IR 05000293/2004008
| ML050140113 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/14/2005 |
| From: | Doerflein L NRC/RGN-I/DRS/SSB |
| To: | Balduzzi M Entergy Nuclear Operations |
| References | |
| IR-04-008 | |
| Download: ML050140113 (26) | |
Text
January 14, 2005
SUBJECT:
PILGRIM NUCLEAR POWER STATION - NRC INSPECTION REPORT 05000293/2004008
Dear Mr. Balduzzi:
On December 3, 2004, the U. S. Nuclear Regulatory Commission (NRC) completed an engineering team inspection at Pilgrim Nuclear Power Station. The enclosed report presents the results of that inspection, which were discussed at an exit meeting on December 3, 2004,
with Messrs. P. Dietrich, R. Smith and other members of your staff.
This 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 inspection included plant walkdowns; examination of selected procedures, drawings, modifications, calculations, surveillance tests and maintenance records; and interviews with station personnel.
The report documents two NRC identified findings of very low safety significance (Green), both of which were determined to involve violations of NRC requirements. However, because of their very low safety significance and because they have been entered into your corrective action program, the NRC is treating these issues as Non-Cited Violations (NCV) consistent with Section VI.A of the NRC Enforcement Policy. If you contest the NCVs in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U. S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Pilgrim Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document
Mr. Michael Balduzzi-2-Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Lawrence T. Doerflein, Chief Safety Systems Branch Division of Reactor Safety Docket No. 50-293 License No. DPR-35
Enclosure:
Inspection Report 05000293/2004008 w/Attachment: Supplemental Information
Mr. Michael Balduzzi-3-
REGION I==
Docket No.
50-293 License No.
DPR-35 Report No.
05000293/2004008 Licensee:
Entergy Nuclear Operations, Inc.
Facility:
Pilgrim Nuclear Power Station (PNPS)
Location:
600 Rocky Hill Road Plymouth, MA 02360 Inspection Period:
November 15, 2004 - December 3, 2004 Inspectors:
F. Bower, Senior Reactor Inspector, DRS (Team Leader)
L. Cheung, Senior Reactor Inspector L. Scholl, Senior Reactor Inspector J. Lilliendahl, Reactor Inspector (Trainee)
A. Passarelli, Reactor Inspector J. Talieri, Reactor Inspector Approved by:
Lawrence T. Doerflein, Chief Safety Systems Branch Division of Reactor Safety
ii Enclosure
SUMMARY OF FINDINGS
IR 05000293/2004008; 11/15/2004 - 12/03/2004; Entergy Nuclear Operations, Inc., Pilgrim
Nuclear Power Station, Safety System Design and Performance Capability.
The inspection was conducted by six regional inspectors. The inspection identified two findings of very low safety significance (Green) that were also non-cited violations. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.
The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
C
- Green.
The team identified a Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, because the licensee failed to establish adequate measures to assure that the design basis minimum water level in each salt service water (SSW) system pump well of the intake structure was correctly translated into the Technical Specifications and SSW System Operating Procedures.
This issue was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objectives of ensuring the availability, reliability, and capability of systems and components that respond to initiating events. Specifically, the lower level specified for ensuring SSW system operability had the potential to affect the capability of the SSW system to perform its safety-related function under worst case design basis loss of coolant accident (DBA-LOCA) conditions.
The issue screened as very low safety significance (Green) in Phase I of the SDP, because it was a design deficiency that was not found to result in a loss of function. The team did not identify any examples where the minimum water level in the pump wells of the Intake Structure was less than design basis minimum water level.
The team also identified that a contributing cause of the finding was related to the problem identification and resolution cross-cutting area, in that, although inconsistencies between the Updated Final Safety Analysis Report (UFSAR) and the SSW design basis document (DBD) regarding the SSW pump minimum water levels relative to mean sea level (msl) were identified during the DBD development process and during previous SSW assessments, these issues were not appropriately resolved. (Section 1R21.2)
C
- Green.
