance the pumps will be capable of performing their safety function.

The finding, associated with the design control attribute of the mitigation systems cornerstone, is more than minor because it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that evaluated per MC-0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," was determinted to be of very low safety significance (Green) because of the extremely low initiationg event frequency for a tornado. A phase III risk analysis was performed because the finding screened potentially risk significant for a severe weathr initiating event. This finding has a cross-cutting aspect in the area of human performance resources, because the licensee's design documentation for DC SU-08-0001 and ET-S-08-0032 was not complete and accurate which led to the installation of inadequate modifications on ESWPs 1-SW-P-1A/1B/1C (H.2.c)(Section 1R18)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Lead to Packing Failure of ESW Pump 1-SW-P-1B.

A self-revealing Green NCV of Technical Specification 6.4, "Unit Operating Procedures and Programs," was identified for the failure to provide adequate work instructions for maintenance on 1-SW-P-1B, a safety-related component, which led to failure of the pump's packing gland on August 26, 2009, and required the pump be removed from service and repacked (CR 346268).

The finding is associated with the equipment performance attribute of the mitigation systems cornerstone and is more than minor because it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequenses. The finding, evaluated per MC-0609, , "Phase 1- Initial Screening and Characterization of Findings" was determined to be of very low safety significance (green) because it did not result in a loss of safety function or loss of a single train of ESW for more than its allowed outage time. Tthis finding has a cross-cutting aspect in the area of human performance, resources, in that a complete and accurate procedure was not available to assure nuclear safety during replacement of 1-Sw-P-1B (H.2.c)

(Section 1R19)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct a Ground on Safety Bus 1H A green NCV of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," was identified by the inspectors for failure to promptly identify and correct a condition adverse to quality related to a ground on emergency safety bus 1H.

This resulted in the degraded condition being allowed to exist for 72 days prior to de-energizing the containment recircution fan and correcting the adverse condition (CR 336041).

This finding is more than minor because it adversely impacted the equipment performance attribute of the reactor safety mitigation system cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," was determined to be of very low safety significance (Green). The finding screened to a phase II assessment on the assumption that a second ground would result in a complete loss of the safety bus and its safety function. The phase II analysis was performed for the core damage sequence "Loss of a 4.16Kv Bus (1J or 1H)" utilizing an increased initiating event likelihood (IEL) value of 1 due to the degraded conditon of the 1H bus. The duration of the degraded conditon was 72 days. The

finding was not greater than Green because full mitigation capability of the opposite train remained available. This finding has a cross cutting aspect in human performance, decision making, in that the licensee did not use conservative assumption in their decision making process and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than to demonstrate that it is unsafe in order to disapprove the action of continuing to operate with a ground on the 1H emergency bus (H.1.b) (Section 4OA2).

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Action for ELU Performance Deficiencies The inspectors identified a Green NCV of Surry Operating Licenses, Section 3.1 "Fire Protection," for failure to promptly identify and correct a condition adverse to fire protection in regard to Appendix R emergency lighting unit performance failures due to inadequate configuration control of the emergency light's defeat switch. Failure to reposition the switch following maintenance and or inadvertent switch manipulation has over time led to numerous Appendix R emergency lights being discovered non-functional. Corrective action to address the failure to restore the switch following maintenance has been taken and actions to prevent inadvertent manipulation are being evaluated (CR 352214).

The finding is more than minor becaue it adversely affected the external factors attribute (fire) of the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability and availability of the emergency lighting units (ELUs) was affected. The finding, evaluated per MC-0609, Appendix F, "Fire Protection Significance Determination Process," was determined to be of very low safety significance (Green). The finding affected post fire safe shutdown and was assigned a low degradation rating because the issue did not have a significant impact on safe shutdown operations because there was not a simultaneous wide spread failure of the ELUs. This finding has a cross-cutting aspect in the area of problem identification and resolution, because the licensee did not take adequate corrective action in a timely mannr to address an adverse trend in ELU functionality (P.1.d). (Section 4OA2)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Remove Blocking Device From Piping Supports The inspectors identified a Green NCV of Technical Specification 6.4, "Unit Operating Procedures," associated with blocking devices not being removed from piping supports following maintenance due to procedure issues related to procedure adequacy and adherence. The blocking devices were removed upon discovery and appropriate corrective action established to address the issue (ACE017736).

the finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated wit the procedure quality attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding evaluated per MC-0609, , "Phase 1-Initial Screening and Characterization of Findings," was determined to be of very low safety significance (Green) because operability of a safety system, though challenged, was never lost. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee's corrective actions were not effective in identifying additional blocked spring hangers on safety-related systems or preventing further configuration control issues assocatied with spring hanger blocking devices (P.1.d). (Section 4OA2)

Inspection Report# : 2009004 (pdf)

Significance: Jun 30, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions for Installation of a Design Change

A self-revealing Green non-cited violation of TS 6.4 "Unit Operating Procedures and Programs" was identified for the failure to provide adequate work instructions for installation of design change SU-08-0001, for engine-driven emergency service water pump 1-SW-P-1A. Corrective action to remove the modification from the A pump was completed and reasonable compensatory measures established for all 3 pumps pending removal/alteration of the exhaust piping modificaiton. the licensee entered this issue into the CA program as CR 3337337 The finding associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone, is more than minor because it adversely affected the cornerstone objectivie to ensure the availability, reliability, and operability of 1-SW-P-1A to perform its safety function during a design basis event. Evaluated using a Phase II SDP risk analysis per Appendix A of MC-0609, the finding was determined to be of very low safety significance (Green) due to availability of the two remainign ESWPs which provided full mitigation capability for the safety functions required.

A cross cutting aspect in the area of human performance work control was assigned to the finding (H.3.a)

Inspection Report# : 2009003 (pdf)

Significance: Jun 26, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Maintenance for Backup Battery for the Halon 1301 System in ESGRs The team identified a performance deficiency and Green NCV for failing to implement a maintenance program for the backup batteries for the Halon 1301 system for the emergency switchgear rooms to ensure on a continuing basis that 24-hour backup power was available as required by the fire protection program (FPP) and Units 1 & 2 Operating License Condition 3.I, Fire Protection. The licensee entered this finding into their corrective action program, and demonstrated that the backup battery had sufficient capacity in the short term until the long term corrective actions can be implemented.

The licensees failure to implement a maintenance program to help ensure that the backup battery for the Halon 1301 system continued to meet its licensing basis requirement of providing backup power for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a performance deficiency. The finding is more than minor because the backup battery actually degraded on several occasions in the past, and the finding is associated with the reactor safety, mitigating systems, cornerstone attribute of protection against external factors, and affected the objective of ensuring reliability and capability of systems that respond to initiating events. The finding was determined to be of very low safety significance because it represented a low degradation of the fixed fire suppression systems. A cross-cutting aspect was not identified in relation to this finding since the cause was not representative of current license performance.

Inspection Report# : 2009007 (pdf)

Barrier Integrity Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inoperability of MCR isolation Damper 1-VS-MOD-103D due to failure to promptly identify and correct internal hydraulic leakage A self-revealing Green NCV of 10 CFR 50 Appendix B, Criterion XVI, was identified for the failure to correct a condition adverse to quality which led to main control room isolation damper 1-VS-MOD-103D being inoperable for approximately 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months on September 21-22, 2009 (CR 349075). The actuator was repaired and is scheduled for replacement in 2010.

The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective, as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated per

MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a crosscutting aspect in the area of human performance, resources, in that equipment and other resources were not made available to assure nuclear safety by minimizing preventative maintenance deferrals (H.2.a).

Inspection Report# : 2009005 (pdf)

Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to perform an adequate operability determination for main control room isolation damper 1-VS-MOD-103D A self-revealing Green Finding was identified for the incorrect operability determination of main control room isolation damper 1-VS-MOD-103D. The damper, declared operable and left in-service following loss of power to its hydraulic pump on September 21, 2009 (CR 349003), failed to close on demand, on September 22, 2009. The damper was inoperable for approximately 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months (CR 349075) before power was restored to the pump, the damper closed, and the actuator repaired.

The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated per MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or the loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a cross-cutting aspect in the area of problem identification, corrective action program, in that an adequate operability assessment that thoroughly evaluated the degraded condition of 1-VS-MOD-103D was not performed (P.1.c).

Inspection Report# : 2009005 (pdf)

Significance: Oct 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate Effective Preventive Maintenance of Safety Injection Check Valves nor Set Goals and Monitor under 10CFR50.65(a)(1)

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants, for failure to demonstrate effective preventive maintenance of Unit 1 low head safety injection (LHSI) cold leg check valves in accordance with 10CFR50.65(a)(2) and not establish goals and monitor against those goals in accordance with 10CFR50.65(a)(1).

The finding is more than minor because it affected the Barrier Integrity cornerstone objective of providing reasonable assurance that physical design barriers (e.g., reactor coolant system (RCS)) protect the public from radionuclide releases caused by accidents or events. Specifically, the finding affected the LHSI cold leg check valves, which provide an isolation barrier from the high pressure RCS when the SI System is in standby to ensure that the integrity of the reactor RCS boundary is maintained. The finding is also associated with the cornerstone attribute of reactor coolant system equipment and barrier performance. The inspectors determined that this performance deficiency was a separate consequence of the degraded performance associated with the LHSI cold leg check valves. Because of this characterization, the inspectors determined that this issue should not be processed through the Significance Determination Process. Therefore, in accordance with the guidance in NRC Inspection Procedure 71111.12, Appendix D, this issue was determined to be a maintenance rule Category II finding and is of very low safety significance (Green). Based on the assessment performed by the team on the current licensees implementation of 10CFR50.65, the results of the licensees extent of condition review for this finding, and because this finding occurred on November 18, 2007, the team determined that this finding was not indicative of current licensee performance and, therefore, no Cross Cutting Aspect was assigned to this issue. This issue was entered in the licensees CAP as CR02560. The licensee restored compliance by establishing goals and monitoring the system performance against those goals in accordance with 10CFR50.65(a)(1). (Section 4OA2.a(3)i)

Inspection Report# : 2009006 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : March 01, 2010

Surry 1 1Q/2010 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a non-conservative error in the quarterly TS surveillance for the Unit 1 A battery The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action for failure to identify that a non-conservative error had been introduced into the Unit 1 A main station battery quarterly technical specification surveillance procedure (CR 366388). The licensee modified the procedure to eliminate the non-conservative error.

The inspectors determined the failure to identify a non-conservative error which was introduced into the TS quarterly surveillance procedure following the replacement of individual battery cells, was a condition adverse to quality and a performance deficiency which was reasonably within the licensees ability to foresee and correct, and should have been prevented. The finding was more than minor because if left uncorrected the non-conservative error in 1-EPT-0103-01 would have the potential to lead to a more significant safety concern. Specifically, this is because the error was large enough to mask cell degradation and an inoperable cell. The finding was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of the safety related 125 VDC station batteries that provide class 1E backup power to risk significant components needed to prevent undesirable consequences during a loss of offsite power event. The finding was evaluated using MC-0609, , Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because operability of the Unit 1 A battery was not lost and the error was removed prior to the next quarterly surveillance. This finding had a cross cutting aspect in the area of problem identification and resolution because the licensee did not evaluate and communicate relevant external OE, including vendor recommendations, to affected internal stakeholders in a timely manner (P.2(a)). Specifically, the caveat to have cells on a float charge for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months was not fully evaluated when the battery cells were replaced.

Inspection Report# : 2010002 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Compensatory Measures for the Impairment of Fire Detection Systems The inspectors identified a Green NCV of the Surry operating license, section 3.1 "Fire Protection," for an inadequate procedure that resulted in compensatory continuous fire watches in MERs #3 and #4 being inadequate (CR342078).

Correctie action, revising the requirements for a continuous fire watch, has been implemented. The finding is greater than minor because it is associated with the reactor safety mitigating systems cornerstone attribute to provide protection against external events and adversely affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used MC 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the condition had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for fixed fire protection system outages or degradations. A low degradation rating was assigned to this finding as the provision affected by this finding is expected to display nearly the same level of effectiveness and reliability. Using MC 0609, Appendix F, this finding was determined to be of very low safety

significance (Green). A cross-cutting aspect was not assigned to this finding because the performance deficiency for the inadequate procedure occurred long ago and is not a reflection of current performance (Section 1R05)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: FIN Finding Failure to Provide an Adequate Basis for Operability of ESW Pump !-SW-P-1B.

