IR 05000282/2003003

IR 05000282-03-003(DRS) & IR 05000306-03-003(DRS) on 03/24/2003 - 04/11/2003, Prairie Island Nuclear Generating Plant, Units 1 and 2, Nuclear Management Co. Safety System Design and Performance Capability Inspection

ML031420406
Person / Time
Site:	Prairie Island
Issue date:	05/20/2003
From:	Julio Lara NRC/RGN-III/DRS/EEB
To:	Solymossy J Nuclear Management Co
References
IR-03-003
Download: ML031420406 (34)

Text

May 20, 2003

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY INSPECTION REPORT 50-282/03-03(DRS) AND 50-306/03-03(DRS)

Dear Mr. Solymossy:

On April 11, 2003, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Prairie Island Nuclear Generating Plant, Units 1 and 2. The enclosed safety system design and performance capability inspection report documents the inspection findings, which were discussed on April 11, 2003, with you and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Specifically, this inspection focused on the design and performance capability of the residual heat removal system, safety injection system and selected portions of the chemical and volume control system to ensure that they were capable of performing their required safety related functions.

Based on the results of this inspection, the inspectors identified three findings of very low safety significance (Green), all of which were determined to involve violations of NRC requirements.

However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these three findings as non-cited violations (NCVs) consistent with Section VI.A of the NRC Enforcement Policy. If you contest any NCV 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 Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator Region III; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Prairie Island Nuclear Generating Plant, Units 1 and 2.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publically Available Records (PARS) component of NRCs 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/

Julio Lara, Chief Electrical Engineering Branch Division of Reactor Safety Docket Nos: 50-282; 50-306 License Nos: DPR-42; DPR-60 Enclosure:

Inspection Report 50-282, 306/03-03(DRS)

w/Attachment: Supplemental Information cc w/encl:

Plant Manager, Prairie Island R. Anderson, Executive Vice President Mano K. Nazar, Senior Vice President John Paul Cowan, Chief Nuclear Officer Manager, Regulatory Affairs Jonathan Rogoff, Esquire General Counsel Nuclear Asset Manager Commissioner, Minnesota Department of Health State Liaison Officer, State of Wisconsin Tribal Council, Prairie Island Indian Community Adonis A. Neblett, Assistant Attorney General Office of the Attorney General Administrator, Goodhue County Courthouse Commissioner, Minnesota Department of Commerce Gene Wilson Commissioner, Minnesota Department of Commerce

SUMMARY OF FINDINGS

IR 05000282/2003-003(DRS), 05000306/2003-003(DRS); 03/24/2003 - 04/11/2003; Prairie

Island Nuclear Generating Plant, Units 1 and 2; Safety System Design and Performance Capability Inspection.

The report covered a three week period of inspection by regional engineering specialists with mechanical engineering consultant assistance. The inspection focused on the design and performance capability of the residual heat removal system, safety injection system and selected portions of the chemical and volume control system to ensure that they were capable of performing their required safety related functions. Three Green non-cited violations (NCVs)of very low safety significance were identified. The significance of most findings is indicated by their color (Green, White, Yellow, 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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

Cornerstone: Mitigating Systems

Green.

The inspection team identified a non-cited violation of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, in that, the design bases for the Units 1 and 2 residual heat removal (RHR) discharge overpressure interlock removal modification was not correctly translated into specifications, procedures, and instructions. Specifically, the modifications safety evaluation took credit for local operator action to manually open the RHR heat exchanger to safety injection pump suction valves during the transfer to recirculation in both units emergency operating procedures (EOPs). However, on March 14, 2003, local operator action to manually open the valves was removed from the EOPs.

This finding was greater than minor because the lack of coordination between the modifications design requirements and EOP procedural guidance affected the mitigating systems cornerstone objective. The cornerstones objective of ensuring the availability, reliability, and capability of the emergency core cooling system to respond to initiating events was affected. The finding was of very low safety significance because it did not represent an actual loss of a safety function. (Section 1R21.2b.1)

Green.

