IR 05000237/2008002
| ML081330658 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 05/12/2008 |
| From: | NRC/RGN-III/DRP/B1 |
| To: | Pardee C Exelon Generation Co |
| References | |
| IR-08-002 | |
| Download: ML081330658 (41) | |
Text
May 12, 2008
SUBJECT:
DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 INTEGRATED INSPECTION REPORT 05000237/2008002; 05000249/2008002
Dear Mr. Pardee:
On March 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Dresden Nuclear Power Station, Units 2 and 3. The enclosed report documents the inspection findings, which were discussed on April 9, 2008, with Mr. T. Hanley and other members of your staff. In addition, three unresolved items described within this report were discussed in a telephone call on May 2, 2008, with Mr. J. Ellis of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of this inspection, no findings of significance were identified.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Mark A. Ring, Chief Branch 1 Division of Reactor Projects Docket Nos. 50-237; 50-249;72-037 License Nos. DPR-19; DPR-25 Enclosure: Inspection Report 05000237/2008002; 05000249/2008002 w/Attachment: Supplemental Information DISTRIBUTION:
See next page
Letter to
SUMMARY OF FINDINGS
IR 05000237/2008002, 05000249/2008002; 01/01/2008 - 03/31/2008, Dresden Nuclear Power
Station, Units 2 and 3.
This report covers a three-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. The U.S. Nuclear Regulatory Commissions program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
NRC-Identified and Self-Revealing Findings
No violations of significance were identified.
Licensee-Identified Violations
No violations of significance were identified.
REPORT DETAILS
Summary of Plant Status
Unit 2 On February 3, 2008, power was reduced to approximately 91 percent electrical output to perform control rod drive testing, and the unit returned to full power on the same day.
On February 25, 2008, power was reduced to approximately 96 percent electrical output to perform control rod drive testing, and the unit returned to full power on the same day.
On March 1, 2008, power was reduced to approximately 67 percent electrical output to perform turbine valve testing, a control rod pattern adjustment, reactor feedwater pump maintenance, and other activities. The unit returned to full power on the same day.
Unit 3 The unit began the month of January slightly derated due to degraded extraction steam pressure to the 3C2 and 3B2 feedwater heaters.
On January 8, 2008, the unit was taken offline to repair/replace several extraction steam bellows and perform low pressure turbine casing repairs. The unit returned to full power on January 21.
On March 8, 2008, power was reduced to approximately 72 percent electrical output to perform turbine valve testing, a control rod pattern adjustment, and various other activities. The unit returned to full power on the same day.
REACTOR SAFETY
Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
.1 Readiness For Impending Adverse Weather Condition - Extreme Cold Conditions
a. Inspection Scope
Since extreme cold conditions were forecast in the vicinity of the facility for the week of January 21, 2008, the inspectors reviewed the licensees overall preparations/protection for the expected weather conditions. On January 22, 2008, the inspectors walked down portions of the Unit 2 and 3 reactor buildings and Unit 2 and 3 turbine buildings because the equipment could be affected or required as a result of the extreme cold conditions forecast for the facility. The inspectors observed insulation, heat trace circuits, space heater operation, and weatherized enclosures to ensure operability of affected systems.
The inspectors reviewed licensee procedures and discussed potential compensatory measures with control room personnel. The inspectors focused on plant managements actions for implementing the stations procedures for ensuring adequate personnel for safe plant operation and emergency response would be available.
This inspection constitutes one readiness for impending adverse weather condition sample as defined in Inspection Procedure 71111.01-05.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment
.1 Quarterly Partial System Walkdowns
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant systems:
- Unit 3A core spray out-of-service, verified Unit 3B core spray alignment;
- Unit 2 125V DC safety related battery after replacement; and
- Unit 3 emergency diesel generator out-of-service, verified Unit 2 and Unit 2/3 emergency diesel generator alignment.
The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Final Safety Analysis Report (UFSAR), Technical Specification (TS) requirements, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program (CAP) with the appropriate significance characterization. Documents reviewed are listed in the Attachment.
These activities constituted three partial system walkdown samples as defined in Inspection Procedure 71111.04-05.
b. Findings
No findings of significance were identified.
.2 Semi-Annual Complete System Walkdown
a. Inspection Scope
During the week of March 3, 2008, the inspectors performed a complete system alignment inspection of the Unit 2 station blackout diesel to verify the functional capability of the system. This system was selected because it was considered risk-significant in the licensees probabilistic risk assessment. The inspectors walked down the system to review mechanical and electrical equipment lineups, electrical power availability, system pressure and temperature indications, as appropriate, component labeling, component lubrication, component and equipment cooling, hangers and supports, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. A review of a sample of past and outstanding work orders (WOs) was performed to determine whether any deficiencies significantly affected the system function. In addition, the inspectors reviewed the CAP database to ensure that system equipment alignment problems were being identified and appropriately resolved. The documents used for the walkdown and issue review are listed in the Attachment.
These activities constituted one complete system walkdown sample as defined in Inspection Procedure 71111.04-05.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:
- Fire Zone 9.0.C, 2/3 Swing emergency diesel generator room, elevation 517;
- Fire Zone 7.0.A.1, 7.0.A.2, 7.0.A.3, B.2.7, Unit 2 turbine building battery rooms, elevation 549;
- Fire Zone 1.1.2.6, Unit 2 reactor building refuel floor, elevation 613;
- Fire Zone 1.1.1.6, Unit 3 reactor building refuel floor, elevation 613; and
- Fire Zone 8.2.4, Units 2 and 3 cable tunnel, elevation 502.