The team identified a Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XI, Test Control, because the licensees station battery test program lacked adequate provisions to assure that all testing prerequisites iii were met and to assure that the available test equipment was adequately used for three cycles of Technical Specification (TS) required surveillance testing of the 125V A & B station batteries and the 250V station battery.
The finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events. Specifically, the lack of procedure quality and detail led to repetitive instances where battery testing was not completed without error. The issue screened as very low safety significance (Green) in Phase I of the SDP, because it was a procedure quality issue that did not result in a loss of function since the capacity margin in the design of the batteries has enabled the licensee to perform engineering evaluations for the incorrectly performed testing and demonstrate operability.
The team also identified that a contributing cause of the finding was related to the problem identification and resolution cross-cutting area, in that, the licensee reviewed each of these events narrowly, determined that each was an isolated case and failed to identify the adverse trend of procedure inadequacies that contributed to the repetitive events. (Section 1R21.3)
Licensee-Identified Violations
None
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
==1R21 Safety System Design and Performance Capability (IP 71111.21)
1.
==
Inspection Sample Selection Process In selecting systems and components for review, the team focused on risk significance and considered the risk information contained in the licensees Probabilistic Risk Assessment (PRA) and the U.S. Nuclear Regulatory Commissions (NRC) Simplified Plant Analysis Risk (SPAR) models. Using the risk insights, the team selected the Salt Service Water (SSW) and Direct Current (DC) Power systems for review. In selecting the components for review, the team considered risk importance measures, such as risk achievement worth (RAW) values, from the PRA and SPAR models, as well as the maintenance and modification history, and operating experience.
The team reviewed design and licensing basis documents for the two systems to understand the system needs, safety functions and regulatory requirements. The documents reviewed included the applicable technical specifications (TS), updated final safety analysis report (UFSAR), design basis documents (DBD) and design calculations.
A list of documents reviewed is included in the attachment to this report. The teams inspection activities were focused on verifying that design bases were being correctly implemented for the selected systems and components to ensure that the systems can be relied upon to meet their design basis functional requirements.
2. Salt Service Water System
a. Inspection Scope
The team reviewed the piping and instrumentation drawings, other supporting documents (e.g., system health reports) and conducted plant walkdowns of the service water intake structure and other accessible portions of the SSW system to verify the physical installation was consistent with the design basis. In addition, during these walkdowns, the team evaluated the material condition of the plant to determine if Entergy personnel were adequately identifying and correcting material equipment problems. The team also visited the main control room, performed control board checks and discussed SSW system design and operation with the licensed operators.
In addition, the team interviewed cognizant system engineers and design engineers regarding the system design, operation, and performance. The team reviewed control diagrams, setpoint calculations, calibration procedures and surveillance tests to verify the capability of the SSW instrumentation and controls to respond to design basis transient and accident conditions. The team reviewed a selected sample of system operating procedures, off-normal operating procedures, and valve line-up lists to determine that they adequately controlled the plant configuration and supported operator actions assumed in the design basis.
The risk significant components selected for detailed review by the team included the SSW pumps and their associated controls. The team reviewed a sample of SSW periodic surveillance test procedures to ensure the tests demonstrated the required component functions, and that the acceptance criteria were consistent with the design basis assumptions and the pump curves. To verify that acceptance criteria were met and that problems identified through testing were corrected, the team reviewed completed surveillance tests and engineering evaluations. The team also reviewed inservice testing (IST) results to verify that acceptance criteria were met and the tested components were appropriately categorized.
The team also reviewed a selected sample of procedures, test and maintenance records and the licensees commitments relative to six SSW system motor operated butterfly valves. The review was done to assess the implementation of the licensees program for periodic testing of motor-operated valves and for implementing NRC Generic Letter (GL) 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Power-Operated Valves.