The inspection team identified a non-cited violation of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, in that, the design bases for the residual heat removal (RHR) system was not correctly maintained in accordance with regulatory requirements. Specifically, a safety evaluation was written for the change in classification from safety related to non-safety related for the Units 1 and 2 RHR heat exchanger flow control valves positioners, hand controllers and signal converters. However, the safety evaluation failed to consider all credible failures in evaluating the single failure criterion. For example, if a required open valves hand controller were to fail high, the valve would close and block the emergency core cooling system (ECCS) flow path.

iii This finding was greater than minor because the change in classification from safety related to non-safety related for the Units 1 and 2 RHR heat exchanger flow control valve components affected the mitigating systems cornerstone objective. The cornerstones objective of ensuring the availability, reliability, and capability of the ECCS to respond to initiating events was affected. The finding was of very low safety significance because it did not represent an actual loss of a safety function. (Section 1R21.2b.2)

Green.

The inspection team identified a non-cited violation of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, due to the licensees failure to maintain the design basis configuration of the residual heat removal (RHR) pit covers. Specifically, the Units 1 and 2 auxiliary buildings RHR pit covers were designed to be closed during plant operation to limit the radiological dose rates to vital plant areas during accident conditions. However, prior to April 4, 2003, the Units 1 and 2 RHR pit covers were maintained in an open position during plant operation.

This finding was greater than minor because the potential to affect the safety injection and RHR systems design basis functions (i.e., degradation of long term heat removal) affected the mitigating systems cornerstone objective.

Specifically, local operator actions in the auxiliary building (e.g., area around the RHR pits) were required to transfer the emergency core cooling system (ECCS)to the recirculation mode. If the operator was prevented from performing the local operator actions during accident conditions due to high dose rates, then both trains of ECCS could be degraded. As a result, the cornerstones objective of ensuring the availability, reliability, and capability of the ECCS to respond to initiating events was affected. The finding was of very low safety significance because it did not represent an actual loss of a safety function.

(Section 1R21.2b.3)

Licensee-Identified Violations

None.

REPORT DETAILS

1R21 Safety System Design and Performance Capability

Introduction Inspection of safety system design and performance verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected systems to perform design bases functions. As plants age, the design bases may be lost and important design features may be altered or disabled. The plants risk assessment model was based on the capability of the as-built safety system to perform the intended safety functions successfully. This inspectable area verifies aspects of the mitigating systems cornerstone for which there are no indicators to measure performance.

The objective of the safety system design and performance capability inspection was to assess the adequacy of calculations, analyses, other engineering documents, and operational and testing practices that were used to support the performance of the selected systems during normal, abnormal, and accident conditions.

The systems and components selected were the residual heat removal (RHR) system, safety injection (SI) system and selected portions of the chemical and volume control system (CVCS). These systems were selected for review based upon:

• having a high probabilistic risk analysis ranking;

• having had recent significant issues;

• not having received recent NRC review; and

• being interacting systems.

The criteria used to determine the acceptability of the systems performance was found in documents such as:

• applicable technical specifications;

• applicable updated safety analysis report (USAR) sections; and

• the systems' design documents.

The following system and component attributes were reviewed in detail:

System Requirements Process Medium - water, air, electrical signal Energy Source - electrical power, steam, air Control Systems - initiation, control, and shutdown actions Operator Actions - initiation, monitoring, control, and shutdown Heat Removal - cooling water and ventilation System Condition and Capability Installed Configuration - elevation and flow path operation Operation - system alignments and operator actions Design - calculations and procedures Testing - level, flow rate, pressure, temperature, voltage, and current Component Level Equipment/Environmental Qualification - temperature and radiation Equipment Protection - fire, flood, missile, high energy line breaks (HELBs), freezing, heating, ventilation and air conditioning

.1 System Requirements

a. Inspection Scope

The inspectors reviewed the USAR, technical specifications, system descriptions, drawings and available design basis information to determine the performance requirements of the RHR system, SI system and selected portions of the CVCS. The reviewed system attributes included process medium, energy sources, control systems, operator actions and heat removal. The rationale for reviewing each of the attributes was:

Process Medium: This attribute required review to ensure that the selected systems flow paths would be available and unimpeded during/following design basis events. To achieve this function, the inspectors verified that the systems would be aligned and maintained in an operable condition as described in the plants USAR, technical specifications and design bases.