The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and had implemented adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event. Using the documents listed in the Attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed, that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees CAP.
These activities constituted five quarterly fire protection inspection samples as defined in Inspection Procedure 71111.05-05.
b. Findings
No findings of significance were identified.
1R07 Annual Heat Sink Performance
a. Inspection Scope
The inspectors reviewed the licensees testing of Unit 3 emergency diesel generator lube oil heat exchangers to verify that potential deficiencies did not mask the licensees ability to detect degraded performance, to identify any common cause issues that had the potential to increase risk, and to ensure that the licensee was adequately addressing problems that could result in initiating events that would cause an increase in risk. The inspectors reviewed the licensees observations as compared against acceptance criteria, the correlation of scheduled testing and the frequency of testing, and the impact of instrument inaccuracies on test results. Inspectors also verified that test acceptance criteria considered differences between test conditions, design conditions, and testing conditions.
This inspection constitutes one sample as defined in Inspection Procedure 71111.07-05.
b. Findings
No findings of significance were identified.
1R07 Triennial Heat Sink Performance
.1 Triennial Review of Heat Sink Performance
a. Inspection Scope
The inspectors reviewed operability determinations, completed surveillances, vendor manual information, associated calculations, performance test results and cooler inspection results associated with the 3A Low Pressure Coolant Injection (LPCI) heat exchanger and the 3B LPCI room cooler. These heat exchangers/coolers were chosen based on their risk significance in the licensees probabilistic safety analysis, their important safety-related mitigating system support functions and their relatively low margin.
For the 3A LPCI heat exchanger and the 3B LPCI room cooler, the inspectors verified that testing, inspection, maintenance, and monitoring of biotic fouling and macrofouling programs were adequate to ensure proper heat transfer. This was accomplished by verifying the test method used was consistent with accepted industry practices, or equivalent, the test conditions were consistent with the selected methodology, the test acceptance criteria were consistent with the design basis values, and results of heat exchanger performance testing were properly evaluated. The inspectors also verified that the test results appropriately considered differences between testing conditions and design conditions, the frequency of testing based on trending of test results was sufficient to detect degradation prior to loss of heat removal capabilities below design basis values, and test results considered test instrument inaccuracies and differences.
For the 3A LPCI heat exchanger and the 3B LPCI room cooler, the inspectors reviewed the methods and results of heat exchanger performance inspections. The inspectors verified the methods used to inspect and clean heat exchangers were consistent with as-found conditions identified and expected degradation trends and industry standards, the licensees inspection and cleaning activities had established acceptance criteria consistent with industry standards, and the as-found results were recorded, evaluated, and appropriately dispositioned such that the as-left condition was acceptable.
In addition, the inspectors verified the condition and operation of the 3A LPCI heat exchanger and the 3B LPCI room cooler were consistent with design assumptions in heat transfer calculations and as described in the Final Safety Analysis Report (FSAR).
This included verification that the number of plugged tubes was within pre-established limits based on capacity and heat transfer assumptions. The inspectors verified the licensee evaluated the potential for water hammer and established adequate controls and operational limits to prevent heat exchanger degradation due to excessive flow induced vibration during operation. In addition, eddy current test reports and visual inspection records were reviewed to determine the structural integrity of the heat exchanger.
The inspectors verified the performance of the following attributes of the ultimate heat sink (UHS) and subcomponents such as piping, intake screens, pumps, valves, etc. by tests or other equivalent methods to ensure availability and accessibility to the cooling water systems in plant.
The inspectors reviewed the results of the licensees inspection of the UHS weirs or excavations. The inspectors verified that identified settlement or movement indicating loss of structural integrity and/or capacity was appropriately evaluated and dispositioned by the licensee. In addition, the inspectors verified the licensee ensured sufficient reservoir capacity by trending and removing debris or sediment buildup in the UHS.
The inspectors reviewed the licensees operation of the UHS. The inspectors also verified the licensees ability to ensure functionality of the UHS during adverse weather conditions. The inspectors verified that the licensees biocide treatments for biotic control were adequately conducted and the results monitored, trended, and evaluated.
In addition, the inspectors reviewed condition reports related to the heat exchangers or coolers and heat sink performance issues to verify that the licensee had an appropriate threshold for identifying issues and to evaluate the effectiveness of the corrective actions. The documents that were reviewed are included at the end of the report.
These inspection activities constituted two samples.
b. Findings
No findings of significance were identified. Three Unresolved Items were identified.
- (1) LPCI Heat Exchangers Design Calculation Deficiencies and Discrepancies
Introduction:
The inspectors identified an unresolved item (URI) regarding the ability of the Unit 2 and Unit 3 LPCI heat exchangers to perform their design safety functions after a design basis loss of coolant accident (Design Basis Accident (DBA)-LOCA).
Description:
During a routine thermal performance test on March 21, 2008, the licensee determined that the 3B LPCI heat exchanger had degraded. Specifically, the heat removal capability was determined to be 70.586 MBtu/hr at design conditions which was below the design heat removal rate of 71 MBtu/hr as shown in UFSAR Table 6.2-3b.
The licensee determined the heat exchanger was operable but degraded.