The inspectors identified a Green finding for the incorrect operability determination for emergency service water pump 1-SW-P-1B on August 1, 2009,after vibrations had increased 391% in the vertical plane (CR 343396). A violation of regulatory requirements was not identified. The pump, declared inoperable on August 2, was replaced within the Technical Specificaiton allowed outage time.

The finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated with the equipment performance attribute of the itigation systems cornerstone and adversely affected the cornerstone objective to ensure the availabillity, reliability, and capabilityy of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-609, Attachement 4, "Phase 1-Initial Screening and Characterization of Findings," was determined to be of very low safety significance (Green) because it did not result in a losss of safety function or the loss of a single train of ESW for greater than the allowed outage time. This finding has a cross-cutting aspect in the are of human performance, decision making, because the licensee failed to use conservative assumptions in their operability decision for !-SW-P-1B (H.1.b). (Section 1R15)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Tornado Protection for Engine Driven Emergency Service Water Pumps 1-SW-P-1A/B/C The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, "Design Control". The design change for the emergency service water pumps (DC-SU-08-0001) was not adequate to protect the diesel-driven emergency service water pumps from damage resulting from a tornado missile as required by the UFSAR (CRs 337720, 337337, 341557). Pending resolution, interim compenatory measues have been established to provide assurance the pumps will be capable of performing their safety function.

The finding, associated with the design control attribute of the mitigation systems cornerstone, is more than minor because it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that evaluated per MC-0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," was determinted to be of very low safety significance (Green) because of the extremely low initiationg event frequency for a tornado. A phase III risk analysis was performed because the finding screened potentially risk significant for a severe weathr initiating event. This finding has a cross-cutting aspect in the area of human performance resources, because the licensee's design documentation for DC SU-08-0001 and ET-S-08-0032 was not complete and accurate which led to the installation of inadequate modifications on ESWPs 1-SW-P-1A/1B/1C (H.2.c)(Section 1R18)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Lead to Packing Failure of ESW Pump 1-SW-P-1B.

A self-revealing Green NCV of Technical Specification 6.4, "Unit Operating Procedures and Programs," was identified for the failure to provide adequate work instructions for maintenance on 1-SW-P-1B, a safety-related component, which led to failure of the pump's packing gland on August 26, 2009, and required the pump be removed from service and repacked (CR 346268).

The finding is associated with the equipment performance attribute of the mitigation systems cornerstone and is more than minor because it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequenses. The finding, evaluated per MC-0609, , "Phase 1- Initial Screening and Characterization of Findings" was determined to be of very low safety significance (green) because it did not result in a loss of safety function or loss of a single train of ESW for more than its allowed outage time. Tthis finding has a cross-cutting aspect in the area of human performance, resources, in that a complete and accurate procedure was not available to assure nuclear safety during replacement of 1-Sw-P-1B (H.2.c)

(Section 1R19)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct a Ground on Safety Bus 1H A green NCV of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," was identified by the inspectors for failure to promptly identify and correct a condition adverse to quality related to a ground on emergency safety bus 1H.

This resulted in the degraded condition being allowed to exist for 72 days prior to de-energizing the containment recircution fan and correcting the adverse condition (CR 336041).

This finding is more than minor because it adversely impacted the equipment performance attribute of the reactor safety mitigation system cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," was determined to be of very low safety significance (Green). The finding screened to a phase II assessment on the assumption that a second ground would result in a complete loss of the safety bus and its safety function. The phase II analysis was performed for the core damage sequence "Loss of a 4.16Kv Bus (1J or 1H)" utilizing an increased initiating event likelihood (IEL) value of 1 due to the degraded conditon of the 1H bus. The duration of the degraded conditon was 72 days. The finding was not greater than Green because full mitigation capability of the opposite train remained available. This finding has a cross cutting aspect in human performance, decision making, in that the licensee did not use conservative assumption in their decision making process and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than to demonstrate that it is unsafe in order to disapprove the action of continuing to operate with a ground on the 1H emergency bus (H.1.b) (Section 4OA2).

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Action for ELU Performance Deficiencies The inspectors identified a Green NCV of Surry Operating Licenses, Section 3.1 "Fire Protection," for failure to promptly identify and correct a condition adverse to fire protection in regard to Appendix R emergency lighting unit performance failures due to inadequate configuration control of the emergency light's defeat switch. Failure to reposition the switch following maintenance and or inadvertent switch manipulation has over time led to numerous Appendix R emergency lights being discovered non-functional. Corrective action to address the failure to restore the switch following maintenance has been taken and actions to prevent inadvertent manipulation are being evaluated (CR 352214).

The finding is more than minor becaue it adversely affected the external factors attribute (fire) of the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability and availability of the emergency lighting units (ELUs) was affected. The finding, evaluated per MC-0609, Appendix F, "Fire Protection Significance Determination Process," was determined to be of very low safety significance (Green). The finding affected post fire safe shutdown and was assigned a low degradation rating because the issue did not have a significant impact on safe shutdown operations because there was not a simultaneous wide spread failure of the ELUs. This finding has a cross-cutting aspect in the area of problem identification and resolution, because the licensee did not take adequate corrective action in a timely mannr to address an adverse trend in ELU functionality (P.1.d). (Section 4OA2)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Remove Blocking Device From Piping Supports The inspectors identified a Green NCV of Technical Specification 6.4, "Unit Operating Procedures," associated with blocking devices not being removed from piping supports following maintenance due to procedure issues related to procedure adequacy and adherence. The blocking devices were removed upon discovery and appropriate corrective action established to address the issue (ACE017736).

the finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated wit the procedure quality attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding evaluated per MC-0609, , "Phase 1-Initial Screening and Characterization of Findings," was determined to be of very low safety significance (Green) because operability of a safety system, though challenged, was never lost. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee's corrective actions were not effective in identifying additional blocked spring hangers on safety-related systems or preventing further configuration control issues assocatied with spring hanger blocking devices (P.1.d). (Section 4OA2)

Inspection Report# : 2009004 (pdf)

Significance: Jun 30, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions for Installation of a Design Change A self-revealing Green non-cited violation of TS 6.4 "Unit Operating Procedures and Programs" was identified for the failure to provide adequate work instructions for installation of design change SU-08-0001, for engine-driven emergency service water pump 1-SW-P-1A. Corrective action to remove the modification from the A pump was completed and reasonable compensatory measures established for all 3 pumps pending removal/alteration of the exhaust piping modificaiton. the licensee entered this issue into the CA program as CR 3337337 The finding associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone, is more than minor because it adversely affected the cornerstone objectivie to ensure the availability, reliability, and operability of 1-SW-P-1A to perform its safety function during a design basis event. Evaluated using a Phase II SDP risk analysis per Appendix A of MC-0609, the finding was determined to be of very low safety significance (Green) due to availability of the two remainign ESWPs which provided full mitigation capability for the safety functions required.

A cross cutting aspect in the area of human performance work control was assigned to the finding (H.3.a)

Inspection Report# : 2009003 (pdf)

Significance: Jun 26, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Maintenance for Backup Battery for the Halon 1301 System in ESGRs The team identified a performance deficiency and Green NCV for failing to implement a maintenance program for the backup batteries for the Halon 1301 system for the emergency switchgear rooms to ensure on a continuing basis that 24-hour backup power was available as required by the fire protection program (FPP) and Units 1 & 2 Operating License Condition 3.I, Fire Protection. The licensee entered this finding into their corrective action program, and demonstrated that the backup battery had sufficient capacity in the short term until the long term corrective actions can be implemented.

The licensees failure to implement a maintenance program to help ensure that the backup battery for the Halon 1301 system continued to meet its licensing basis requirement of providing backup power for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a performance deficiency. The finding is more than minor because the backup battery actually degraded on several occasions in the past, and the finding is associated with the reactor safety, mitigating systems, cornerstone attribute of protection against external factors, and affected the objective of ensuring reliability and capability of systems that respond to initiating events. The finding was determined to be of very low safety significance because it represented a low degradation of the fixed fire suppression systems. A cross-cutting aspect was not identified in relation to this finding since the cause was not representative of current license performance.

Inspection Report# : 2009007 (pdf)

Barrier Integrity Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inoperability of MCR isolation Damper 1-VS-MOD-103D due to failure to promptly identify and correct internal hydraulic leakage A self-revealing Green NCV of 10 CFR 50 Appendix B, Criterion XVI, was identified for the failure to correct a condition adverse to quality which led to main control room isolation damper 1-VS-MOD-103D being inoperable for approximately 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months on September 21-22, 2009 (CR 349075). The actuator was repaired and is scheduled for replacement in 2010.

The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective, as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated per MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a crosscutting aspect in the area of human performance, resources, in that equipment and other resources were not made available to assure nuclear safety by minimizing preventative maintenance deferrals (H.2.a).

Inspection Report# : 2009005 (pdf)

Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to perform an adequate operability determination for main control room isolation damper 1-VS-MOD-103D A self-revealing Green Finding was identified for the incorrect operability determination of main control room isolation damper 1-VS-MOD-103D. The damper, declared operable and left in-service following loss of power to its hydraulic pump on September 21, 2009 (CR 349003), failed to close on demand, on September 22, 2009. The damper was inoperable for approximately 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months (CR 349075) before power was restored to the pump, the damper closed, and the actuator repaired.

The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated per MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or the loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a cross-cutting aspect in the area of problem identification, corrective action program, in that an adequate operability assessment that thoroughly evaluated the degraded condition of 1-VS-MOD-103D was not performed (P.1.c).

Inspection Report# : 2009005 (pdf)

Significance: Oct 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate Effective Preventive Maintenance of Safety Injection Check Valves nor Set Goals and Monitor under 10CFR50.65(a)(1)

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants, for failure to demonstrate effective preventive maintenance of Unit 1 low head safety injection (LHSI) cold leg check valves in accordance with 10CFR50.65(a)(2) and not establish goals and monitor against those goals in accordance with 10CFR50.65(a)(1).

The finding is more than minor because it affected the Barrier Integrity cornerstone objective of providing reasonable assurance that physical design barriers (e.g., reactor coolant system (RCS)) protect the public from radionuclide releases caused by accidents or events. Specifically, the finding affected the LHSI cold leg check valves, which provide an isolation barrier from the high pressure RCS when the SI System is in standby to ensure that the integrity of the reactor RCS boundary is maintained. The finding is also associated with the cornerstone attribute of reactor coolant system equipment and barrier performance. The inspectors determined that this performance deficiency was a separate consequence of the degraded performance associated with the LHSI cold leg check valves. Because of this characterization, the inspectors determined that this issue should not be processed through the Significance Determination Process. Therefore, in accordance with the guidance in NRC Inspection Procedure 71111.12, Appendix D, this issue was determined to be a maintenance rule Category II finding and is of very low safety significance (Green). Based on the assessment performed by the team on the current licensees implementation of 10CFR50.65, the results of the licensees extent of condition review for this finding, and because this finding occurred on November 18, 2007, the team determined that this finding was not indicative of current licensee performance and, therefore, no Cross Cutting Aspect was assigned to this issue. This issue was entered in the licensees CAP as CR02560. The licensee restored compliance by establishing goals and monitoring the system performance against those goals in accordance with 10CFR50.65(a)(1). (Section 4OA2.a(3)i)

Inspection Report# : 2009006 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : May 26, 2010

Surry 1 2Q/2010 Plant Inspection Findings Initiating Events Significance: Jun 30, 2010 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Rigging Practices Result in Damage to Safety Related Equipment A self-revealing Green Finding was identified for failure to adequately rig a 300 pound motor in the auxiliary building in accordance with the manufacturers recommendations on May 11, 2010. As a result, the motor slipped from its rigging and dropped approximately 15 feet onto the A component cooling water (CCW) pump motor below, damaging the motors cabling and electrical junction box. The CCW pump was declared inoperable (CR 380834), the damage was repaired, and the CCW pump restored to an operable status on May 15, 2010.