Energy Sources: This attribute required review to ensure that the selected systems motive/electrical source would be available/adequate and unimpeded during/following design basis events, that appropriate valves and system control functions would have sufficient power to change state when required. To achieve this function, the inspectors verified that the interactions between the systems and their support systems were appropriate such that all components would operate properly when required.

Controls: This attribute required review to ensure that the automatic controls for operating the systems and associated systems were properly established and maintained. Additionally, review of alarms and indicators was necessary to ensure that operator actions would be accomplished in accordance with design requirements.

Operations: This attribute was reviewed because the operators perform a number of actions during normal, abnormal and emergency operating conditions that have the potential to affect the selected systems operation. In addition, the emergency operating procedures (EOPs) require the operators to manually realign the systems flow paths during and following design basis events. Therefore, operator actions play an important role in the ability of the selected systems to achieve their safety related functions.

Heat Removal: This attribute was reviewed to ensure that there was adequate and sufficient heat removal capability for the selected systems.

b. Findings

No findings of significance were identified.

.2 System Condition and Capability

a. Inspection Scope

The inspectors reviewed design basis documents and plant drawings, abnormal and emergency operating procedures, requirements, and commitments identified in the USAR and technical specifications. The inspectors compared the information in these documents to applicable electrical, instrumentation and control, and mechanical calculations, setpoint changes and plant modifications. The inspectors also reviewed operational procedures to verify that instructions to operators were consistent with design assumptions.

The inspectors reviewed information to verify that the actual system condition and tested capability was consistent with the identified design bases. Specifically, the inspectors reviewed the installed configuration, the system operation, the detailed design, and the system testing, as described below.

Installed Configuration: The inspectors confirmed that the installed configuration of the RHR system, SI system and selected portions of the CVCS met the design basis by performing detailed system walkdowns. The walkdowns focused on the installation and configuration of piping, components, and instruments; the placement of protective barriers and systems; the susceptibility to flooding, fire, or other environmental concerns; physical separation; provisions for seismic and other pressure transient concerns; and the conformance of the currently installed configuration of the systems with the design and licensing bases.

Operation: The inspectors performed procedure walk-throughs of selected manual operator actions to confirm that the operators had the knowledge and tools necessary to accomplish actions credited in the design basis.

Design: The inspectors reviewed the mechanical, electrical and instrumentation design of the RHR system, SI system and selected portions of the CVCS to verify that the systems and subsystems would function as required under accident conditions. The review included a review of the design basis, design changes, design assumptions, calculations, boundary conditions, and models as well as a review of selected modification packages. Instrumentation was reviewed to verify appropriateness of applications and set-points based on the required equipment function. Additionally, the inspectors performed limited analyses in several areas to verify the appropriateness of the design values.

Testing: The inspectors reviewed records of selected periodic testing and calibration procedures and results to verify that the design requirements of calculations, drawings, and procedures were incorporated in the system and were adequately demonstrated by test results. Test results were also reviewed to ensure automatic initiations occurred within required times and that testing was consistent with design basis information.

.1 RHR Discharge Overpressure Interlock

Introduction:

The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that, the design bases for the Units 1 and 2 RHR discharge overpressure interlock removal modification was not correctly translated into specifications, procedures, and instructions. Specifically, the modifications safety evaluation took credit for local operator action to manually open the RHR heat exchanger to SI pump suction valves during the transfer to recirculation in both units EOPs. However, on March 14, 2003, local operator action to manually open the valves was removed from the EOPs.