On April 14, 2008, while completing actions associated with a concern raised during the NRC Component Design Bases Inspection (05000237/2007006; 05000249/2007006),the licensee discovered that the increased containment pressure expected as a result of extended power uprate (EPU) post-LOCA conditions was not considered in the containment analysis. The increased containment pressure would result in an increased LPCI system pressure which required an increase in the assumed containment cooling service water (CCSW) system pressure to ensure that the CCSW system pressure would always be 20 psid more than the LPCI system. This differential pressure was required to prevent radioactive water from leaking from the LPCI system into the CCSW system and discharging to the Kankakee River. The licensee determined that the required throttling of the CCSW system flow to maintain a higher pressure would reduce the CCSW system flowrate below the 5,000 gpm assumed in DBA-LOCA design calculations. The licensee determined that the CCSW system flowrate could be as low as 3,798 gpm.
The inspectors were also concerned that the associated increase in containment temperature resulting from the increased pressure may not have been considered by the licensee in determining the maximum reactor building room temperatures assumed for the environmental qualification of equipment throughout the reactor building. This potential omission will be considered by the next URI.
The licensee initiated Assignment Report (AR) 763663 and determined the heat exchangers were operable at a maximum river temperature of 85 degrees Fahrenheit.
(Current temperature was about 75 degrees Fahrenheit.) This issue is considered unresolved (URI 05000237/2008002-01 (DRS) URI 05000249/2008002-01 (DRS))
pending the licensees completion of a past operability determination and inspectors review of the evaluation.
- (2) LPCI Room Cooler Calculation Deficiencies and Discrepancies
Introduction:
The inspectors identified an unresolved item regarding the calculation of the post-LOCA, post-EPU temperatures in the LPCI Room after a DBA-LOCA without LPCI room cooling.
Description:
The inspectors requested documentation with respect to LPCI room heat-up calculations. The inspectors noted that UFSAR (revision 7, dated June 2007)section 3.11.4, Loss of Ventilation, stated, The LPCI/CS (core spray) room coolers are not required to support the operation of the LPCI/CS systems. All safety-related electrical equipment located in the LPCI/CS rooms that is subject to 10 CFR 50.49 has been environmentally qualified for the elevated temperatures resulting from a LOCA concurrent with a loss of the room coolers (LORC). The inspectors were provided Calculation DRE 00-0075, Evaluation of Dresden binders for EPU environmental conditions, Revision 2, dated December 18, 2005, to support the determination that the room coolers were not needed and therefore, were not safety-related. The calculation concluded that safety-related equipment throughout the reactor building was adequately environmentally qualified for post-EPU post-LOCA conditions without the LPCI room coolers and the associated fans. The inspectors had the following concerns:
- As discussed in the previous URI, the inspectors were concerned that the associated increase in containment temperature resulting from the increased containment pressure post-EPU may also not have been considered by the licensee in determining the post-EPU maximum reactor building room temperatures post-LOCA.
Failure to do so may have underestimated the maximum reactor building (including the LPCI room) temperatures post-LOCA and the impact on the environmental qualification of equipment throughout the reactor building. The licensee initiated AR 763663763663to address the increased containment pressure post-EPU.
- The inspectors noted that Calculation DRE 00-0075, Evaluation of Dresden Environmentally Qualified (EQ) binders for EPU environmental conditions, Rev 2, dated 12-18-05, stated that safety-related equipment throughout the reactor building was adequately environmentally qualified for post-EPU post-LOCA conditions without the LPCI/CS room coolers. However, Calculation DRE 01-0041, Revision 0, dated February 21, 2007 changed the LPCI rooms post-DBA conditions.
Specifically, the post-LOCA LPCI room temperature changed from 168 to 190 degrees Fahrenheit. In addition, this revision changed the reactor building post-DBA conditions and the relevant EQ binders had not been updated.
- Calculation DRE 00-0075 assumed by engineering judgment that for zone 4 & 5 (LPCI corner rooms), the post LOCA area temperatures would be 22 degrees Fahrenheit higher based on the difference in suppression pool temperature profiles between pre-EPU and post-EPU post-LOCA conditions as shown in the Reference 5.1.2. The inspectors noted that Reference 5.1.2 did not indicate where the torus temperature profiles were calculated but rather referred back to DRE 00-0075 as the source. The licensee initiated an Issue Report (IR) indicating that calculation BSA-D-97-02, revision 1, Calculation of Long Term Suppression Pool Temperature Response for Dresden Station, was the source of the pre-EPU post-LOCA torus temperature profile and that DRE 00-0075 was a historical calculation and would not be updated.
- The inspectors noted that Calculation DRE 01-0041, revision 2 stated in Section 6.4 that the turbine building ventilation system was assumed to run after a design basis LOCA. Likewise, Assumption 3 in RSA-D-92-07, LPCI Room Temperature response due to Loss of Room Cooler at Dresden Station, stated that the turbine building remained at 104 degrees Fahrenheit throughout the post-LOCA period which also indicated the turbine building ventilation system was assumed to be operating during and after the DBA-LOCA. The inspectors questioned whether turbine building ventilation system would be running since it was not safety-related and would not be powered by the emergency diesel generators.
- The inspectors questioned the assumed heat load from one core spray and one LCPI pump. In Calculation BSA-D-00-01, DRES ECCS room temperature response with Loss of LPCI Room Cooling, Revision 0, dated February 9, 2000, Design Input 1 indicated that the heat load in the LPCI room 10 minutes after a LOCA was assumed to be 302,050 BTU/hr. This is equal to about 84 Btu/sec or 118 horsepower (HP). However, the inspector estimated that the heat load due to the two pumps would be at least 130 HP. In addition, the calculation did not consider the increased heat due to the increased current in the motor due to the reduced voltage in the induction motor due to the increased room temperature (190 vs 168 degree Fahrenheit). In addition, the previous EQ qualification calculation did not consider the increased heat due to the increased current in the motor due to the reduced voltage in the induction motor due to due to the assumption of an increased room temperature due to assumption that LPCI room coolers were not needed (168 vs 104 degree Fahrenheit).