Inspectors determined that the failure to implement adequate rigging practices in accordance with vendor recommendations as required by procedure MA-AA-101, Revision 5, Fleet Lifting and Material Handling constituted a performance deficiency and a finding which was reasonably within the licensees ability to foresee and correct and which should have been prevented. The finding is similar to MC 0612, Appendix E example 4.f, and is more than minor because it resulted in damage to and inoperability of a risk significant component. The finding is associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events which upset plant stability and challenge critical safety functions during shutdown as well as power operations because a loss of the component cooling water system would have resulted in a unit transient. The finding, evaluated per Attachment 4 of MC-0609, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a plant transient and the loss of accident mitigation equipment. This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not make safety/risk significant decisions using a systematic process, especially when faced with uncertain decisions, to ensure safety is maintained (H.1(a)). Specifically, the rigging team made safety/risk significant decisions within lifting/rigging procedures that did not include a systematic process for evaluating each lift, especially loads <5000 lbs in the vicinity of risk significant equipment.

Inspection Report# : 2010003 (pdf)

Mitigating Systems Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate that the Reliability of Systems or Components were effectively controlled per 10 CFR 50.65 (a)(2)

The NRC identified a Green Non-Cited Violation of 10CFR50.65 a(2) for the licensees failure to demonstrate that the reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by was being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function. Specifically, the licensees MR program would not demonstrate that a system should remain in category a(2) as defined by regulatory requirements.

The inspectors determined the licensees MR program could not demonstrate that reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by were being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function is a performance deficiency. Specifically, the monitoring established by the

license did not effectively demonstrate that systems in a(2) were being effectively controlled through performance of appropriate preventative maintenance. This masking of poor equipment performance does not allow the licensee to determine if a system should be in increased monitoring of a(1).

The finding was more than minor because it adversely affected the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of HSS and LSS systems to perform their functions when required. Specifically, multiple HSS and LSS systems could have a high probability of failure, because poor equipment performance would not be recognized by the licensee. This could prevent a poor performing system from being placed into the a(1) category when required and appropriate corrective action would not be taken.

The finding was evaluated using MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green), because the finding did not involve an actual failure of equipment. This finding had a crosscutting aspect in the area of human performance and resources because the licensee did not ensure that personnel, procedures, and other resources were available and adequate to assure proper implementation of MR program. The MR personnel did not have the training required to implement the program within the required industry regulations and guidelines (H.2.b).

Inspection Report# : 2010003 (pdf)

Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a non-conservative error in the quarterly TS surveillance for the Unit 1 A battery The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action for failure to identify that a non-conservative error had been introduced into the Unit 1 A main station battery quarterly technical specification surveillance procedure (CR 366388). The licensee modified the procedure to eliminate the non-conservative error.

The inspectors determined the failure to identify a non-conservative error which was introduced into the TS quarterly surveillance procedure following the replacement of individual battery cells, was a condition adverse to quality and a performance deficiency which was reasonably within the licensees ability to foresee and correct, and should have been prevented. The finding was more than minor because if left uncorrected the non-conservative error in 1-EPT-0103-01 would have the potential to lead to a more significant safety concern. Specifically, this is because the error was large enough to mask cell degradation and an inoperable cell. The finding was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of the safety related 125 VDC station batteries that provide class 1E backup power to risk significant components needed to prevent undesirable consequences during a loss of offsite power event. The finding was evaluated using MC-0609, , Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because operability of the Unit 1 A battery was not lost and the error was removed prior to the next quarterly surveillance. This finding had a cross cutting aspect in the area of problem identification and resolution because the licensee did not evaluate and communicate relevant external OE, including vendor recommendations, to affected internal stakeholders in a timely manner (P.2(a)). Specifically, the caveat to have cells on a float charge for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months was not fully evaluated when the battery cells were replaced.

Inspection Report# : 2010002 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Compensatory Measures for the Impairment of Fire Detection Systems The inspectors identified a Green NCV of the Surry operating license, section 3.1 "Fire Protection," for an inadequate procedure that resulted in compensatory continuous fire watches in MERs #3 and #4 being inadequate (CR342078).

Correctie action, revising the requirements for a continuous fire watch, has been implemented. The finding is greater than minor because it is associated with the reactor safety mitigating systems cornerstone attribute to provide

protection against external events and adversely affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used MC 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the condition had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for fixed fire protection system outages or degradations. A low degradation rating was assigned to this finding as the provision affected by this finding is expected to display nearly the same level of effectiveness and reliability. Using MC 0609, Appendix F, this finding was determined to be of very low safety significance (Green). A cross-cutting aspect was not assigned to this finding because the performance deficiency for the inadequate procedure occurred long ago and is not a reflection of current performance (Section 1R05)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: FIN Finding Failure to Provide an Adequate Basis for Operability of ESW Pump !-SW-P-1B.

The inspectors identified a Green finding for the incorrect operability determination for emergency service water pump 1-SW-P-1B on August 1, 2009,after vibrations had increased 391% in the vertical plane (CR 343396). A violation of regulatory requirements was not identified. The pump, declared inoperable on August 2, was replaced within the Technical Specificaiton allowed outage time.

The finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated with the equipment performance attribute of the itigation systems cornerstone and adversely affected the cornerstone objective to ensure the availabillity, reliability, and capabilityy of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-609, Attachement 4, "Phase 1-Initial Screening and Characterization of Findings," was determined to be of very low safety significance (Green) because it did not result in a losss of safety function or the loss of a single train of ESW for greater than the allowed outage time. This finding has a cross-cutting aspect in the are of human performance, decision making, because the licensee failed to use conservative assumptions in their operability decision for !-SW-P-1B (H.1.b). (Section 1R15)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Tornado Protection for Engine Driven Emergency Service Water Pumps 1-SW-P-1A/B/C The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, "Design Control". The design change for the emergency service water pumps (DC-SU-08-0001) was not adequate to protect the diesel-driven emergency service water pumps from damage resulting from a tornado missile as required by the UFSAR (CRs 337720, 337337, 341557). Pending resolution, interim compenatory measues have been established to provide assurance the pumps will be capable of performing their safety function.

The finding, associated with the design control attribute of the mitigation systems cornerstone, is more than minor because it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that evaluated per MC-0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," was determinted to be of very low safety significance (Green) because of the extremely low initiationg event frequency for a tornado. A phase III risk analysis was performed because the finding screened potentially risk significant for a severe weathr initiating event. This finding has a cross-cutting aspect in the area of human performance resources, because the licensee's design documentation for DC SU-08-0001 and ET-S-08-0032 was not complete and accurate which led to the installation of inadequate modifications on ESWPs 1-SW-P-1A/1B/1C (H.2.c)(Section 1R18)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: Self-Revealing

Item Type: NCV NonCited Violation Inadequate Work Instructions Lead to Packing Failure of ESW Pump 1-SW-P-1B.

A self-revealing Green NCV of Technical Specification 6.4, "Unit Operating Procedures and Programs," was identified for the failure to provide adequate work instructions for maintenance on 1-SW-P-1B, a safety-related component, which led to failure of the pump's packing gland on August 26, 2009, and required the pump be removed from service and repacked (CR 346268).

The finding is associated with the equipment performance attribute of the mitigation systems cornerstone and is more than minor because it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequenses. The finding, evaluated per MC-0609, , "Phase 1- Initial Screening and Characterization of Findings" was determined to be of very low safety significance (green) because it did not result in a loss of safety function or loss of a single train of ESW for more than its allowed outage time. Tthis finding has a cross-cutting aspect in the area of human performance, resources, in that a complete and accurate procedure was not available to assure nuclear safety during replacement of 1-Sw-P-1B (H.2.c)

(Section 1R19)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct a Ground on Safety Bus 1H A green NCV of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," was identified by the inspectors for failure to promptly identify and correct a condition adverse to quality related to a ground on emergency safety bus 1H.

This resulted in the degraded condition being allowed to exist for 72 days prior to de-energizing the containment recircution fan and correcting the adverse condition (CR 336041).

This finding is more than minor because it adversely impacted the equipment performance attribute of the reactor safety mitigation system cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," was determined to be of very low safety significance (Green). The finding screened to a phase II assessment on the assumption that a second ground would result in a complete loss of the safety bus and its safety function. The phase II analysis was performed for the core damage sequence "Loss of a 4.16Kv Bus (1J or 1H)" utilizing an increased initiating event likelihood (IEL) value of 1 due to the degraded conditon of the 1H bus. The duration of the degraded conditon was 72 days. The finding was not greater than Green because full mitigation capability of the opposite train remained available. This finding has a cross cutting aspect in human performance, decision making, in that the licensee did not use conservative assumption in their decision making process and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than to demonstrate that it is unsafe in order to disapprove the action of continuing to operate with a ground on the 1H emergency bus (H.1.b) (Section 4OA2).

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Action for ELU Performance Deficiencies The inspectors identified a Green NCV of Surry Operating Licenses, Section 3.1 "Fire Protection," for failure to promptly identify and correct a condition adverse to fire protection in regard to Appendix R emergency lighting unit performance failures due to inadequate configuration control of the emergency light's defeat switch. Failure to reposition the switch following maintenance and or inadvertent switch manipulation has over time led to numerous Appendix R emergency lights being discovered non-functional. Corrective action to address the failure to restore the switch following maintenance has been taken and actions to prevent inadvertent manipulation are being evaluated (CR 352214).

The finding is more than minor becaue it adversely affected the external factors attribute (fire) of the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating

events to prevent undesirable consequences. Specifically, the reliability and availability of the emergency lighting units (ELUs) was affected. The finding, evaluated per MC-0609, Appendix F, "Fire Protection Significance Determination Process," was determined to be of very low safety significance (Green). The finding affected post fire safe shutdown and was assigned a low degradation rating because the issue did not have a significant impact on safe shutdown operations because there was not a simultaneous wide spread failure of the ELUs. This finding has a cross-cutting aspect in the area of problem identification and resolution, because the licensee did not take adequate corrective action in a timely mannr to address an adverse trend in ELU functionality (P.1.d). (Section 4OA2)

Inspection Report# : 2009004 (pdf)

Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Remove Blocking Device From Piping Supports The inspectors identified a Green NCV of Technical Specification 6.4, "Unit Operating Procedures," associated with blocking devices not being removed from piping supports following maintenance due to procedure issues related to procedure adequacy and adherence. The blocking devices were removed upon discovery and appropriate corrective action established to address the issue (ACE017736).

the finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated wit the procedure quality attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding evaluated per MC-0609, , "Phase 1-Initial Screening and Characterization of Findings," was determined to be of very low safety significance (Green) because operability of a safety system, though challenged, was never lost. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee's corrective actions were not effective in identifying additional blocked spring hangers on safety-related systems or preventing further configuration control issues assocatied with spring hanger blocking devices (P.1.d). (Section 4OA2)

Inspection Report# : 2009004 (pdf)

Barrier Integrity Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inoperability of MCR isolation Damper 1-VS-MOD-103D due to failure to promptly identify and correct internal hydraulic leakage A self-revealing Green NCV of 10 CFR 50 Appendix B, Criterion XVI, was identified for the failure to correct a condition adverse to quality which led to main control room isolation damper 1-VS-MOD-103D being inoperable for approximately 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months on September 21-22, 2009 (CR 349075). The actuator was repaired and is scheduled for replacement in 2010.

The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective, as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated per MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a crosscutting aspect in the area of human performance, resources, in that equipment and other resources were not made available to assure nuclear safety by minimizing preventative maintenance deferrals (H.2.a).

Inspection Report# : 2009005 (pdf)

Significance: Dec 31, 2009

Identified By: Self-Revealing Item Type: FIN Finding Failure to perform an adequate operability determination for main control room isolation damper 1-VS-MOD-103D A self-revealing Green Finding was identified for the incorrect operability determination of main control room isolation damper 1-VS-MOD-103D. The damper, declared operable and left in-service following loss of power to its hydraulic pump on September 21, 2009 (CR 349003), failed to close on demand, on September 22, 2009. The damper was inoperable for approximately 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months (CR 349075) before power was restored to the pump, the damper closed, and the actuator repaired.

The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated per MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or the loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a cross-cutting aspect in the area of problem identification, corrective action program, in that an adequate operability assessment that thoroughly evaluated the degraded condition of 1-VS-MOD-103D was not performed (P.1.c).

Inspection Report# : 2009005 (pdf)

Significance: Oct 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate Effective Preventive Maintenance of Safety Injection Check Valves nor Set Goals and Monitor under 10CFR50.65(a)(1)

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants, for failure to demonstrate effective preventive maintenance of Unit 1 low head safety injection (LHSI) cold leg check valves in accordance with 10CFR50.65(a)(2) and not establish goals and monitor against those goals in accordance with 10CFR50.65(a)(1).