Relative to potential system and component degradation, the team conducted interviews with PNPS inservice inspection (ISI), systems and design engineering personnel and reviewed design specifications and plant design change documents regarding the material selection processes used for the SSW pumps, piping materials and pipe coatings. The team verified the structural integrity of these components through the review of design calculations, problem reports and corrective actions, field revision notices, and test results. The inspector also reviewed the ISI program and the piping spool database used in the visual inspection of the SSW piping.
To verify the effectiveness of the licensees corrective actions related to SSW system design issues, the team selected the 1999 closeout report for the service water operational performance inspection (SWOPI) for a follow up evaluation. The evaluation included a review of the adequacy of the licensees implementation of corrective actions that included procedure, UFSAR, and diesel generator loading calculation changes.
b. Findings
Introduction.
The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to establish adequate measures to assure that the design basis minimum water level in each SSW pump well of the intake structure was correctly translated into the Technical Specifications and SSW System Operating Procedures. This finding was determined to be of very low safety significance since the team did not identify an actual loss of safety function of the SSW System as a result of low level in the SSW intake bay.
Description.
UFSAR Section 10.7, Salt Service Water System, describes, in part, the relationship between the SSW system performance analysis and the sea water tide level used in accident analysis calculations as 7.1 feet below mean sea level (msl). The
-7.1 foot level is also the yearly astronomical minimum low tide. The -7.1 foot value was chosen for the design basis analysis of the minimum SSW system performance required to perform the emergency containment cooling function. This lowest tide level was assumed to be a constant, thereby yielding a conservatively low SSW system flow rate during accident analyses that span a several day period. The SSW pumps are also assumed to be operating at their minimum performance (allowed by inservice testing)thereby providing only the required 4500 gpm to the reactor building closed cooling water (RBCCW) heat exchanger. The team concluded that this description was consistent with the supporting design calculation, but noted that the potential for reducing level in an SSW pump well of the intake structure due to failure of the non-safety-related screen wash system and debris impinging on the traveling screens was not considered.
In addition to the SSW system design basis, the UFSAR also describes the manufacturers design rated performance of the SSW pumps. UFSAR Section 10.7 states that the Technical Specifications originally described the minimum required SSW pump performance as 2700 gpm at 55 ft total developed head (TDH). The team noted that this was before the re-analysis (design calculation M630) of the minimum SSW system performance required to perform the emergency containment cooling function.
The actual manufacturers rating of the SSW pump is 2700 gpm at 95 ft TDH and minimum required performance for in-service testing is defined as 2700 gpm at 87.5 ft TDH. These TDH values are across the pump bowl not including the 40 ft vertical pump column. At this minimum performance point (87.5 ft TDH), the minimum water level in the SSW pump wells of the intake structure must be greater than or equal to 7.1 feet below msl.
The 55 ft value represents the minimum required pressure, in feet, measured at the centerline of the pump discharge piping (EL 23.9 ft) for a pump bowl operating at 2700 GPM at 87.5 ft TDH. The minimum sea water level for maintaining SSW pump manufacturers design performance is approximately 139 below msl. This represents the lowest sea water level at which an SSW pump bowl operating at its rated performance of 95 ft TDH at 2700 gpm will produce a discharge head of 55 ft at 2700 gpm as measured at EL 23.9 ft.
However, the team noted that based on the SSW system performance analysis described above, SSW pumps operating at the minimum required performance will not satisfy the system requirements unless the pump suction is submerged in an SSW pump well of the intake structure at a level equivalent to 7.1 ft below msl. The team confirmed this analysis through the review of the four applicable cases in design calculation M630, SSW System Hydraulic Analysis, that all assumed a minimum submergence of the SSW pumps suctions at the level of 7.1 ft below msl. This information was confirmed in Section 5.2.2, Minimum Water Level of the system design basis document (SDBD)-29 for the SSW system.