The licensee stated that a new calculation would be completed to address the predicted post-EPU and post-DBA-LOCA temperatures in the reactor building. This issue is considered unresolved (URI 05000237/2008002-02 (DRS) URI 05000249/2008002-02)pending review of the final calculation.
Introduction:
The inspectors identified an unresolved item regarding the ability of the LPCI room safety-related equipment to perform its post-LOCA design safety function without LPCI room cooling.
Description:
The inspectors noted that Dresden EQ binders concluded that safety-related equipment throughout the reactor building was adequately environmentally qualified for post-EPU post-LOCA conditions without the LPCI room coolers and the associated fans. The inspectors noted that the motor used in the EQ test was not identical to the motors installed at Dresden. The inspectors questioned whether the motor used in testing was acceptable. These issues are considered unresolved (URI 05000237/2008002-03 (DRS) URI 05000249/2008002-03 (DRS))
pending the inspectors review of the licensing basis.
1R11 Licensed Operator Requalification Program
a. Inspection Scope
On January 28, 2008, and again on February 4, 2008, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator requalification examinations to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:
- licensed operator performance;
- crews clarity and formality of communications;
- ability to take timely actions in the conservative direction;
- prioritization, interpretation, and verification of annunciator alarms;
- correct use and implementation of abnormal and emergency procedures;
- control board manipulations;
- oversight and direction from supervisors; and
- ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.
The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements.
This inspection constitutes two quarterly licensed operator requalification program samples as defined in Inspection Procedure 71111.11.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk-significant systems:
- Unit 3 emergency diesel generator; and
- Unit 3 high pressure coolant injection system.
The inspectors reviewed events such as where ineffective equipment maintenance has resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:
- implementing appropriate work practices;
- identifying and addressing common cause failures;
- scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;
- characterizing system reliability issues for performance;
- charging unavailability for performance;
- trending key parameters for condition monitoring;
- ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and
- verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2) or appropriate and adequate goals and corrective actions for systems classified as (a)(1).
The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment.
This inspection constitutes two quarterly maintenance effectiveness samples as defined in Inspection Procedure 71111.12-05.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:
- Work Order 802846-01, 3A Core spray pump motor surveillance;
- Work Order 99053253, Dresden 2 - 17 year replace 125V DC station main battery 2-83125;
- Work Order 431632, Dresden 2 - 6 year preventative maintenance 250V DC breaker isolation condenser supply line isolation valve 1301-3;
- Work Order 1094260-01, Dresden 3 - 1 month TS Unit Diesel Generator Operability; and
- Work Order 1065405-01, 2/3 EDG [emergency diesel generator] Cooling Water Flow Reversal with U3 HPCI [high pressure coolant injection] Unavailable.
These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.
These activities constituted five samples as defined in Inspection Procedure 71111.13-05.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed the following issues:
- Issue Report 719750, Yoke Found Broke During Walkdown/Stem Also Bent;
- Issue Report 688977, Error in Reduced FW [feedwater] Temperature Region With Westinghouse Fuel;
- Operability Evaluation 08-001, Unit 3 EDG Pre-Lubrication System;
- Issue Report 728225, Higher Than Expected Amount of Air When Venting LPCI
[low pressure coolant injection] Cross Tie; and
- Issue Report 732758, Failure of Unit 3 B SBLC [standby liquid control}
Accumulator Charging Valve.
The inspectors selected these potential operability issues based on the risk-significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and UFSAR to the licensees evaluations, to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the
.
This inspection constitutes five samples as defined in Inspection Procedure 71111.15-05.
b. Findings
No findings of significance were identified.
1R18 Modifications
.1 Temporary Plant Modifications
a. Inspection Scope
The inspectors reviewed the following temporary modification:
- Engineering Change 368695, Install Test Equipment to Monitor 2A Reactor Feed Pump, Revision 1 and Engineering Change 368860, Install Test Equipment to Monitor 3A and 3B Reactor Feed Pumps, Revision 0.
The inspectors compared the temporary configuration change and associated 10 CFR 50.59 screening and evaluation information against the design basis, the UFSAR and the TS, as applicable, to verify that the modification did not affect the operability or availability of the affected system. The inspectors, as applicable, performed field verifications to ensure that the modifications were installed as directed; the modifications operated as expected; modification testing adequately demonstrated continued system operability, availability, and reliability; and that operation of the modifications did not impact the operability of any interfacing systems. Lastly, the inspectors discussed the temporary modification with operations, engineering, and training personnel to ensure that the individuals were aware of how extended operation with the temporary modification in place could impact overall plant performance.
This inspection constitutes one temporary modification sample as defined in Inspection Procedure 71111.18-05.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed the following post-maintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:
- Work order 882328-01, Two Year 3A Core Spray Pump Motor Environmental Qualification Surveillance;
- Work order 99053253: Dresden 2 - 17 year PM [preventative maintenance]
Replace 125V STA Main Battery 2-83125;
- Work order 867233-06, IM [instrument maintenance] Post Installation Test Relay 2-1530-105;
- Work order 977159, Tasks 06 and 07, Install MCC [motor control center]
Replacement Buckets per EC [engineering change] 363270;
- Work order 1094260-01, Dresden 3 - 1M [month] TS Unit Diesel Generator Operability;
- Work order 1028190, U2/3 RBCCW [reactor building closed cooling water] PP
[pump] Dead-Headed in Parallel with 3B RBCCW PP; and
- Work order 911669, D2/3 2Y [year] Standby Diesel Generator Inspection.