The finding is more than minor because it affected the Barrier Integrity cornerstone objective of providing reasonable assurance that physical design barriers (e.g., reactor coolant system (RCS)) protect the public from radionuclide releases caused by accidents or events. Specifically, the finding affected the LHSI cold leg check valves, which provide an isolation barrier from the high pressure RCS when the SI System is in standby to ensure that the integrity of the reactor RCS boundary is maintained. The finding is also associated with the cornerstone attribute of reactor coolant system equipment and barrier performance. The inspectors determined that this performance deficiency was a separate consequence of the degraded performance associated with the LHSI cold leg check valves. Because of this characterization, the inspectors determined that this issue should not be processed through the Significance Determination Process. Therefore, in accordance with the guidance in NRC Inspection Procedure 71111.12, Appendix D, this issue was determined to be a maintenance rule Category II finding and is of very low safety significance (Green). Based on the assessment performed by the team on the current licensees implementation of 10CFR50.65, the results of the licensees extent of condition review for this finding, and because this finding occurred on November 18, 2007, the team determined that this finding was not indicative of current licensee performance and, therefore, no Cross Cutting Aspect was assigned to this issue. This issue was entered in the licensees CAP as CR02560. The licensee restored compliance by establishing goals and monitoring the system performance against those goals in accordance with 10CFR50.65(a)(1). (Section 4OA2.a(3)i)

Inspection Report# : 2009006 (pdf)

Emergency Preparedness

Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : September 02, 2010

Surry 1 3Q/2010 Plant Inspection Findings Initiating Events Significance: Jun 30, 2010 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Rigging Practices Result in Damage to Safety Related Equipment A self-revealing Green Finding was identified for failure to adequately rig a 300 pound motor in the auxiliary building in accordance with the manufacturers recommendations on May 11, 2010. As a result, the motor slipped from its rigging and dropped approximately 15 feet onto the A component cooling water (CCW) pump motor below, damaging the motors cabling and electrical junction box. The CCW pump was declared inoperable (CR 380834), the damage was repaired, and the CCW pump restored to an operable status on May 15, 2010.

Inspectors determined that the failure to implement adequate rigging practices in accordance with vendor recommendations as required by procedure MA-AA-101, Revision 5, Fleet Lifting and Material Handling constituted a performance deficiency and a finding which was reasonably within the licensees ability to foresee and correct and which should have been prevented. The finding is similar to MC 0612, Appendix E example 4.f, and is more than minor because it resulted in damage to and inoperability of a risk significant component. The finding is associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events which upset plant stability and challenge critical safety functions during shutdown as well as power operations because a loss of the component cooling water system would have resulted in a unit transient. The finding, evaluated per Attachment 4 of MC-0609, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a plant transient and the loss of accident mitigation equipment. This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not make safety/risk significant decisions using a systematic process, especially when faced with uncertain decisions, to ensure safety is maintained (H.1(a)). Specifically, the rigging team made safety/risk significant decisions within lifting/rigging procedures that did not include a systematic process for evaluating each lift, especially loads <5000 lbs in the vicinity of risk significant equipment.

Inspection Report# : 2010003 (pdf)

Mitigating Systems Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate that the Reliability of Systems or Components were effectively controlled per 10 CFR 50.65 (a)(2)

The NRC identified a Green Non-Cited Violation of 10CFR50.65 a(2) for the licensees failure to demonstrate that the reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by was being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function. Specifically, the licensees MR program would not demonstrate that a system should remain in category a(2) as defined by regulatory requirements.

The inspectors determined the licensees MR program could not demonstrate that reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by were being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function is a performance deficiency. Specifically, the monitoring established by the

license did not effectively demonstrate that systems in a(2) were being effectively controlled through performance of appropriate preventative maintenance. This masking of poor equipment performance does not allow the licensee to determine if a system should be in increased monitoring of a(1).

The finding was more than minor because it adversely affected the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of HSS and LSS systems to perform their functions when required. Specifically, multiple HSS and LSS systems could have a high probability of failure, because poor equipment performance would not be recognized by the licensee. This could prevent a poor performing system from being placed into the a(1) category when required and appropriate corrective action would not be taken.

The finding was evaluated using MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green), because the finding did not involve an actual failure of equipment. This finding had a crosscutting aspect in the area of human performance and resources because the licensee did not ensure that personnel, procedures, and other resources were available and adequate to assure proper implementation of MR program. The MR personnel did not have the training required to implement the program within the required industry regulations and guidelines (H.2.b).

Inspection Report# : 2010003 (pdf)

Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a non-conservative error in the quarterly TS surveillance for the Unit 1 A battery The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action for failure to identify that a non-conservative error had been introduced into the Unit 1 A main station battery quarterly technical specification surveillance procedure (CR 366388). The licensee modified the procedure to eliminate the non-conservative error.

The inspectors determined the failure to identify a non-conservative error which was introduced into the TS quarterly surveillance procedure following the replacement of individual battery cells, was a condition adverse to quality and a performance deficiency which was reasonably within the licensees ability to foresee and correct, and should have been prevented. The finding was more than minor because if left uncorrected the non-conservative error in 1-EPT-0103-01 would have the potential to lead to a more significant safety concern. Specifically, this is because the error was large enough to mask cell degradation and an inoperable cell. The finding was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of the safety related 125 VDC station batteries that provide class 1E backup power to risk significant components needed to prevent undesirable consequences during a loss of offsite power event. The finding was evaluated using MC-0609, , Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because operability of the Unit 1 A battery was not lost and the error was removed prior to the next quarterly surveillance. This finding had a cross cutting aspect in the area of problem identification and resolution because the licensee did not evaluate and communicate relevant external OE, including vendor recommendations, to affected internal stakeholders in a timely manner (P.2(a)). Specifically, the caveat to have cells on a float charge for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months was not fully evaluated when the battery cells were replaced.

Inspection Report# : 2010002 (pdf)

Barrier Integrity Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inoperability of MCR isolation Damper 1-VS-MOD-103D due to failure to promptly identify and correct

internal hydraulic leakage A self-revealing Green NCV of 10 CFR 50 Appendix B, Criterion XVI, was identified for the failure to correct a condition adverse to quality which led to main control room isolation damper 1-VS-MOD-103D being inoperable for approximately 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months on September 21-22, 2009 (CR 349075). The actuator was repaired and is scheduled for replacement in 2010.

The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective, as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated per MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a crosscutting aspect in the area of human performance, resources, in that equipment and other resources were not made available to assure nuclear safety by minimizing preventative maintenance deferrals (H.2.a).

Inspection Report# : 2009005 (pdf)

Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to perform an adequate operability determination for main control room isolation damper 1-VS-MOD-103D A self-revealing Green Finding was identified for the incorrect operability determination of main control room isolation damper 1-VS-MOD-103D. The damper, declared operable and left in-service following loss of power to its hydraulic pump on September 21, 2009 (CR 349003), failed to close on demand, on September 22, 2009. The damper was inoperable for approximately 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months (CR 349075) before power was restored to the pump, the damper closed, and the actuator repaired.

The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated per MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or the loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a cross-cutting aspect in the area of problem identification, corrective action program, in that an adequate operability assessment that thoroughly evaluated the degraded condition of 1-VS-MOD-103D was not performed (P.1.c).

Inspection Report# : 2009005 (pdf)

Significance: Oct 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate Effective Preventive Maintenance of Safety Injection Check Valves nor Set Goals and Monitor under 10CFR50.65(a)(1)

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants, for failure to demonstrate effective preventive maintenance of Unit 1 low head safety injection (LHSI) cold leg check valves in accordance with 10CFR50.65(a)(2) and not establish goals and monitor against those goals in accordance with 10CFR50.65(a)(1).

The finding is more than minor because it affected the Barrier Integrity cornerstone objective of providing reasonable assurance that physical design barriers (e.g., reactor coolant system (RCS)) protect the public from radionuclide releases caused by accidents or events. Specifically, the finding affected the LHSI cold leg check valves, which provide an isolation barrier from the high pressure RCS when the SI System is in standby to ensure that the integrity of the reactor RCS boundary is maintained. The finding is also associated with the cornerstone attribute of reactor coolant system equipment and barrier performance. The inspectors determined that this performance deficiency was a separate consequence of the degraded performance associated with the LHSI cold leg check valves. Because of this characterization, the inspectors determined that this issue should not be processed

through the Significance Determination Process. Therefore, in accordance with the guidance in NRC Inspection Procedure 71111.12, Appendix D, this issue was determined to be a maintenance rule Category II finding and is of very low safety significance (Green). Based on the assessment performed by the team on the current licensees implementation of 10CFR50.65, the results of the licensees extent of condition review for this finding, and because this finding occurred on November 18, 2007, the team determined that this finding was not indicative of current licensee performance and, therefore, no Cross Cutting Aspect was assigned to this issue. This issue was entered in the licensees CAP as CR02560. The licensee restored compliance by establishing goals and monitoring the system performance against those goals in accordance with 10CFR50.65(a)(1). (Section 4OA2.a(3)i)

Inspection Report# : 2009006 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : November 29, 2010

Surry 1 4Q/2010 Plant Inspection Findings Initiating Events Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadvertent Actuation of Safety Injection A self-revealing Green NCV of TS 6.4, Unit Operating Procedures and Programs, was identified for the failure to follow procedure 1-OPT-ZZ-001, ESF Actuation with Undervoltage and Degraded Voltage 1H Bus. Specifically, on October 26, 2010, a test lead was incorrectly installed in the Unit 1 relay room for the logic circuit associated with the A train of Consequence Limiting Safeguards (CLS). This resulted in an inadvertent safety injection, isolated component cooling water supply to the standby residual heat removal (RHR) train, and automatically initiated several safety-related components including emergency diesel generator (EDG) #1. Operators entered AP-10.20, Response To Spurious Safety Injection With RCS Temperature Less Than 350°F, and terminated the safety injection in approximately three minutes. The licensee entered this issue into the CAP (CR 400908).

Failure to install the test leads as required by procedure 1-OPT-ZZ-001, is a performance deficiency. The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstones objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding, evaluated in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 3, identified the finding is of very low safety significance (Green) because the finding did not lead to a loss of decay heat removal. This finding has a cross cutting aspect in the work practices component in the Human Performance area, because human error prevention techniques were not properly used commensurate with the risk significance of the assigned task (H.4(a)).

Inspection Report# : 2010005 (pdf)

Significance: Aug 16, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Degraded Unit 1 Nuclear Instrument RC Filters An NRC identified non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, was identified for the licensees failure to identify and correct degraded RC filters associated with Unit 1 Nuclear Instrument (NI) cabinets for N-42 and N-44 based on a similar degraded condition identified on Unit 2 NI cabinet N-43 in November 2009. The issue was entered into the licensees corrective action program as condition report CR383881.

All the RC filters in the Surry Unit 1 and 2 NI cabinets have been replaced with new RC filters.

The finding was determined to be of more than minor significance because it is associated with the equipment performance attribute of the Initiating Events cornerstone. It adversely affected the cornerstone objective of protection against external events, i.e., fire. The performance deficiency was screened using phase 1 of the Significance Determination Process (SDP) and was determined to be a fire initiator contributor and to have impact on post fire safe shutdown, therefore a phase 2 analysis utilizing Inspection Manual chapter 0609 Appendix F was required. Since the finding involved MCR fire scenarios, a phase 3 analysis was required. A phase 3 risk analysis was performed by a regional SRA in accordance with IMC 0609 Appendix F, NUREG/CR6850, and utilizing the latest Surry SPAR probabilistic risk analysis model. The fire scenarios were determined to impact MCR operator actions but would not credibly require MCR evacuation for either habitability or safe shutdown functional requirements. The dominant sequence was a fire induced reactor trip transient initiator, with failures of auxiliary feedwater, main feedwater and failure to implement feed and bleed leading to core damage. Factors which mitigated the risk of the fire were the minimal fire growth potential and the potential for NI cabinet fires to damage SSD equipment. The risk evaluation result was an increase of <1E-6 for core damage frequency, a finding of very low risk significance (Green). This finding involved the cross cutting area of problem identification and resolution, the component of operating experience (OE), and the aspect of evaluating

internal OE (P.2.a), because the licensee did not effectively evaluate the internal operating experience gained from the November 2009 RC filter failure prior to the failure of the RC filters on June 8, 2010.