Description:

The inspectors reviewed a corrective action program (CAP) document (GEN20001718), which was written on June 2, 2000, that identified a lack of isolation between safety related and non-safety related instrumentation associated with the RHR discharge overpressure interlock. The purpose of this interlock was to prevent opening the RHR heat exchanger to SI pump suction valves (i.e., MV-32206, 32207, 32208 and 32209) from the control room, if RHR discharge pressure exceeded the allowable pressure for the SI pump suction piping. The equipment was determined to be operable but degraded with corrective action projected for completion in 2001.

The licensee had initiated a modification, 01RH01, RHR Disch Press Loop 1E/Non-1E Separation, to address this non-conformance. The safety evaluation prepared for this modification established that the design basis for this loop was non-safety related with credit taken for local operator action to manually open the valve. Having taken credit for local operator action to manually open the valve, the RHR system was considered operable by the licensee.

On March 14, 2003, the inspectors noted that local operator action to manually open these valves had been removed from the EOPs 1ES-1.2, Unit 1 Transfer to Recirculation, and 1ES-1.3, Unit 1 Transfer to Recirculation With One Safeguard Train Out of Service. As a result, the licensee initiated CAP029269 and revised the EOPs to open the valves locally if they did not open from the control room. Additionally, the licensee initiated CAP029598 to track removal of the reliance on local operator manual action.

Analysis:

Evaluation of this issue concluded that it was a design control deficiency resulting in a finding of very low safety significance (Green). The design control deficiency was due to the licensee removing steps from the EOPs that implemented a design basis requirement. The mitigating systems cornerstone was affected since the unqualified interlock could prevent the emergency core cooling system (ECCS) from performing a safety related function. No other cornerstones were determined to be degraded as a result of this issue.

This finding was greater than minor because the lack of coordination between the modifications design requirements and EOP procedural guidance affected the mitigating systems cornerstone objective. The cornerstones objective of ensuring the availability, reliability, and capability of the ECCS to respond to initiating events was affected.

The issue was assessed through Phase I of the significance determination process.

The inspectors agreed with the licensees position that, notwithstanding the reliance on procedural guidance and the lack of coordination with design requirements, the system would perform its safety function. Therefore, the inspectors concluded that the finding was a design deficiency that did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, procedures, and instructions.

Contrary to the above, as of March 14, 2003, the design basis for the Units 1 and 2 RHR discharge overpressure interlock removal modification was not correctly translated into specifications, procedures, and instructions. Specifically, the modifications safety evaluation took credit for local operator action to manually open the RHR heat exchanger to SI pump suction valves during the transfer to recirculation in both units EOPs. However, on March 14, 2003, local operator action to manually open the valves was removed from the EOPs. Because failure to correctly translate/maintain the RHR discharge overpressure interlock removal modifications design basis was of very low safety significance and has been entered into the corrective action program (CAP029269 and CAP029598), this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 50-282, 306/03-03-01(DRS), Failure to Correctly Translate/Maintain the RHR Discharge Overpressure Interlock Removal Modifications Design Basis.

.2 RHR Heat Exchanger Flow Control Valves

Introduction:

The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that, the design bases for the RHR system was not correctly maintained in accordance with regulatory requirements.

Specifically, a safety evaluation was written for the change in classification from safety related to non-safety related for the Units 1 and 2 RHR heat exchanger flow control valves positioners, hand controllers and signal converters. However, the safety evaluation failed to consider all credible failures in evaluating the single failure criterion.

For example, if a required open valves hand controller were to fail high the valve would close and block the ECCS flow path.

Description:

The inspectors reviewed safety evaluation 311 that was written to justify the change in classification of the valve positioners, hand controllers and signal converters for the RHR heat exchanger flow control valves (i.e., CV-31235, CV-31236, CV-31238, and CV-31239) from safety related to non-safety related. The valves were in the ECCS flow path and were required to remain open during a design basis event. The safety evaluation failed to consider all credible failures in evaluating the effect on compliance with the single failure criterion.