These activities were selected based upon the structure, system, or component's ability to impact risk. The inspectors evaluated these activities for the following (as applicable):
the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion), and test documentation was properly evaluated. The inspectors evaluated the activities against TS, the UFSAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with post-maintenance tests to determine whether the licensee was identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety. Documents reviewed are listed in the Attachment.
This inspection constitutes seven samples as defined in Inspection Procedure 71111.19.
b. Findings
No findings of significance were identified.
1R20 Outage Activities
a. Inspection Scope
The inspectors evaluated outage activities for a maintenance outage (D3M14) that began on January 8, 2008, and continued through January 21, 2008, when Unit 3 achieved full power. The inspectors reviewed activities to ensure that the licensee considered risk in developing, planning, and implementing the outage schedule.
The inspectors observed or reviewed portions of the reactor shutdown and cooldown, outage equipment configuration and risk management, electrical lineups, selected clearances, control and monitoring of decay heat removal, control of containment activities, startup and heatup activities, and identification and resolution of problems associated with the outage. The outage had been planned because of changes in feedwater heater pressures indicating the beginning of an extraction steam line failure in the condenser. About 20 minutes before the start of the maintenance outage downpower, feedwater heater pressures changed dramatically. This occurred because one of the two 24 inch diameter extraction steam line bellows failed inside the B condenser bay.
This inspection constitutes one other outage sample as defined in Inspection Procedure 71111.20-05.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:
- Unit 3, DOS 0250-02, Full Closure Timing and Exercising of Main Steam Isolation Valves, Revision 25;
Integrated Functional Test For Unit 2/3 Diesel Generator To Unit 2;
- DIS 7500-01, Unit 2/3 SBGT [standby gas treatment] Auto-Actuation and Secondary Containment Isolation System Logic System Functional Test, Revision 24; and
- Work order 01100465-01, U3 EDG Lube Oil Gallery (Lower Bulls Eye) Site Glass Empty.
The inspectors observed in plant activities and reviewed procedures and associated records to determine whether: preconditioning occurred; effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing; acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis; plant equipment calibration was correct, accurate, and properly documented; as left setpoints were within required ranges; and the calibration frequencies were in accordance with TSs, the UFSAR, procedures, and applicable commitments; measuring and test equipment calibration was current; test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied; test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used; test data and results were accurate, complete, within limits, and valid; test equipment was removed after testing; where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable; where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure; where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished; prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test; equipment was returned to a position or status required to support the performance of its safety functions; and all problems identified during the testing were appropriately documented and dispositioned in the CAP. Documents reviewed are listed in the Attachment.
This inspection constitutes three routine surveillance testing samples, one inservice inspection sample, and one containment isolation valve inspection sample as defined in Inspection Procedure 71111.22-05, for a total of five samples.
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP2 Alert and Notification System (ANS) Evaluation
a. Inspection Scope
The inspectors discussed with Emergency Preparedness (EP) staff the operation, maintenance, and periodic testing of the ANS in the Dresden Stations plume pathway Emergency Planning Zone to determine whether the ANS equipment was adequately maintained and tested in accordance with Emergency Plan commitments and procedures. The inspectors reviewed records of January 2006 through December 2007 monthly trend reports and siren test failures, as well as 2006 and 2007 maintenance documents.
These activities completed one inspection sample as defined in Inspection Procedure 71114.02-05.
b. Findings
No findings of significance were identified.
1EP3 Emergency Response Organization (ERO) Augmentation Testing
a. Inspection Scope
The inspectors reviewed and discussed with plant EP staff the emergency plan commitments and procedures that addressed the primary and alternate methods of initiating an ERO activation to augment the on-shift ERO as well as the provisions for maintaining the stations ERO call-out roster. The inspectors also reviewed reports and a sample of CAP records of unannounced off-hour augmentation tests, which were conducted January 2006 through December 2007, to determine the adequacy of the drills critiques and associated corrective actions. The inspectors also reviewed the EP training records of a sample of approximately 74 Dresden Station ERO personnel, who were assigned to key and support positions, to determine whether they were currently trained for their assigned ERO positions.
These activities completed one inspection sample as defined in Inspection Procedure 71114.03-05.
b. Findings
No findings of significance were identified.
1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies
a. Inspection Scope
The inspectors reviewed a sample of Nuclear Oversight staffs 2006 and 2007 audits of the Dresden Station emergency preparedness program to verify that these independent assessments met the requirements of 10 CFR 50.54(t). The inspectors also reviewed critique reports and samples of CAP records associated with the 2007 biennial exercise, as well as various EP drills conducted in 2006 and 2007 in order to verify that the licensee fulfilled its drill commitments and to evaluate the licensees efforts to identify, track, and resolve concerns identified during these activities. Additionally, the inspectors reviewed and discussed with station EP staff the CAP, a sample of EP items, and corrective actions related to the facilitys EP program and activities to determine whether corrective actions were acceptably completed.
These activities completed one inspection sample as defined in Inspection Procedure 71114.05-05.
b. Findings
No findings of significance were identified.
1EP6 Drill Evaluation
.1 Emergency Preparedness Drill Observation
a. Inspection Scope
The inspectors evaluated the conduct of a routine licensee emergency drill on February 4, 2008, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the simulator to determine whether the event classification, notifications, and protective action recommendations were performed in accordance with procedures. The inspectors also attended the licensee drill critique to compare any inspector-observed weakness with those identified by the licensee staff in order to evaluate the critique and to verify whether the licensee staff was properly identifying weaknesses and entering them into the CAP. As part of the inspection, the inspectors reviewed the drill package and other documents listed in the Attachment.