Inspection Report# : 2010006 (pdf)

Significance: Jun 30, 2010 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Rigging Practices Result in Damage to Safety Related Equipment A self-revealing Green Finding was identified for failure to adequately rig a 300 pound motor in the auxiliary building in accordance with the manufacturers recommendations on May 11, 2010. As a result, the motor slipped from its rigging and dropped approximately 15 feet onto the A component cooling water (CCW) pump motor below, damaging the motors cabling and electrical junction box. The CCW pump was declared inoperable (CR 380834), the damage was repaired, and the CCW pump restored to an operable status on May 15, 2010.

Inspectors determined that the failure to implement adequate rigging practices in accordance with vendor recommendations as required by procedure MA-AA-101, Revision 5, Fleet Lifting and Material Handling constituted a performance deficiency and a finding which was reasonably within the licensees ability to foresee and correct and which should have been prevented. The finding is similar to MC 0612, Appendix E example 4.f, and is more than minor because it resulted in damage to and inoperability of a risk significant component. The finding is associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events which upset plant stability and challenge critical safety functions during shutdown as well as power operations because a loss of the component cooling water system would have resulted in a unit transient. The finding, evaluated per Attachment 4 of MC-0609, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a plant transient and the loss of accident mitigation equipment. This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not make safety/risk significant decisions using a systematic process, especially when faced with uncertain decisions, to ensure safety is maintained (H.1(a)). Specifically, the rigging team made safety/risk significant decisions within lifting/rigging procedures that did not include a systematic process for evaluating each lift, especially loads <5000 lbs in the vicinity of risk significant equipment.

Inspection Report# : 2010003 (pdf)

Mitigating Systems Significance: TBD Dec 31, 2010 Identified By: NRC Item Type: AV Apparent Violation Failure to Correct Multiple Conditions Adverse to Fire Protection A self-revealing apparent violation (AV) of Condition 1.B to the Surry Unit 1 and Unit 2 Updated Facility Operating Licenses, DPR-32 and DPR-37, was identified for the licensees failure to take corrective action for degraded conditions adverse to the fire protection program. Specifically, in 2003-2004, three breakers with loads including the Unit 2 1B Refueling Water Storage Tank (RWST) chiller motor, the Unit 1 2B charging component cooling water pump, and the Unit 2 B hydrogen recombiner were identified as being oversized with respect to the Surry design standard for breaker sizing and cable protection. The failure to take corrective action on the affected breakers led to a fault on the Unit 2 RWST Chiller Motor 1B on October 11, 2010, and a resulting fire which damaged the electrical cable and motor controller. The fire was promptly extinguished by the fire brigade. The licensee entered this issue into the CAP (CR 398628) and isolated the remaining breakers to prevent additional failures.

The inspectors found that the failure to take action to correct multiple oversized breakers constituted a performance deficiency. The finding is more than minor because it adversely affected the external factors attribute (fire) of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the Unit 2 1B RWST chiller motor and the Unit 2 B hydrogen recombiner breakers were the most susceptible to fire due to their size; also a cable fault could

potentially damage safety related cables routed nearby. In addition, the Unit 1 2B charging component cooling water pump is safety related and was also unprotected. The inspectors reviewed IMC 0609, Appendix F, Attachment 1, and determined the category of post fire safe shutdown was affected and the finding required a phase 3 analysis. The significance of this finding is to be determined pending completion of the phase 3 evaluation. This finding has a cross cutting aspect in the work control component in the Human Performance area because the licensee did not appropriately plan work activities by incorporating risk insights. Specifically, although work orders were planned in 2006 they were neither prioritized consistent with their safety significance nor scheduled and completed in a timely manner. (H.3(a)).

Inspection Report# : 2010005 (pdf)

Significance: TBD Dec 31, 2010 Identified By: NRC Item Type: AV Apparent Violation Inadequate Risk Evaluation for Leaving Common ESGR HELB Door Open A licensee identified AV of 10CFR50.65 (a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, was revealed after the licensee discovered that 2-BS-DR-21, common emergency switchgear room (ESGR) door was blocked open for two hours without clear communication to licensed operators. The licensee did not adequately assess the increase in operational risk that resulted in the required risk management actions of fire and environmentally qualified watches not being established. The licensee immediately corrected the condition by shutting the HELB door and having security control personnel access. The issue was entered into the licensees CAP as CR397720.

The failure to adequately assess the increased risk associated with blocking open the common ESGR door and to take the required risk management actions is a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, both Unit 1 and Unit 2 plant risk were not evaluated and risk management activities were not put in place when the common ESGR door was blocked open for maintenance and unable to perform its function as a fire barrier, a halon suppression pressure boundary, a main control room pressure boundary, and a HELB boundary. In accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, this finding will require a phase 3 analysis. The significance of this finding is to be determined pending completion of the phase 3 evaluation. The inspectors determined that this finding had a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately plan work activities by incorporating risk insights (H.3(a)).

Inspection Report# : 2010005 (pdf)

Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate that the Reliability of Systems or Components were effectively controlled per 10 CFR 50.65 (a)(2)

The NRC identified a Green Non-Cited Violation of 10CFR50.65 a(2) for the licensees failure to demonstrate that the reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by was being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function. Specifically, the licensees MR program would not demonstrate that a system should remain in category a(2) as defined by regulatory requirements.

The inspectors determined the licensees MR program could not demonstrate that reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by were being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function is a performance deficiency. Specifically, the monitoring established by the license did not effectively demonstrate that systems in a(2) were being effectively controlled through performance of appropriate preventative maintenance. This masking of poor equipment performance does not allow the licensee to determine if a system should be in increased monitoring of a(1).

The finding was more than minor because it adversely affected the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of HSS and LSS systems to perform their functions when required. Specifically, multiple HSS and LSS systems could have a high probability of failure, because poor equipment performance would not be recognized by the licensee. This could prevent a poor performing system from being placed into the a(1) category when required and appropriate corrective action would not be taken.

The finding was evaluated using MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green), because the finding did not involve an actual failure of equipment. This finding had a crosscutting aspect in the area of human performance and resources because the licensee did not ensure that personnel, procedures, and other resources were available and adequate to assure proper implementation of MR program. The MR personnel did not have the training required to implement the program within the required industry regulations and guidelines (H.2.b).

Inspection Report# : 2010003 (pdf)

Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a non-conservative error in the quarterly TS surveillance for the Unit 1 A battery The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action for failure to identify that a non-conservative error had been introduced into the Unit 1 A main station battery quarterly technical specification surveillance procedure (CR 366388). The licensee modified the procedure to eliminate the non-conservative error.

The inspectors determined the failure to identify a non-conservative error which was introduced into the TS quarterly surveillance procedure following the replacement of individual battery cells, was a condition adverse to quality and a performance deficiency which was reasonably within the licensees ability to foresee and correct, and should have been prevented. The finding was more than minor because if left uncorrected the non-conservative error in 1-EPT-0103-01 would have the potential to lead to a more significant safety concern. Specifically, this is because the error was large enough to mask cell degradation and an inoperable cell. The finding was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of the safety related 125 VDC station batteries that provide class 1E backup power to risk significant components needed to prevent undesirable consequences during a loss of offsite power event. The finding was evaluated using MC-0609, , Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because operability of the Unit 1 A battery was not lost and the error was removed prior to the next quarterly surveillance. This finding had a cross cutting aspect in the area of problem identification and resolution because the licensee did not evaluate and communicate relevant external OE, including vendor recommendations, to affected internal stakeholders in a timely manner (P.2(a)). Specifically, the caveat to have cells on a float charge for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months was not fully evaluated when the battery cells were replaced.

Inspection Report# : 2010002 (pdf)

Barrier Integrity Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Heavy Load Lift of the 1B RCP Motor Over Exposed Reactor Fuel An NRC-identified Green NCV of Technical Specification (TS) 6.4, Unit Operating Procedures and Programs, was identified. Personnel failed to follow the defined heavy load shipping path inside containment as specified in procedure, GMP-001, Heavy Load Rigging and Movement, which resulted in the movement of the 1B reactor coolant pump motor over exposed reactor fuel. The licensee has entered the issue into the CAP (CR 404106).

Transport of the 1B reactor coolant pump motor over the exposed reactor core is a performance deficiency. The finding is more than minor because it can reasonably be viewed as a precursor to a significant event, the heavy load traveled over exposed irradiated fuel with the reactor vessel head removed. In accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, the inspectors conducted a Phase 1 SDP screening and determined the finding required a Phase 2 analysis. The Phase 2 analysis determined the finding is of very low safety significance (Green) because: (1) there is a low probability of dropping the load based on a study in NUREG-1774 performed for crane operating experience; (2) the polar crane was in working condition and had no known deficiencies that would have affected the cranes ability to lift the load; and, (3) the duration of the heavy load lift over the exposed reactor core was very short. In addition, in accordance with NRC IMC 0609, Appendix H, Containment Integrity SDP, the finding would not contribute to LERF due to the time since the reactor was shutdown. The finding has a cross-cutting aspect in the work practices component of the Human Performance area because plant supervisors failed to properly supervise workers executing procedure steps (H.4(c)).

Inspection Report# : 2010005 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : March 03, 2011

Surry 1 1Q/2011 Plant Inspection Findings Initiating Events Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadvertent Actuation of Safety Injection A self-revealing Green NCV of TS 6.4, Unit Operating Procedures and Programs, was identified for the failure to follow procedure 1-OPT-ZZ-001, ESF Actuation with Undervoltage and Degraded Voltage 1H Bus. Specifically, on October 26, 2010, a test lead was incorrectly installed in the Unit 1 relay room for the logic circuit associated with the A train of Consequence Limiting Safeguards (CLS). This resulted in an inadvertent safety injection, isolated component cooling water supply to the standby residual heat removal (RHR) train, and automatically initiated several safety-related components including emergency diesel generator (EDG) #1. Operators entered AP-10.20, Response To Spurious Safety Injection With RCS Temperature Less Than 350°F, and terminated the safety injection in approximately three minutes. The licensee entered this issue into the CAP (CR 400908).

Failure to install the test leads as required by procedure 1-OPT-ZZ-001, is a performance deficiency. The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstones objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding, evaluated in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 3, identified the finding is of very low safety significance (Green) because the finding did not lead to a loss of decay heat removal. This finding has a cross cutting aspect in the work practices component in the Human Performance area, because human error prevention techniques were not properly used commensurate with the risk significance of the assigned task (H.4(a)).

Inspection Report# : 2010005 (pdf)

Significance: Aug 16, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Degraded Unit 1 Nuclear Instrument RC Filters An NRC identified non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, was identified for the licensees failure to identify and correct degraded RC filters associated with Unit 1 Nuclear Instrument (NI) cabinets for N-42 and N-44 based on a similar degraded condition identified on Unit 2 NI cabinet N-43 in November 2009. The issue was entered into the licensees corrective action program as condition report CR383881.

All the RC filters in the Surry Unit 1 and 2 NI cabinets have been replaced with new RC filters.

The finding was determined to be of more than minor significance because it is associated with the equipment performance attribute of the Initiating Events cornerstone. It adversely affected the cornerstone objective of protection against external events, i.e., fire. The performance deficiency was screened using phase 1 of the Significance Determination Process (SDP) and was determined to be a fire initiator contributor and to have impact on post fire safe shutdown, therefore a phase 2 analysis utilizing Inspection Manual chapter 0609 Appendix F was required. Since the finding involved MCR fire scenarios, a phase 3 analysis was required. A phase 3 risk analysis was performed by a regional SRA in accordance with IMC 0609 Appendix F, NUREG/CR6850, and utilizing the latest Surry SPAR probabilistic risk analysis model. The fire scenarios were determined to impact MCR operator actions but would not credibly require MCR evacuation for either habitability or safe shutdown functional requirements. The dominant sequence was a fire induced reactor trip transient initiator, with failures of auxiliary feedwater, main feedwater and failure to implement feed and bleed leading to core damage. Factors which mitigated the risk of the fire were the minimal fire growth potential and the potential for NI cabinet fires to damage SSD equipment. The risk evaluation result was an increase of <1E-6 for core damage frequency, a finding of very low risk significance (Green). This finding involved the cross cutting area of problem identification and resolution, the component of operating experience (OE), and the aspect of evaluating

internal OE (P.2.a), because the licensee did not effectively evaluate the internal operating experience gained from the November 2009 RC filter failure prior to the failure of the RC filters on June 8, 2010.