The licensee committed to following IEEE Standard 279-1971, Criteria for Protection System for Nuclear Power Generating Stations, in meeting the single failure criterion.

IEEE Standard 279-1971 defines credible failures of non-safety related components to include application of the maximum credible direct current

(dc) potential. If a hand controller were to fail high, producing a 50ma dc signal, the valve would close and block the ECCS flow path. As a result, the licensee initiated CAP029616 to correct the noncompliance.

Analysis:

Evaluation of this issue concluded that it was a design control deficiency resulting in a finding of very low safety significance (Green). The design control deficiency was due to the licensee changing the components safety related classification to non-safety related, which eliminated the requirement for the licensee to meet the single failure criterion, that placed the plant in noncompliance with regulatory requirements. The mitigating systems cornerstone was affected since an unqualified valve control loop could prevent the ECCS from performing a safety related function.

No other cornerstones were determined to be degraded as a result of this finding.

This finding was greater than minor because the change in classification from safety related to non-safety related for the Units 1 and 2 RHR heat exchanger flow control valve components affected the mitigating systems cornerstone objective. The cornerstones objective of ensuring the availability, reliability, and capability of the ECCS to respond to initiating events was affected.

The finding was assessed through Phase I of the significance determination process.

The inspectors agreed with the licensee's position that there was reasonable assurance that the system would perform its safety function. Therefore, the inspectors concluded that the finding was a design deficiency that did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, as of March 19, 1992, the licensee incorrectly changed the classification of safety related components to non-safety related which created a conflict between the regulatory requirements and the design basis. Specifically, a safety evaluation was written for the change in classification from safety related to non-safety related for the Units 1 and 2 RHR heat exchanger flow control valves positioners, hand controllers and signal converters. However, the safety evaluation failed to consider all credible failures in evaluating the single failure criterion. For example, if a required open valves hand controller were to fail high the valve would close and block the ECCS flow path. Because failure to consider all credible failures during the change in classification of the RHR heat exchanger outlet control valve components was of very low safety significance and has been entered into the CAP (CAP029616), this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 50-282, 306/03-03-02(DRS), Failure to Consider All Credible Failures During the Change in Classification of the RHR Heat Exchanger Outlet Control Valve Components.

.3 RHR Pit Covers

Introduction:

The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, due to the licensees failure to maintain the design basis configuration of the RHR pit covers. Specifically, the Units 1 and 2 auxiliary buildings RHR pit covers were designed to be closed during plant operation to limit the radiological dose rates to vital plant areas during accident conditions. However, prior to April 4, 2003, the Units 1 and 2 RHR pit covers were maintained in an open position during plant operation.

Description:

During the inspection walkdown, the inspectors observed that the RHR system pumps and associated heat exchangers/valves/piping were contained in RHR pits in the auxiliary building. The entrances to the two RHR pits were initially designed with heavy steel covers placed over the RHR pits. The RHR pit covers were subsequently modified with rollers allowing each of them to be opened/closed locally by an operator. The modification installed a control switch adjacent to each RHR pit, which controlled an electric motor that was used to open/close the RHR pit cover. During normal plant operation the RHR pit covers were maintained in the open position to allow for temporary ventilation into the RHR pits, thereby, permitting the licensee normal ingress/egress without having to establish the RHR pits as a confined space entry. In addition, the inspectors noted that EOPs 1ES-1.2 (2ES-1.2), Transfer to Recirculation, Revision 16, included a step for local operator action to close the RHR pit covers during accident conditions (Attachment K, Step 5).

The inspectors asked if the RHR pit covers were provided with safety related electrical power to ensure their capability to be closed after an accident. The licensee stated that they were supplied with a non-safety related electrical power supply and could not be closed by the operator if the power supply was lost.