The team found that this design control error regarding the minimum level in an SSW pump well of the intake structure was carried into several plant documents. Pilgrim Technical Specification 4.5.B.4.1 specifies the surveillance requirement to verify the water level in the pump wells of the Intake Structure as greater than or equal to 13 feet, 9 inches below msl to ensure SSW system operability. This is reiterated in Section 5.3.2.6 of SDBD-29 for the SSW system. The error was also translated into the definition for operability of the ultimate heat sink in procedure PNPS 2.2.32, Salt Service Water System (SSW).
Analysis.
The performance deficiency was the licensees failure to establish adequate measures to assure that the design basis minimum water level in each SSW pump well of the intake structure was correctly translated into the Technical Specifications and SSW System Operating Procedures. Specifically, the minimum water level above the SSW pump suction (-7.1 feet relative to mean sea level (msl) of 0 feet, 0 inches),assumed in design basis calculation M-630 to assure that the residual heat removal (RHR), RBCCW and SSW systems are operated to maximize their containment heat removal capability under worst case Design Basis Loss of Coolant Accident (DBA-LOCA) was not correctly translated into Pilgrim Technical Specifications 3/4.5.
Also, the -7.1 foot level was not correctly translated into SSW system operating procedure 2.2.32. Both of these documents specify the minimum water level (to assure system operability) in the pump wells of the Intake Structure as greater than or equal to 13 feet, 9 inches below msl.
This issue was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objectives of ensuring the availability, reliability, and capability of systems and components that respond to initiating events. Specifically, the lower level specified for ensuring SSW system operability had the potential to affect the capability of the SSW system to perform its safety-related function under worst case DBA-LOCA conditions.
The issue screened as very low safety significance (Green) in Phase I of the SDP, because it was a design deficiency that was not found to result in a loss of function.
During the inspection, the team did not identify any recent examples of where the minimum water level in the pump wells of the Intake Structure were less than -7.1 feet.
The team also identified that a contributing cause of the finding was related to the problem identification and resolution cross-cutting area, in that, although inconsistencies between the UFSAR and the SSW DBD regarding the SSW pump minimum water levels relative to msl were identified during the DBD development process (OI-SSW-2) and during previous SSW assessments (SW95.0020.01), these issues were not appropriately resolved (CR-PNP-2004-3707).
Enforcement.
10 CFR Part 50, Appendix B, Criterion III, Design Control, requires measures shall be established to assure that the design basis systems and components are correctly translated into specifications and procedures. Contrary to the above, Pilgrim incorrectly translated the design basis minimum water level (-7.1 feet below mean sea level as assumed in design bases calculations M630 and M500) in each SSW pump well of the intake structure to Technical Specifications 3/4.5 and the SSW system operating procedure 2.2.32 which specify the minimum water level as greater than or equal to 13 feet, 9 inches below msl. Because this finding is of very low safety significance and has been entered into the PNPS corrective action program (CR-PNP-2004-3832 and CR-PNP-2004-3707), it is being treated as a non-cited violation, consistent with Section VI.A of the NRCs Enforcement Policy.
(NCV 05000293/2004008-01; Failure To Adequately Translate Design Basis Minimum SSW Pump Well Level to Technical Specifications)
3. 125 and 250 Volt Direct Current (Vdc) System
a. Inspection Scope
The team reviewed the one-line diagrams for the dc power system and distribution panels, and other supporting documents (e.g., system health reports). The team also reviewed the heat, ventilation and air-conditioning (HVAC) drawings to confirm that the battery rooms were adequately ventilated to prevent hydrogen accumulation. The team walked down all station batteries, battery chargers, and associated switchgear rooms to verify that the physical installation was consistent with the design basis and to confirm the operating status of the battery ventilation systems. The team also reviewed the physical condition of the batteries and battery chargers to help determine that Entergy personnel were adequately identifying and correcting material equipment problems.