This inspection constitutes one sample as defined in Inspection Procedure 71114.06-05.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03)
.1 Inspection Planning and Identification of Instrumentation
a. Inspection Scope
The inspectors reviewed the plant FSAR to identify applicable radiation monitors associated with measuring transient high and very high radiation areas, including those intended for remote emergency assessment. The inspectors identified the types of portable radiation detection instrumentation used for job coverage of high radiation area work, including instruments used for underwater surveys, portable and fixed area radiation monitors used to provide radiological information in various plant areas, and continuous air monitors used to assess airborne radiological conditions and consequently work areas with the potential for workers to receive a 50 millirem or greater committed effective dose equivalent (CEDE). Whole body counters used to monitor for internal exposure and those radiation detection instruments utilized to conduct surveys for the release of personnel and equipment from the radiologically controlled area (RCA), including contamination monitors and portal monitors, were also identified.
These reviews constitute two samples as defined in Inspection Procedure 71121.03.
b. Findings
No findings of significance were identified.
.2 Calibration and Testing of Radiation Monitoring Instrumentation
a. Inspection Scope
The inspectors reviewed radiological instrumentation to determine if it had been calibrated as required by the licensees procedures, consistent with industry and regulatory standards. The inspectors also reviewed alarm setpoints for selected instruments to determine whether they were established consistent with the FSAR or TSs, as applicable, and were consistent with industry practices and regulatory guidance.
Specifically, the inspectors reviewed calibration procedures and the most recent calibration records for the following radiation monitoring instrumentation and calibration equipment:
- Unit 2 and Unit 3 Drywell High Range (Accident) Radiation Monitors;
- Unit 2 and Unit 3 Reactor Building Exhaust Radiation Monitors;
- Portal monitors used at RCA and protected area egresses;
- Personnel contamination monitors used at RCA egress;
- Instrument Calibrator (and the associated instruments used to measure calibrator output);
- Unit 2 Reactor Water Cleanup, High Pressure Coolant Injection and Air Ejector Area Radiation Monitors (ARMs);
- Unit 3 Control Rod Drive Module and Radwaste Pump Room ARMs;
- Several portable survey instruments; and
- Whole Body Counter.
The inspectors determined what actions were taken when, during calibration or source checks, an instrument was found significantly out of calibration or exceeded as-found acceptance criteria. Should that occur, the inspectors determined whether the licensees actions would include a determination of the instruments previous usages and the possible consequences of that use since the prior successful calibration. The inspectors also reviewed the results of the licensees most recent 10 CFR Part 61 source term (radionuclide mix) evaluation to determine if radiation sources used for instrument calibration and for instrument checks were representative of the plant source term.
Also, the inspectors observed the licensees use of the portable survey instrument calibration units, discussed calibrator output validation methods, and compared calibrator exposed readings with calculated/expected values. The inspectors evaluated compliance with licensee procedures while radiation protection (RP) personnel demonstrated the methods for performing source checks of portable survey instruments and source checks of personnel contamination and portal monitors.
These reviews constitute one sample as defined in Inspection Procedure 71121.03.
b. Findings
No findings of significance were identified.
.3 Problem Identification and Resolution
a. Inspection Scope
The inspectors reviewed licensee CAP documents and any Licensee Event Reports or special reports that involved personnel contamination monitor alarms due to personnel internal exposures to determine whether identified problems were entered into the CAP for resolution.
While no internal exposure with a CEDE greater than 50 millirem occurred since the last inspection in this area, the inspectors reviewed the licensees methodology for internal dose assessment to determine if affected personnel would be properly monitored using calibrated equipment and if the data would be analyzed and exposures properly assessed.
These reviews constitute one sample as defined in Inspection Procedure 71121.03.
The inspectors reviewed CAP reports related to exposure significant radiological incidents that involved radiation monitoring instrument deficiencies since the last inspection in this area, as applicable. Members of the radiation protection staff were interviewed and corrective action documents were reviewed to determine whether follow-up activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk based on the following:
- Initial problem identification, characterization, and tracking;
- Disposition of operability/reportability issues;
- Evaluation of safety significance/risk and priority for resolution;
- Identification of repetitive problems;
- Identification of contributing causes;
- Resolution of Non-Cited Violations (NCVs) tracked in the corrective action system; and
- Identification and implementation of effective corrective actions.
These reviews constitute one sample and defined in Inspection Procedure 71121.03.
The inspectors determined if the licensees self-assessment and audit activities completed for the approximate 2-year period that preceded the inspection were identifying and addressing repetitive deficiencies or significant individual deficiencies in problem identification and resolution, as applicable.
These reviews constitute one sample as defined by Inspection Procedure 71121.03.
b. Findings
No findings of significance were identified.
.4 Radiation Protection Technician Instrument Use
a. Inspection Scope
The inspectors determined whether calibrations for those survey instruments used to perform job coverage surveys and for those currently designated for use had not lapsed.
The inspectors determined if response checks of portable survey instruments and checks of instruments used for unconditional release of materials and workers from the RCA were completed prior to instrument use, as required by the licensees procedure.
The inspectors also discussed instrument calibration methods and source response check practices with radiation protection staff and observed staff demonstrate instrument source checks.
These reviews constitute one sample as defined in Inspection Procedure 71121.03.
b. Findings
No findings of significance were identified.