The inspectors reviewed applicable portions of NUREG-0737, Item II.B.2, Prairie Island Shielding Study, dated January 1981. The objective of this study was to identify if vital plant areas, requiring personnel occupancy under post-accident conditions, could have high dose rates due to systems containing highly radioactive fluids. The study identified corrective actions required to limit the dose to an operator to 5 Rem whole body, or equivalent. The ECCS alignment area (i.e., the area around the RHR pits) was identified as a vital area requiring infrequent access. Post-accident dose rates from the RHR, SI, and containment spray systems were calculated based on the assumption that 100 percent of the core equilibrium Noble Gas inventory and 50 percent of the core equilibrium radioactive Halogen inventory had been diluted into the combined volume of the reactor coolant system and the refueling water storage tank. This radioactive fluid could be contained in the RHR and SI systems during the ECCSs recirculation mode of operation. One of the corrective actions identified by the study stated, RHR pit covers to be redesigned so that they may be left in place except when maintenance or inspection is going on in the pits.

The inspectors also reviewed Design Change No. 80Y103, RHR Pit Cover Access Fix.

This design change added the rollers and electric motors to the RHR pit covers. Based on the description included in the modification package, the RHR pit covers were intended to remain closed, except when access to the pits was required. In addition, the modification package indicated that the motor assembly was purchased as non-safety related.

At the time of the inspection the licensee had not determined how long the RHR pit covers had been left open. However, the licensee stated that a step to close the RHR pit covers during a design basis accident was first included in a previous EOP, dated July 8, 1982.

On April 3, 2003, in response to the inspectors concerns, the licensee initiated CAP029501, RHR Pit Covers Powered from Non-Safety Related Power Supplies, dated April 3, 2003. The CAP stated that the dose received by an operator performing EOP local operator actions could exceed the assumed values and that documentation could not be found to support the RHR pit covers being open. The CAP recommended that the RHR pit covers be maintained in a closed position. On April 4, 2003, to ensure operability, the licensee closed the RHR pit covers and tagged-out the power supply to assure the covers remain closed. The licensee also issued form PINGP 1224, Crew Meeting Review of Noteworthy Event/Near Miss/Change - RHR Pit Covers to Remain Closed, to inform operating personnel of the issue.

The licensee had not completed their review of this condition for past operability and potential reportability during the inspection period. The licensee stated that this review would address the potential impact of the open RHR pit covers on post-accident access to vital areas, as well as the potential impact on the environmental qualification of electrical equipment in the area. In addition, the licensee stated that they would determine the appropriate controls associated with opening the RHR pit covers as required during plant operation. These activities would be tracked by CAP029501.

Analysis:

Evaluation of this issue concluded that it was a licensee performance deficiency resulting in a finding of very low safety significance (Green). The performance deficiency was due to the licensees failure to maintain the design basis configuration of the RHR pit covers. The mitigating systems cornerstone was affected due to the potential of long term heat removal being degraded by this condition.

This finding was greater than minor because the potential to affect the SI and RHR systems design basis functions (i.e., degradation of long term heat removal) affected the mitigating systems cornerstone objective. Specifically, local operator actions in the auxiliary building (e.g., area around the RHR pits) were required to transfer the ECCS to the recirculation mode. The local operator actions were included in both of the units EOPs 1ES-1.2 and 2ES-1.2, Transfer to Recirculation, Attachment K. The required local operator actions included closing the breakers for the RHR to SI pump suction valves (i.e., MV-32206 & MV-32207 and MV-32208 & MV-32209). These valves were required to be repositioned to establish high head safety injection recirculation flow. If the operator was prevented from performing the local operator actions during accident conditions due to high dose rates, then both trains of ECCS could be degraded. As a result, the cornerstones objective of ensuring the availability, reliability, and capability of the ECCS to respond to initiating events was affected.

The finding was assessed through Phase I of the significance determination process.