Inspection Report# : 2010006 (pdf)

Significance: Jun 30, 2010 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Rigging Practices Result in Damage to Safety Related Equipment A self-revealing Green Finding was identified for failure to adequately rig a 300 pound motor in the auxiliary building in accordance with the manufacturers recommendations on May 11, 2010. As a result, the motor slipped from its rigging and dropped approximately 15 feet onto the A component cooling water (CCW) pump motor below, damaging the motors cabling and electrical junction box. The CCW pump was declared inoperable (CR 380834), the damage was repaired, and the CCW pump restored to an operable status on May 15, 2010.

Inspectors determined that the failure to implement adequate rigging practices in accordance with vendor recommendations as required by procedure MA-AA-101, Revision 5, Fleet Lifting and Material Handling constituted a performance deficiency and a finding which was reasonably within the licensees ability to foresee and correct and which should have been prevented. The finding is similar to MC 0612, Appendix E example 4.f, and is more than minor because it resulted in damage to and inoperability of a risk significant component. The finding is associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events which upset plant stability and challenge critical safety functions during shutdown as well as power operations because a loss of the component cooling water system would have resulted in a unit transient. The finding, evaluated per Attachment 4 of MC-0609, Phase 1 - Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a plant transient and the loss of accident mitigation equipment. This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not make safety/risk significant decisions using a systematic process, especially when faced with uncertain decisions, to ensure safety is maintained (H.1(a)). Specifically, the rigging team made safety/risk significant decisions within lifting/rigging procedures that did not include a systematic process for evaluating each lift, especially loads <5000 lbs in the vicinity of risk significant equipment.

Inspection Report# : 2010003 (pdf)

Mitigating Systems Significance: Mar 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct Multiple Conditions Adverse to Fire Protection A Green, self-revealing non-cited violation of Condition 3.I to the Surry Unit 1 and Unit 2 Updated Facility Operating Licenses, DPR-32 and DPR-37, was identified for the licensees failure to take corrective action for degraded conditions adverse to the fire protection program. The licensee entered this issue into their corrective action program as condition report 398628.

The inspectors found that the failure to take action to correct multiple oversized breakers constituted a performance deficiency. The finding is more than minor because it adversely affected the external factors attribute (fire) of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the Unit 2 1B RWST chiller motor and the Unit 2B hydrogen recombiner breakers were the most susceptible to fire due to their size; also a cable fault could potentially damage safety related cables routed nearby. In addition, the Unit 1 2B charging component cooling water pump is safety related and its cable was also unprotected. The inspectors reviewed IMC 0609, Appendix F, , and determined the category of post fire safe shutdown was affected and the finding required a phase 3 analysis. A phase 3 risk analysis was performed by a regional SRA in accordance with IMC 0609 Appendix F,

NUREG/CR6850, NUREG/CR 6850 supplement 1, and utilizing the latest NRC Surry SPAR probabilistic risk analysis model and determined that the risk increase in core damage frequency was <1E-6, a finding of very low risk significance, Green. The cause of this finding involved the cross-cutting area of human performance, the component of work control, and the aspect of work planning, H.3(a), because the licensee failed to appropriately prioritize, schedule, and complete work activities consistent with risk insights and the safety significance of the equipment.

Inspection Report# : 2011002 (pdf)

Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate that the Reliability of Systems or Components were effectively controlled per 10 CFR 50.65 (a)(2)

The NRC identified a Green Non-Cited Violation of 10CFR50.65 a(2) for the licensees failure to demonstrate that the reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by was being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function. Specifically, the licensees MR program would not demonstrate that a system should remain in category a(2) as defined by regulatory requirements.

The inspectors determined the licensees MR program could not demonstrate that reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by were being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function is a performance deficiency. Specifically, the monitoring established by the license did not effectively demonstrate that systems in a(2) were being effectively controlled through performance of appropriate preventative maintenance. This masking of poor equipment performance does not allow the licensee to determine if a system should be in increased monitoring of a(1).

The finding was more than minor because it adversely affected the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of HSS and LSS systems to perform their functions when required. Specifically, multiple HSS and LSS systems could have a high probability of failure, because poor equipment performance would not be recognized by the licensee. This could prevent a poor performing system from being placed into the a(1) category when required and appropriate corrective action would not be taken.

The finding was evaluated using MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green), because the finding did not involve an actual failure of equipment. This finding had a crosscutting aspect in the area of human performance and resources because the licensee did not ensure that personnel, procedures, and other resources were available and adequate to assure proper implementation of MR program. The MR personnel did not have the training required to implement the program within the required industry regulations and guidelines (H.2.b).

Inspection Report# : 2010003 (pdf)

Barrier Integrity Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Heavy Load Lift of the 1B RCP Motor Over Exposed Reactor Fuel An NRC-identified Green NCV of Technical Specification (TS) 6.4, Unit Operating Procedures and Programs, was identified. Personnel failed to follow the defined heavy load shipping path inside containment as specified in procedure, GMP-001, Heavy Load Rigging and Movement, which resulted in the movement of the 1B reactor coolant pump motor over exposed reactor fuel. The licensee has entered the issue into the CAP (CR 404106).

Transport of the 1B reactor coolant pump motor over the exposed reactor core is a performance deficiency. The finding is more than minor because it can reasonably be viewed as a precursor to a significant event, the heavy load traveled over exposed irradiated fuel with the reactor vessel head removed. In accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, the inspectors conducted a Phase 1 SDP screening and determined the finding required a Phase 2 analysis. The Phase 2 analysis determined the finding is of very low safety significance (Green) because: (1) there is a low probability of dropping the load based on a study in NUREG-1774 performed for crane operating experience; (2) the polar crane was in working condition and had no known deficiencies that would have affected the cranes ability to lift the load; and, (3) the duration of the heavy load lift over the exposed reactor core was very short. In addition, in accordance with NRC IMC 0609, Appendix H, Containment Integrity SDP, the finding would not contribute to LERF due to the time since the reactor was shutdown. The finding has a cross-cutting aspect in the work practices component of the Human Performance area because plant supervisors failed to properly supervise workers executing procedure steps (H.4(c)).

Inspection Report# : 2010005 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Last modified : June 07, 2011

Surry 1 2Q/2011 Plant Inspection Findings Initiating Events Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadvertent Actuation of Safety Injection A self-revealing Green NCV of TS 6.4, Unit Operating Procedures and Programs, was identified for the failure to follow procedure 1-OPT-ZZ-001, ESF Actuation with Undervoltage and Degraded Voltage 1H Bus. Specifically, on October 26, 2010, a test lead was incorrectly installed in the Unit 1 relay room for the logic circuit associated with the A train of Consequence Limiting Safeguards (CLS). This resulted in an inadvertent safety injection, isolated component cooling water supply to the standby residual heat removal (RHR) train, and automatically initiated several safety-related components including emergency diesel generator (EDG) #1. Operators entered AP-10.20, Response To Spurious Safety Injection With RCS Temperature Less Than 350°F, and terminated the safety injection in approximately three minutes. The licensee entered this issue into the CAP (CR 400908).

Failure to install the test leads as required by procedure 1-OPT-ZZ-001, is a performance deficiency. The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstones objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding, evaluated in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 3, identified the finding is of very low safety significance (Green) because the finding did not lead to a loss of decay heat removal. This finding has a cross cutting aspect in the work practices component in the Human Performance area, because human error prevention techniques were not properly used commensurate with the risk significance of the assigned task (H.4(a)).

Inspection Report# : 2010005 (pdf)

Significance: Aug 16, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Degraded Unit 1 Nuclear Instrument RC Filters An NRC identified non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, was identified for the licensees failure to identify and correct degraded RC filters associated with Unit 1 Nuclear Instrument (NI) cabinets for N-42 and N-44 based on a similar degraded condition identified on Unit 2 NI cabinet N-43 in November 2009. The issue was entered into the licensees corrective action program as condition report CR383881.

All the RC filters in the Surry Unit 1 and 2 NI cabinets have been replaced with new RC filters.

The finding was determined to be of more than minor significance because it is associated with the equipment performance attribute of the Initiating Events cornerstone. It adversely affected the cornerstone objective of protection against external events, i.e., fire. The performance deficiency was screened using phase 1 of the Significance Determination Process (SDP) and was determined to be a fire initiator contributor and to have impact on post fire safe shutdown, therefore a phase 2 analysis utilizing Inspection Manual chapter 0609 Appendix F was required. Since the finding involved MCR fire scenarios, a phase 3 analysis was required. A phase 3 risk analysis was performed by a regional SRA in accordance with IMC 0609 Appendix F, NUREG/CR6850, and utilizing the latest Surry SPAR probabilistic risk analysis model. The fire scenarios were determined to impact MCR operator actions but would not credibly require MCR evacuation for either habitability or safe shutdown functional requirements. The dominant sequence was a fire induced reactor trip transient initiator, with failures of auxiliary feedwater, main feedwater and failure to implement feed and bleed leading to core damage. Factors which mitigated the risk of the fire were the minimal fire growth potential and the potential for NI cabinet fires to damage SSD equipment. The risk evaluation result was an increase of <1E-6 for core damage frequency, a finding of very low risk significance (Green). This finding involved the cross cutting area of problem identification and resolution, the component of operating experience (OE), and the aspect of evaluating

internal OE (P.2.a), because the licensee did not effectively evaluate the internal operating experience gained from the November 2009 RC filter failure prior to the failure of the RC filters on June 8, 2010.

Inspection Report# : 2010006 (pdf)

Mitigating Systems Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Qualification Testing of Fire Barrier Penetration Seals A Green non-cited violation of Surry Units 1 and 2 Operating License Condition 3.I, Fire Protection, was identified by the inspectors for failure to have adequate qualification testing results, as directed by Appendix A to Branch Technical Position APCSB 9.5-1. Specifically, the licensee did not have sufficient testing results to qualify certain aluminum conduit configurations that penetrate 3-hour fire rated barriers separating fire areas containing redundant equipment required for safe shutdown. As part of the corrective actions, the licensee performed testing to determine the qualification of aluminum conduit penetrations, and performed modifications, as appropriate, to restore compliance.

The finding is more than minor because it is associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e., fire) and it affects the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events. Specifically, not having qualification testing results for aluminum conduits that penetrate fire rated barriers adversely affected the fire confinement capability defense-in-depth element because subsequent testing revealed some conduit configurations that did not meet the penetration seal criteria established in Branch Technical Position APCSB 9.5-1. The inspectors used the guidance of NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and determined that the performance deficiency represented a finding of very low safety significance (Green).

Specifically, the fire areas in question either contained a non degraded automatic gaseous or water-based fire suppression system, or the exposed fire areas did not contain potential damage targets that are unique from those in the exposing fire areas. Inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. (Section 4OA5.3)

Inspection Report# : 2011003 (pdf)

Significance: Mar 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct Multiple Conditions Adverse to Fire Protection A Green, self-revealing non-cited violation of Condition 3.I to the Surry Unit 1 and Unit 2 Updated Facility Operating Licenses, DPR-32 and DPR-37, was identified for the licensees failure to take corrective action for degraded conditions adverse to the fire protection program. The licensee entered this issue into their corrective action program as condition report 398628.

The inspectors found that the failure to take action to correct multiple oversized breakers constituted a performance deficiency. The finding is more than minor because it adversely affected the external factors attribute (fire) of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the Unit 2 1B RWST chiller motor and the Unit 2B hydrogen recombiner breakers were the most susceptible to fire due to their size; also a cable fault could potentially damage safety related cables routed nearby. In addition, the Unit 1 2B charging component cooling water pump is safety related and its cable was also unprotected. The inspectors reviewed IMC 0609, Appendix F, , and determined the category of post fire safe shutdown was affected and the finding required a phase 3 analysis. A phase 3 risk analysis was performed by a regional SRA in accordance with IMC 0609 Appendix F, NUREG/CR6850, NUREG/CR 6850 supplement 1, and utilizing the latest NRC Surry SPAR probabilistic risk

analysis model and determined that the risk increase in core damage frequency was <1E-6, a finding of very low risk significance, Green. The cause of this finding involved the cross-cutting area of human performance, the component of work control, and the aspect of work planning, H.3(a), because the licensee failed to appropriately prioritize, schedule, and complete work activities consistent with risk insights and the safety significance of the equipment.