.5 Self-Contained Breathing Apparatus (SCBA) Maintenance/Inspection and Emergency
Response Staff Qualifications
a. Inspection Scope
The inspectors reviewed the status and surveillance records of SCBAs staged in the plant and ready for use, and evaluated the licensees capabilities for refilling and transporting SCBA air bottles to and from the control room and operation support center during emergency conditions. The inspectors determined if control room staff and other emergency response and radiation protection personnel were trained, respirator fit tested, and medically certified to use SCBAs, including personal bottle change-out.
Additionally, the inspectors reviewed SCBA qualification records for numerous members of the licensees radiological emergency teams to determine if a sufficient number of staff were qualified to fulfill emergency response positions, consistent with the licensees emergency plan and the requirements of 10 CFR 50.47.
These reviews constitute one sample as defined in Inspection Procedure 71121.03.
The inspectors reviewed the qualification documentation for at least 50 percent of the onsite, or as applicable, offsite contractor personnel that perform maintenance on manufacturer designated vital SCBA components. The inspectors also reviewed vital component maintenance records for several SCBA units that were designated as ready-for-use. The inspectors also evaluated through record review and observations if the required air cylinder hydrostatic testing was documented and current and if the Department of Transportation required retest air cylinder markings were in place for several randomly selected SCBA units and spare air bottles. The inspectors reviewed the onsite maintenance procedures governing vital component work, as applicable, including those for the low-pressure alarm and pressure-demand air regulator. The inspectors reviewed the licensees maintenance procedures and the SCBA manufacturers recommended practices to determine if there were inconsistencies between them.
These reviews constitute one sample as defined in Inspection Procedure 71121.03.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES (OA)
4OA1 Performance Indicator (PI) Verification
.1 Data Submission Issue
a. Inspection Scope
The inspectors performed a review of the data submitted by the licensee for the First Quarter 2008 performance indicators for any obvious inconsistencies prior to its public release in accordance with IMC 0608, Performance Indicator Program.
This review was performed as part of the inspectors normal plant status activities and, as such, did not constitute a separate inspection sample.
b. Findings
No findings of significance were identified.
.2 Unplanned Scrams per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Scrams per 7000 Critical Hours PI for Units 2 and 3 for the period from the First Quarter 2007 through the Fourth Quarter 2007 to determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in Revision 5 of the Nuclear Energy Institute (NEI)
Document 99-02, Regulatory Assessment Performance Indicator Guideline, were used.
The inspectors reviewed the licensees operator narrative logs, issue reports, event reports and NRC Inspection reports for the period of January 2007 through January 2008 to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the Attachment to this report.
This inspection constitutes two unplanned scrams per 7000 critical hours samples as defined in Inspection Procedure 71151.
b. Findings
No findings of significance were identified.
.3 Unplanned Scrams with Complications
a. Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Scrams with Complications performance indicator for Unit 2 and 3 for the period from the First Quarter 2007 through the Fourth Quarter 2007 to determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in Revision 5 of the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, were used. The inspectors reviewed the licensees operator narrative logs, issue reports, event reports and NRC Integrated Inspection reports for the period of January 2007 through January 2008 to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the Attachment to this report.
This inspection constitutes two unplanned scrams with complications samples as defined in Inspection Procedure 71151.
b. Findings
No findings of significance were identified.
.4 Unplanned Transients per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Transients per 7000 Critical Hours PI for Units 2 and 3, for the period from the First Quarter 2007 through the Fourth Quarter 2007 to determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in Revision 5 of the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, were used. The inspectors reviewed the licensees operator narrative logs, issue reports, maintenance rule records, event reports and NRC Integrated Inspection reports for the period of January 2007 through January 2008 to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the Attachment to this report.
This inspection constitutes two unplanned transients per 7000 critical hours samples as defined in Inspection Procedure 71151.
b. Findings
No findings of significance were identified.
Cornerstone: Public Radiation Safety
.5 Radiation Safety Strategic Area
a. Inspection Scope
The Inspectors sampled the licensees PI submittals for the period indicated below. The inspectors used PI definitions and guidance contained in Revision 5 of NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, to verify the accuracy of the PI data. The following PI was reviewed:
- Radiological Effluent Technical Specification/Offsite Dose Calculation Manual (RETS/ODCM) Radiological Effluent Occurrence.
The inspectors reviewed data associated with the RETS/ODCM PI to determine if the indicator was accurately assessed and reported. The inspectors reviewed the licensees assignment report (AR) database and individual ARs generated in 2007 to identify any potential occurrences such as unmonitored, uncontrolled or improperly calculated effluent releases that may have impacted offsite dose. The inspectors also reviewed gaseous and liquid effluent summary data and the results of associated offsite dose calculations for selected periods in 2007 to determine if indicator results were accurately reported. The inspectors also discussed with the licensee the methods for quantifying gaseous and liquid effluents and for determining effluent dose.
These reviews constitute one sample as defined in Inspection Procedure 71151.
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
.6 Emergency Preparedness Strategic Areas
a. Inspection Scope
The inspectors reviewed the licensees records associated with the three EP performance indicators (PIs) listed below. The inspectors verified that the licensee accurately reported these indicators in accordance with relevant procedures and NEI guidance endorsed by the NRC. Specifically, the inspectors reviewed licensee records associated with PI data reported to the NRC for the period January 2007 through September 2007. Reviewed records and processes discussed included: procedural guidance on assessing opportunities for the three PIs; assessments of PI opportunities during predesignated control room simulator training sessions, the 2007 biennial exercise, and other drills; revisions of the roster of personnel assigned to key Emergency Response Organization (ERO) positions; and results of periodic Alert and Notification System operability tests. The following PIs were reviewed:
Common
- Alert and Notification System;
- ERO Drill Participation; and
- Drill and Exercise Performance.