The team reviewed the loading and voltage-drop calculations of the three safety-related batteries and the sizing of associated battery chargers to verify that all dc loads were accounted for and that adequate voltages could be provided for the safety equipment to operate up to eight hours following a postulated design basis accident. The fuses and circuit breakers of the dc system were reviewed to confirm that they were properly coordinated and adequately rated for the expected short circuit currents. The control logic diagrams of the automatic switching circuitry of the swing bus was reviewed to confirm the adequacy of the design. Also, included in the review were the dc ground detection system and the vendor manuals.
In addition, the team interviewed cognizant system engineers and design engineers regarding the system design, operation, and performance. The dc system operating procedures were reviewed to verify that they adequately controlled the plant configuration and supported operator actions assumed in the design basis. In addition, the team reviewed the test results of the battery weekly, quarterly, service, and performance tests, battery charger tests, and the test of the automatic switching of the swing bus, to ensure that operability status was demonstrated.
The risk significant components selected for detailed review by the team included the three station batteries. The team reviewed the service and performance tests procedures for the batteries to determine whether the calculated voltages and capacity of the batteries were appropriately translated into the test procedures.
To verify the effectiveness of the licensees corrective actions related to station battery issues, the team selected the station battery testing program for a follow up evaluation based on inspector identified findings in NRC Inspection Report (IR)05000293/2000012.
The follow-up evaluation was performed to determine if the licensees corrective actions were appropriate and if testing was being performed in accordance with the battery design bases. The inspectors reviewed the station battery (125 & 250 Vdc) test procedures and revision histories, records of station battery testing completed subsequent to IR 05000293/2000012 and the design calculations for sizing of the batteries and determining plant minimum voltages. The inspectors also conducted interviews with knowledgeable battery testing personnel including the DC systems engineer and DC design engineer.
b. Findings
Introduction.
The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B Criterion XI, Test Control, because the licensees station battery test program lacked adequate provisions to assure that all testing prerequisites were met and to assure that the available test equipment was adequately used for three cycles of TS required surveillance testing of the 125V A & B station batteries and the 250V station battery. The issue screened as very low safety significance (Green) because it was a procedure quality issue that did not result in a loss of battery function.
Description.
The team reviewed the results of the 125V and 250V station battery TS surveillance testing conducted for the last three operating cycles (1999, 2001 and 2003).
The team noted that the 1999 service test procedures for both 125V and the 250V station batteries lacked the necessary detail to properly control the testing with respect to recording and documenting test results. This finding (NCV 05000293/2000-012-03)was identified by the NRC in 2000, when the inspectors noted that the licensee could not demonstrate that the batteries were correctly discharged according to their service test duty cycle and that the test records were insufficient during the critical first hour of each of the tests. The licensees root cause analysis (PR00.9494) concluded that the root cause was that the test equipment failed to function as expected during battery testing and the test equipment failure went unrecognized because the test procedures lacked necessary detail to properly control the test.
In 2001, the licensee identified that improper use of the battery test equipment led to early termination of the 250V battery test. This issue was documented in PR01.1893.
This team observed that the test procedure lacked the necessary detail regarding the operation of the timer, resulting in the early test termination. In 2003, the 250V station battery test procedure lacked the necessary detail to prevent improperly connecting the test equipment to the 250V battery. Discussions with licensee personnel revealed that the connections were made based on hand written sketches developed during a training class for using the test equipment. The improper connection resulted in damaging test equipment (a load bank) and causing current oscillations in excess of 1200 amperes.
The design test value was 900 amperes. The inspectors noted that the apparent cause (CR-PNP-2003-01876) for the event was a lack of critical guidance in the test procedure. The team verified that additional inspector identified discrepancies in the battery test procedures were entered into the PNPS corrective action program (CR-PNP-2004-3659 and CR-PNP-2004-3828).
Analysis.
The performance deficiency was the licensees failure to establish an adequate test program and adequate procedures for testing the station batteries.