Inspection Report# : 2011002 (pdf)

Significance: SL-IV Feb 28, 2011 Identified By: NRC Item Type: VIO Violation Inaccurate Fire Watch Records The licensee identified a violation of 10 CFR 50.48 Fire Protection requirements when it was determined that a laborer failed to conduct a roving fire watch patrol. The licensee took substantial disciplinary actions and entered the deficiency into the corrective action program for resolution as CR 379888.

This issue was dispositioned using traditional enforcement due to the deliberate aspects of the performance deficiency. Furthermore, the failure to provide complete and accurate information has the potential to impact the NRCs ability to perform its regulatory function.

An individual assigned as a fire watch deliberately documented the completion of fire watch rounds (Fire Watch Tour Documentation Sheet, Attachment 14) for locations in which he did not conduct the fire watches. This issue was considered more than minor due to the deliberate aspects of the performance deficiency. In accordance with the guidance in Supplement VII of the Enforcement Policy, this issue is considered a Severity Level IV violation because it involved information that the NRC required to be maintained by a licensee that was incomplete or inaccurate and of more than minor safety significance. No cross-cutting aspect was identified because this performance deficiency was dispositioned using traditional enforcement.

Inspection Report# : 2011012 (pdf)

Barrier Integrity Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Heavy Load Lift of the 1B RCP Motor Over Exposed Reactor Fuel An NRC-identified Green NCV of Technical Specification (TS) 6.4, Unit Operating Procedures and Programs, was identified. Personnel failed to follow the defined heavy load shipping path inside containment as specified in procedure, GMP-001, Heavy Load Rigging and Movement, which resulted in the movement of the 1B reactor coolant pump motor over exposed reactor fuel. The licensee has entered the issue into the CAP (CR 404106).

Transport of the 1B reactor coolant pump motor over the exposed reactor core is a performance deficiency. The finding is more than minor because it can reasonably be viewed as a precursor to a significant event, the heavy load traveled over exposed irradiated fuel with the reactor vessel head removed. In accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, the inspectors conducted a Phase 1 SDP screening and determined the finding required a Phase 2 analysis. The Phase 2 analysis determined the finding is of very low safety significance (Green) because: (1) there is a low probability of dropping the load based on a study in NUREG-1774 performed for crane operating experience; (2) the polar crane was in working condition and had no known deficiencies that would have affected the cranes ability to lift the load; and, (3) the duration of the heavy load lift over the exposed reactor core was very short. In addition, in accordance with NRC IMC 0609, Appendix H, Containment Integrity SDP, the finding would not contribute to LERF due to the time since the reactor was shutdown. The finding has a cross-cutting aspect in the work practices component of the Human Performance area because plant supervisors failed to properly supervise workers executing procedure steps (H.4(c)).

Inspection Report# : 2010005 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Significance: N/A Jun 24, 2011 Identified By: NRC Item Type: FIN Finding PI&R inspection results The inspection team concluded that, in general, problems were adequately identified, prioritized, and evaluated; and effective corrective actions were implemented. Site management was actively involved in the corrective action program (CAP) and focused appropriate attention on significant plant issues. The team found that employees were encouraged by management to initiate condition reports (CRs) as appropriate to address plant issues.

The licensee was effective at identifying problems and entering them into the CAP for resolution, as evidenced by the relatively few deficiencies identified by the NRC that had not been previously identified by the licensee during the review period. The threshold for initiating CRs was appropriately low, as evidenced by the type of problems identified and large number of CRs entered annually into the CAP. In addition, CRs normally provided complete and accurate characterization of the problem.

Generally, prioritization and evaluation of issues were adequate and consistent with the licensees CAP guidance.

Formal root cause evaluations for significant problems were adequate, and corrective actions specified for problems did address the cause of the problems. The age and extensions for completing evaluations were closely monitored by plant management, both for high priority condition reports, as well as for adverse conditions of less significant priority. Also, the technical adequacy and depth of evaluations (e.g., root cause investigations) were typically adequate. However, the team identified two minor issues associated with the licensees identification of issues and effectiveness of corrective actions.

Corrective actions were generally effective, timely, and commensurate with the safety significance of the issues.

The operating experience program was effective in screening operating experience for applicability to the plant, entering items determined to be applicable into the CAP, and taking adequate corrective actions to address the issues.

External and internal operating experience was adequately utilized and considered as part of formal root cause

evaluations for supporting the development of lessons learned and corrective actions for CAP issues.

The licensees audits and self-assessments were critical and effective in identifying issues and entering them into the corrective action program. These audits and assessments identified issues similar to those identified by the NRC with respect to the effectiveness of the CAP.

Based on general discussions with licensee employees during the inspection, targeted interviews with plant personnel, and reviews of selected employee concerns records, the inspectors determined that personnel at the site felt free to raise safety concerns to management and use the CAP as well as the employee concerns program to resolve those concerns.

Inspection Report# : 2011008 (pdf)

Last modified : October 14, 2011

Surry 1 3Q/2011 Plant Inspection Findings Initiating Events Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Consider Instrument Uncertainty and Establish Calibration Controls for Rotameters Used to Vent Gas from ECCS Systems An NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XI, Test Control, (with two examples) was identified for the failure to establish measures to apply rotameter instrument measurement error and appropriate instrument calibration controls or standards when using instruments of this type to determine the size of voids discovered as a result of ECCS system venting. The issue was entered into the licensees corrective action program (CAP) as CR419024 and CR419243.

The failure to establish and implement measures (1) to ensure the application of +/- 5% rotameter instrument error to as-found void measurement, and (2) to ensure that rotameters calibrated to standard pressure conditions were used when utilizing those instruments to evaluate the size of as-found voids were performance deficiencies. The performance deficiencies were greater than minor, because, if left uncorrected, they could result in a more significant safety concern. Specifically, the performance deficiencies represented programmatic issues and if instrument error and/or appropriate calibration standards were not applied to instruments used for future void characterization, then sufficient measurement error could reasonably result such that as-found voids, which challenge or exceed established acceptance criteria, may not be identified as intended by post venting evaluations. The finding was screened for significance using the Mitigating Systems cornerstone column of Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because the finding did not represent a design or qualification deficiency, did not represent the loss of a safety system function, did not represent the loss of a train for greater than the allowed outage time, did not represent the loss of risk significant equipment for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and was not potentially risk significant due to external events.

Because the licensee had failed to implement complete, accurate, and up-to-date controls necessary to ensure that rotameter error and calibration standards were adequately addressed by procedures used to evaluate the impact of voids on emergency core cooling systems, this finding is assigned a cross-cutting aspect in resources component of the human performance area H.2(c). (Section 4OA5.1)

Inspection Report# : 2011004 (pdf)

Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadvertent Actuation of Safety Injection A self-revealing Green NCV of TS 6.4, Unit Operating Procedures and Programs, was identified for the failure to follow procedure 1-OPT-ZZ-001, ESF Actuation with Undervoltage and Degraded Voltage 1H Bus. Specifically, on October 26, 2010, a test lead was incorrectly installed in the Unit 1 relay room for the logic circuit associated with the A train of Consequence Limiting Safeguards (CLS). This resulted in an inadvertent safety injection, isolated component cooling water supply to the standby residual heat removal (RHR) train, and automatically initiated several safety-related components including emergency diesel generator (EDG) #1. Operators entered AP-10.20, Response To Spurious Safety Injection With RCS Temperature Less Than 350°F, and terminated the safety injection in approximately three minutes. The licensee entered this issue into the CAP (CR 400908).

Failure to install the test leads as required by procedure 1-OPT-ZZ-001, is a performance deficiency. The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstones objective to limit the likelihood of those events that upset plant stability and

challenge critical safety functions during shutdown as well as power operations. The finding, evaluated in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 3, identified the finding is of very low safety significance (Green) because the finding did not lead to a loss of decay heat removal. This finding has a cross cutting aspect in the work practices component in the Human Performance area, because human error prevention techniques were not properly used commensurate with the risk significance of the assigned task (H.4(a)).

Inspection Report# : 2010005 (pdf)

Mitigating Systems Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Scaffolding Procedure Requirements The inspectors identified a NCV of Technical Specifications (TS) 6.4.D for failing to follow the requirements of procedure MA-AA-105, Scaffolding. Specifically, the licensee did not adequately implement scaffold evaluation, screening, and risk requirements for multiple scaffolds constructed in the vicinity of safety-related equipment.

The inspectors determined that the failure to follow TS required procedure MA-AA-105, Scaffolding, by not properly identifying scaffolds for safety-related systems and performing the required engineering evaluations, constitutes a performance deficiency. This finding is considered more than minor because it is similar to IMC 0612, Appendix E, Example 4.a in that the licensee routinely failed to perform the required engineering reviews and evaluations for scaffolding. This finding is also associated with the external factors and equipment performance attributes of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors screened this finding in accordance with IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance since it was a deficiency determined not to have resulted in the loss of operability or functionality. The cause of this finding involved the cross-cutting area of human performance, the component of resources and the aspect of training H.2(b), because the licensee failed to implement training sufficient to ensure that operators were aware of plant equipment which is designated as safety-related. (Section 1R04)

Inspection Report# : 2011004 (pdf)

Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Qualification Testing of Fire Barrier Penetration Seals A Green non-cited violation of Surry Units 1 and 2 Operating License Condition 3.I, Fire Protection, was identified by the inspectors for failure to have adequate qualification testing results, as directed by Appendix A to Branch Technical Position APCSB 9.5-1. Specifically, the licensee did not have sufficient testing results to qualify certain aluminum conduit configurations that penetrate 3-hour fire rated barriers separating fire areas containing redundant equipment required for safe shutdown. As part of the corrective actions, the licensee performed testing to determine the qualification of aluminum conduit penetrations, and performed modifications, as appropriate, to restore compliance.

The finding is more than minor because it is associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e., fire) and it affects the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events. Specifically, not having qualification testing results for aluminum conduits that penetrate fire rated barriers adversely affected the fire confinement capability defense-in-depth element because subsequent testing revealed some conduit configurations that did not meet the penetration seal criteria established in Branch Technical Position APCSB 9.5-1. The inspectors used the guidance of NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and

determined that the performance deficiency represented a finding of very low safety significance (Green).

Specifically, the fire areas in question either contained a non degraded automatic gaseous or water-based fire suppression system, or the exposed fire areas did not contain potential damage targets that are unique from those in the exposing fire areas. Inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. (Section 4OA5.3)

Inspection Report# : 2011003 (pdf)

Significance: Mar 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct Multiple Conditions Adverse to Fire Protection A Green, self-revealing non-cited violation of Condition 3.I to the Surry Unit 1 and Unit 2 Updated Facility Operating Licenses, DPR-32 and DPR-37, was identified for the licensees failure to take corrective action for degraded conditions adverse to the fire protection program. The licensee entered this issue into their corrective action program as condition report 398628.

The inspectors found that the failure to take action to correct multiple oversized breakers constituted a performance deficiency. The finding is more than minor because it adversely affected the external factors attribute (fire) of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the Unit 2 1B RWST chiller motor and the Unit 2B hydrogen recombiner breakers were the most susceptible to fire due to their size; also a cable fault could potentially damage safety related cables routed nearby. In addition, the Unit 1 2B charging component cooling water pump is safety related and its cable was also unprotected. The inspectors reviewed IMC 0609, Appendix F, , and determined the category of post fire safe shutdown was affected and the finding required a phase 3 analysis. A phase 3 risk analysis was performed by a regional SRA in accordance with IMC 0609 Appendix F, NUREG/CR6850, NUREG/CR 6850 supplement 1, and utilizing the latest NRC Surry SPAR probabilistic risk analysis model and determined that the risk increase in core damage frequency was <1E-6, a finding of very low risk significance, Green. The cause of this finding involved the cross-cutting area of human performance, the component of work control, and the aspect of work planning, H.3(a), because the licensee failed to appropriately prioritize, schedule, and complete work activities consistent with risk insights and the safety significance of the equipment.

Inspection Report# : 2011002 (pdf)

Significance: SL-IV Feb 28, 2011 Identified By: NRC Item Type: VIO Violation Inaccurate Fire Watch Records The licensee identified a violation of 10 CFR 50.48 Fire Protection requirements when it was determined that a laborer failed to conduct a roving fire watch patrol. The licensee took substantial disciplinary actions and entered the deficiency into the corrective action program for resolution as CR 379888.