These reviews constitute three samples as defined in Inspection Procedure 71151.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
.1 Routine Review of items Entered Into the CAP Scope
As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Attributes reviewed included: the complete and accurate identification of the problem; that timeliness was commensurate with the safety significance; that evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrence reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue.
Minor issues entered into the licensees CAP as a result of the inspectors observations are included in the attached List of Documents Reviewed.
These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.
a. Findings
No findings of significance were identified.
.2 Daily Corrective Action Program Reviews
a. Scope
In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees CAP. This review was accomplished through inspection of the stations daily condition report packages.
These daily reviews were performed by procedure as part of the inspectors daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.
b. Findings
No findings of significance were identified.
.3 In-depth Review
Identification and Corrective Actions associated with IR 720709, Revision to M&TE
[maintenance and test equipment] Calculation DRE98-0047 Extended Multiple Times.
a. Effectiveness of Problem Identification
- (1) Inspection Scope The inspectors reviewed IR 720709, Revision to M&TE Calculation DRE98-0047 Extended Multiple Times, regarding corrective actions associated with IR 372385, Calculation Number DRE98-0047 Requires Revision.
- (2) Issues None.
b. Prioritization and Evaluation of Issues
- (1) Inspection Scope The inspectors reviewed IR 720709, Revision to M&TE Calculation DRE98-0047 Extended Multiple Times, regarding corrective actions associated with IR 372385, Calculation Number DRE98-0047 Requires Revision. The inspectors also reviewed MA-AA-716-040, Control of Portable Measurement and Test Equipment Program, EC 351583, Evaluate Yokogawa FG110/FG300 Function Generators For Use As M&TE, The inspectors interviewed design engineering personnel and instrument maintenance personnel.
- (2) Issues None. The licensee has been slow to update the calculation. However, the calculation appears to be a tool only. The evaluations required by the stations M&TE program document were all correct and in place.
c. Effectiveness of Corrective Actions
- (1) Inspection Scope The inspectors reviewed IR 720709, Revision to M&TE Calculation DRE98-0047 Extended Multiple Times, regarding corrective actions associated with IR 372385, Calculation Number DRE98-0047 Requires Revision. The inspectors also reviewed MA-AA-716-040, Control of Portable Measurement and Test Equipment Program, Engineering Change (EC) 351583, Evaluate Yokogawa FG110/FG300 Function Generators For Use As M&TE, The inspectors interviewed design engineering personnel and instrument maintenance personnel.
- (2) Issues None.
This represented one inspection sample as an in-depth review as defined in Inspection Procedure 71152.
d. Findings
No findings of significance were identified.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel and Activities
a. Inspection Scope
During the inspection period, the inspectors conducted the following observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours.
- Multiple tours of operations within the Central and Secondary Security Alarm Stations;
- Tours of selected security towers/security officer response posts;
- Security force shift turnover activities; and
- Observed a security force drill.
These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status review and inspection activities.
b. Findings
No findings of significance were identified.
4OA6 Meetings
.1 Exit Meeting Summary
On April 9, 2008, the inspectors presented the inspection results to Mr. T. Hanley, and other members of the licensee staff. The licensee acknowledged the issues presented.
The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified. An additional telephone discussion between Ms. A. M. Stone and Mr. J. Ellis was conducted on May 2, 2008, to describe and clarify the three unresolved items discussed in Section
1R07 , Triennial Heat Sink Performance.
.2 Interim Exit Meetings
Interim exits were conducted for:
- Occupational radiation safety cornerstone radiation monitoring instrumentation and protective equipment with Mr. D. Wozniak and other licensee staff on January 18, 2008.
- Emergency Preparedness inspection with Mr. D. Wozniak on January 18, 2008.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- D. Wozniak, Site Vice President
- T. Hanley, Plant Manager
- K. Aleshire, Exelon Corporate Emergency Preparedness Manager
- C. Barajas, Operations Director
- H. Bush, Radiation Protection Manager
- J. Ellis, Regulatory Assurance Manager
- D. Galanis, Design Engineering Manager
- D. Glick, Shipping Specialist
- G. Graff, Operations Training Manager
- J. Griffin, Regulatory Assurance - NRC Coordinator
- D. Leggett, Nuclear Oversight Manager
- M. Overstreet, Lead RP Supervisor
- C. Podczerwinski, Maintenance Rule Coordinator
- P. Quealy, Emergency Preparedness Manager
- E. Rowley, Chemistry
- R. Rybak, Regulatory Assurance
- J. Sipek, Engineering Director
- N. Starcevich, Radiation Protection Instrumentation Coordinator
- J. Strmec, Chemistry Manager
- C. Symonds, Training Director
NRC
- M. Ring, Chief, Division of Reactor Projects, Branch 1
Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened:
- 05000237/2008002-01; URI LPCI Heat Exchangers Design Calculation Deficiencies
- 05000249/2008002-01 and Discrepancies
- 05000237/2008002-02; URI LPCI Room Cooler Calculation Deficiencies and
- 05000249/2008002-02 Discrepancies
- 05000237/2008002-03; URI LPCI Room Equipment EQ Qualification Deficiencies
- 05000249/2008002-03 and Discrepancies Closed:
None Discussed:
None Attachment