Specifically, the test program lacked adequate provisions to assure that all battery testing prerequisites were met and to assure that the available test equipment was adequately used. The finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events. Specifically, the lack of procedure quality and detail led to repetitive instances where battery testing was not completed without error. The issue screened as very low safety significance (Green) in Phase I of the SDP, because it was a procedure quality issue and a qualification deficiency that did not result in a loss of safety function. The capacity margin in the design of the batteries has enabled the licensee to perform engineering evaluations for the incorrectly performed testing and demonstrate operability.
The team also identified that a contributing cause of the finding was related to the problem identification and resolution cross-cutting area, in that, the licensee reviewed each of these events narrowly and determined that each was an isolated case. The licensee failed to identify the adverse trend of human errors and the collective significance of events during the performance of battery testing caused by procedure inadequacies that contributed to the repetitive events. The training provided prior to the 2003 event was not an adequate barrier to overcome the lack of detail in the procedures and preclude the human error experienced during connection of the test equipment.
Enforcement.
10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires that a test program be established to assure that all testing required to demonstrate that systems and components will perform satisfactorily in service is performed in accordance with written test procedures and the test procedures shall include provisions for assuring that all prerequisites for the given test have been met, and that adequate test instrumentation is available and used. Contrary to the above, the licensees test program did not establish adequate procedures to assure that all prerequisites were met and to assure that the available test equipment was adequately used, for testing the station batteries. Specifically, test procedures for the batteries lacked adequate detail and the training on the battery test equipment was inadequate, as evidenced by the repetitive failures to correctly connect and use the equipment for testing the station batteries. Because this finding is of very low safety significance and the licensee entered this issue into its corrective action program (CR-PNP-2004-03820), it is considered a non-cited violation consistent with Section VI.A.1 of the NRCs Enforcement Policy. (NCV 05000293/2004008-02; Inadequate Program for Station Battery Test Control)
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
(IP 71152)
4. Annual Sample Review
Not applicable.
5. Cross Reference to PI&R Findings Documented Elsewhere
Section 1R21.2 of this report describes a finding associated with licensees inadequate design controls and failure to adequately translate design basis minimum SSW pump well level into the TS. Although the licensee had several opportunities to identify the design control problems related to this parameter, sufficient corrective actions were not developed or implemented.
Section 1R21.3 of this report describes a finding associated with the licensees inadequate test control and failure to identify the adverse trend of human errors and the collective significance of events during the performance of battery testing caused by procedure inadequacies that contributed to the repetitive events. Although the licensee had several opportunities to identify the repetitive problems, sufficient corrective actions were not developed or implemented since each of these events were reviewed narrowly and determined to be an isolated case.
4OA6 Exit Meeting Summary
The team presented the inspection results to Messrs. P. Dietrich, R. Smith and other members of the Pilgrim staff at the conclusion of the inspection on December 3, 2004.
The team verified that the inspection report does not contain proprietary information.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- P. Dietrich, General Manager - Plant Operations
- R. Smith, Director - Engineering
- S. Bethay, Director - Nuclear Safety Assurance
- T. White, Manager - Design Engineering
- B. Ford, Manager - Licensing
- F. McGinnis, Licensing
- R. Pace, Supervisor - Design Engineering
- S. Wollman, Supervisor - Engineering
- J. Gaedtke, SSW System Engineer
- B. Sullivan, Assistant Operations Manager - Shifts
- R. Daverio, Supervisor - Electrical Design Engineering
- N. Eisenmann, Supervisor - I&C Design Engineering
- D. Landecite, Manager - CA&A
NRC
- W. Raymond, Senior Resident Inspector - Pilgrim
- A. Ziedonis, Reactor Engineer (Trainee)
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000293/2004008-01 NCV Failure To Adequately Translate Design Basis Minimum SSW Pump Well Level to Technical Specifications (Section 1R21.2)
- 05000293/2004008-02 NCV Inadequate Program for Station Battery Test Control (Section 1R21.3 )