This issue was dispositioned using traditional enforcement due to the deliberate aspects of the performance deficiency. Furthermore, the failure to provide complete and accurate information has the potential to impact the NRCs ability to perform its regulatory function.

An individual assigned as a fire watch deliberately documented the completion of fire watch rounds (Fire Watch Tour Documentation Sheet, Attachment 14) for locations in which he did not conduct the fire watches. This issue was considered more than minor due to the deliberate aspects of the performance deficiency. In accordance with the guidance in Supplement VII of the Enforcement Policy, this issue is considered a Severity Level IV violation because it involved information that the NRC required to be maintained by a licensee that was incomplete or inaccurate and of more than minor safety significance. No cross-cutting aspect was identified because this performance deficiency was dispositioned using traditional enforcement.

Inspection Report# : 2011004 (pdf)

Inspection Report# : 2011012 (pdf)

Barrier Integrity Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Heavy Load Lift of the 1B RCP Motor Over Exposed Reactor Fuel An NRC-identified Green NCV of Technical Specification (TS) 6.4, Unit Operating Procedures and Programs, was identified. Personnel failed to follow the defined heavy load shipping path inside containment as specified in procedure, GMP-001, Heavy Load Rigging and Movement, which resulted in the movement of the 1B reactor coolant pump motor over exposed reactor fuel. The licensee has entered the issue into the CAP (CR 404106).

Transport of the 1B reactor coolant pump motor over the exposed reactor core is a performance deficiency. The finding is more than minor because it can reasonably be viewed as a precursor to a significant event, the heavy load traveled over exposed irradiated fuel with the reactor vessel head removed. In accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, the inspectors conducted a Phase 1 SDP screening and determined the finding required a Phase 2 analysis. The Phase 2 analysis determined the finding is of very low safety significance (Green) because: (1) there is a low probability of dropping the load based on a study in NUREG-1774 performed for crane operating experience; (2) the polar crane was in working condition and had no known deficiencies that would have affected the cranes ability to lift the load; and, (3) the duration of the heavy load lift over the exposed reactor core was very short. In addition, in accordance with NRC IMC 0609, Appendix H, Containment Integrity SDP, the finding would not contribute to LERF due to the time since the reactor was shutdown. The finding has a cross-cutting aspect in the work practices component of the Human Performance area because plant supervisors failed to properly supervise workers executing procedure steps (H.4(c)).

Inspection Report# : 2010005 (pdf)

Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.

Miscellaneous Significance: N/A Jun 24, 2011 Identified By: NRC Item Type: FIN Finding PI&R inspection results The inspection team concluded that, in general, problems were adequately identified, prioritized, and evaluated; and effective corrective actions were implemented. Site management was actively involved in the corrective action program (CAP) and focused appropriate attention on significant plant issues. The team found that employees were encouraged by management to initiate condition reports (CRs) as appropriate to address plant issues.

The licensee was effective at identifying problems and entering them into the CAP for resolution, as evidenced by the relatively few deficiencies identified by the NRC that had not been previously identified by the licensee during the review period. The threshold for initiating CRs was appropriately low, as evidenced by the type of problems identified and large number of CRs entered annually into the CAP. In addition, CRs normally provided complete and accurate characterization of the problem.

Generally, prioritization and evaluation of issues were adequate and consistent with the licensees CAP guidance.

Formal root cause evaluations for significant problems were adequate, and corrective actions specified for problems did address the cause of the problems. The age and extensions for completing evaluations were closely monitored by plant management, both for high priority condition reports, as well as for adverse conditions of less significant priority. Also, the technical adequacy and depth of evaluations (e.g., root cause investigations) were typically adequate. However, the team identified two minor issues associated with the licensees identification of issues and effectiveness of corrective actions.

Corrective actions were generally effective, timely, and commensurate with the safety significance of the issues.

The operating experience program was effective in screening operating experience for applicability to the plant, entering items determined to be applicable into the CAP, and taking adequate corrective actions to address the issues.

External and internal operating experience was adequately utilized and considered as part of formal root cause evaluations for supporting the development of lessons learned and corrective actions for CAP issues.

The licensees audits and self-assessments were critical and effective in identifying issues and entering them into the corrective action program. These audits and assessments identified issues similar to those identified by the NRC with respect to the effectiveness of the CAP.

Based on general discussions with licensee employees during the inspection, targeted interviews with plant personnel, and reviews of selected employee concerns records, the inspectors determined that personnel at the site felt free to raise safety concerns to management and use the CAP as well as the employee concerns program to resolve those concerns.

Inspection Report# : 2011008 (pdf)

Last modified : January 04, 2012

Surry 1 4Q/2011 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions to Address Rainwater Intrusion into the Unit 1 RPS Relay Room The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failing to correct a condition adverse to quality. Specifically, the licensee failed to correct a known degraded building seam which allowed rainwater into the Unit 1 Reactor Protection System (RPS) Relay Room on four separate occasions over a four year time period.

The inspectors concluded that the failure to correct the known degraded building seam was a performance deficiency that was within the licensees ability to foresee and correct and which should have been prevented. The finding was more than minor because it is associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors - Weather (heavy rain), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the potential existed that under design basis rainfall conditions the water would migrate to a point where RPS equipment would be impacted. The inspectors determined that this finding was of very low safety significance because the finding was not a design issue, did not result in a loss of a safety function of a mitigating system, and did not screen potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding involved the cross-cutting area of problem identification and resolution, the component of corrective action program, and the aspect of appropriate corrective actions P.1(d), because the licensee failed to take appropriate and timely corrective actions commensurate with the safety significance of the rainwater intrusion events. (Section 4OA2)

Inspection Report# : 2011005 (pdf)

Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Scaffolding Procedure Requirements The inspectors identified a NCV of Technical Specifications (TS) 6.4.D for failing to follow the requirements of procedure MA-AA-105, Scaffolding. Specifically, the licensee did not adequately implement scaffold evaluation, screening, and risk requirements for multiple scaffolds constructed in the vicinity of safety-related equipment.

The inspectors determined that the failure to follow TS required procedure MA-AA-105, Scaffolding, by not properly identifying scaffolds for safety-related systems and performing the required engineering evaluations, constitutes a performance deficiency. This finding is considered more than minor because it is similar to IMC 0612, Appendix E, Example 4.a in that the licensee routinely failed to perform the required engineering reviews and evaluations for scaffolding. This finding is also associated with the external factors and equipment performance attributes of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors screened this finding in accordance with IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance since it was a deficiency determined not to have resulted in the loss of operability or functionality. The

cause of this finding involved the cross-cutting area of human performance, the component of resources and the aspect of training H.2(b), because the licensee failed to implement training sufficient to ensure that operators were aware of plant equipment which is designated as safety-related. (Section 1R04)

Inspection Report# : 2011004 (pdf)

Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Consider Instrument Uncertainty and Establish Calibration Controls for Rotameters Used to Vent Gas from ECCS Systems An NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XI, Test Control, (with two examples) was identified for the failure to establish measures to apply rotameter instrument measurement error and appropriate instrument calibration controls or standards when using instruments of this type to determine the size of voids discovered as a result of ECCS system venting. The issue was entered into the licensees corrective action program (CAP) as CR419024 and CR419243.

The failure to establish and implement measures (1) to ensure the application of +/- 5% rotameter instrument error to as-found void measurement, and (2) to ensure that rotameters calibrated to standard pressure conditions were used when utilizing those instruments to evaluate the size of as-found voids were performance deficiencies. The performance deficiencies were greater than minor, because, if left uncorrected, they could result in a more significant safety concern. Specifically, the performance deficiencies represented programmatic issues and if instrument error and/or appropriate calibration standards were not applied to instruments used for future void characterization, then sufficient measurement error could reasonably result such that as-found voids, which challenge or exceed established acceptance criteria, may not be identified as intended by post venting evaluations. The finding was screened for significance using the Mitigating Systems cornerstone column of Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because the finding did not represent a design or qualification deficiency, did not represent the loss of a safety system function, did not represent the loss of a train for greater than the allowed outage time, did not represent the loss of risk significant equipment for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and was not potentially risk significant due to external events.

Because the licensee had failed to implement complete, accurate, and up-to-date controls necessary to ensure that rotameter error and calibration standards were adequately addressed by procedures used to evaluate the impact of voids on emergency core cooling systems, this finding is assigned a cross-cutting aspect in resources component of the human performance area H.2(c). (Section 4OA5.1)

Inspection Report# : 2011004 (pdf)

Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Qualification Testing of Fire Barrier Penetration Seals A Green non-cited violation of Surry Units 1 and 2 Operating License Condition 3.I, Fire Protection, was identified by the inspectors for failure to have adequate qualification testing results, as directed by Appendix A to Branch Technical Position APCSB 9.5-1. Specifically, the licensee did not have sufficient testing results to qualify certain aluminum conduit configurations that penetrate 3-hour fire rated barriers separating fire areas containing redundant equipment required for safe shutdown. As part of the corrective actions, the licensee performed testing to determine the qualification of aluminum conduit penetrations, and performed modifications, as appropriate, to restore compliance.

The finding is more than minor because it is associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e., fire) and it affects the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events. Specifically, not having qualification testing results for aluminum conduits that penetrate fire rated barriers adversely affected the fire confinement capability defense-in-depth element because subsequent testing revealed some conduit configurations that did not meet the

penetration seal criteria established in Branch Technical Position APCSB 9.5-1. The inspectors used the guidance of NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and determined that the performance deficiency represented a finding of very low safety significance (Green).

Specifically, the fire areas in question either contained a non degraded automatic gaseous or water-based fire suppression system, or the exposed fire areas did not contain potential damage targets that are unique from those in the exposing fire areas. Inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. (Section 4OA5.3)

Inspection Report# : 2011003 (pdf)

Significance: Mar 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct Multiple Conditions Adverse to Fire Protection A Green, self-revealing non-cited violation of Condition 3.I to the Surry Unit 1 and Unit 2 Updated Facility Operating Licenses, DPR-32 and DPR-37, was identified for the licensees failure to take corrective action for degraded conditions adverse to the fire protection program. The licensee entered this issue into their corrective action program as condition report 398628.

The inspectors found that the failure to take action to correct multiple oversized breakers constituted a performance deficiency. The finding is more than minor because it adversely affected the external factors attribute (fire) of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the Unit 2 1B RWST chiller motor and the Unit 2B hydrogen recombiner breakers were the most susceptible to fire due to their size; also a cable fault could potentially damage safety related cables routed nearby. In addition, the Unit 1 2B charging component cooling water pump is safety related and its cable was also unprotected. The inspectors reviewed IMC 0609, Appendix F, , and determined the category of post fire safe shutdown was affected and the finding required a phase 3 analysis. A phase 3 risk analysis was performed by a regional SRA in accordance with IMC 0609 Appendix F, NUREG/CR6850, NUREG/CR 6850 supplement 1, and utilizing the latest NRC Surry SPAR probabilistic risk analysis model and determined that the risk increase in core damage frequency was <1E-6, a finding of very low risk significance, Green. The cause of this finding involved the cross-cutting area of human performance, the component of work control, and the aspect of work planning, H.3(a), because the licensee failed to appropriately prioritize, schedule, and complete work activities consistent with risk insights and the safety significance of the equipment.

Inspection Report# : 2011002 (pdf)

Significance: SL-IV Feb 28, 2011 Identified By: NRC Item Type: VIO Violation Inaccurate Fire Watch Records The licensee identified a violation of 10 CFR 50.48 Fire Protection requirements when it was determined that a laborer failed to conduct a roving fire watch patrol. The licensee took substantial disciplinary actions and entered the deficiency into the corrective action program for resolution as CR 379888.

This issue was dispositioned using traditional enforcement due to the deliberate aspects of the performance deficiency. Furthermore, the failure to provide complete and accurate information has the potential to impact the NRCs ability to perform its regulatory function.

An individual assigned as a fire watch deliberately documented the completion of fire watch rounds (Fire Watch Tour Documentation Sheet, Attachment 14) for locations in which he did not conduct the fire watches. This issue was considered more than minor due to the deliberate aspects of the performance deficiency. In accordance with the guidance in Supplement VII of the Enforcement Policy, this issue is considered a Severity Level IV violation because it involved information that the NRC required to be maintained by a licensee that was incomplete or inaccurate and of more than minor safety significance. No cross-cutting aspect was identified because this performance deficiency was dispositioned

using traditional enforcement.