The inspectors agreed with the licensees position that with the RHR pit covers in the closed position that the system would perform its safety function. The specific accident conditions that could have challenged these systems have not existed and the systems have not been operated under these operating modes. Therefore, the inspectors concluded that the finding was a performance deficiency that did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, prior to April 4, 2003, the design basis of the Units 1 and 2 auxiliary buildings RHR pit covers was not correctly maintained, in that, the position of the RHR pit covers was not effectively controlled. Although the design basis for personnel access to vital areas during accident conditions was based on the RHR pit covers being closed during plant operation, the covers were maintained in an open position prior to April 4, 2003. As a result, the potential existed for safety system operability concerns during post-accident conditions. The licensee implemented appropriate corrective actions to address this finding. Because failure to maintain the RHR pit covers design basis configuration was of very low safety significance and has been entered into the CAP (CAP029501), this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 50-282, 306/03-03-03(DRS), Failure to Maintain the RHR Pit Covers Design Basis Configuration.

.3 Components

a. Inspection Scope

The inspectors examined the RHR, SI and selected portions of the CVCS systems associated pumps, heat exchangers and instrumentation to ensure that component level attributes were satisfied.

Equipment/Environmental Qualification: This attribute verifies that the equipment was qualified to operate under the environment in which it was expected to be subjected to under normal and accident conditions. The inspectors reviewed design information, specifications, and documentation to ensure that the RHR system, SI system and selected portions of the CVCS were qualified to operate within the temperatures and radiation fields specified in the environmental qualification documentation.

Equipment Protection: This attribute verifies that the RHR system, SI system and selected portions of CVCS were adequately protected from natural phenomenon and other hazards, such as HELBs, floods or missiles. The inspectors reviewed design information, specifications, and documentation to ensure that the systems were adequately protected from those hazards identified in the USAR, which could impact the systems ability to perform their safety function.

b. Findings

No findings of significance were identified.

4OA2 Problem Identification and Resolution (PI&R)

.1 Review of Condition Reports

a. Inspection Scope

The inspectors reviewed a sample of problems associated with the RHR system, SI system and selected portions of the CVCS that were identified and entered into the CAP by the licensee. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions related to design issues. In addition, condition reports written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

Exit Meeting On April 11, 2003, the inspectors presented the inspection results to Mr. J. Solymossy and other members of his staff. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. Proprietary information was reviewed during the inspection, as documented in the list of documents. The inspectors confirmed that the proprietary material had been returned and discussed the likely content of the inspection report. The licensee did not indicate any potential conflicts with information presented.

A1 ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

J. Solymossy, Site Vice President

S. Northard, Director - Engineering

M. Wadley, Senior Vice President Operations Support

S. Cook, Manager NOS

B. Alexander, Corporate Engineering

J. Kivi, Senior Regulatory Compliance Engineer

G. Eckholt, Regulatory Affairs Manager

E. Weinkam, Director Regulatory Services

T. Verbout, I&C/Electrical Design Supervisor

D. Anderson, Response Team Technical Leader

S. Thomas, Design Engineering

B. Rogers, Design Engineering

R. Pond, Design Engineering

T. Lillehei, Design Engineering

B. Peterson, Engineering

L. Johnson, System Engineering

G. Thoraldson, System Engineering

D. Molback, System Engineering

D. Price, System Engineering

J. Kapitz, System Engineering

R. Wirkkala, System Engineering

D. Smith, Shift Manager Operations

R. Williston, Programs Engineering

C. Mundt, Planning Manager

A. Johnson, Rad. Protection and Chemistry Manager

Nuclear Regulatory Commission

J. Adams, Senior Resident Inspector

D. Karjala, Resident Inspector

C. Pederson, Director, Division of Reactor Safety

A2

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

NONE

Opened and Closed

50-282, 306/03-03-01(DRS)

NCV Failure to Correctly Translate/Maintain the RHR Discharge Overpressure Interlock Removal Modifications Design Basis (Section 1R21.2b.1)

50-282, 306/03-03-02(DRS)

NCV Failure to Consider All Credible Failures During the Change in Classification of the RHR Heat Exchanger Outlet Control Valve Components (Section 1R21.2b.2)

50-282, 306/03-03-03(DRS)

NCV Failure to Maintain the RHR Pit Covers Design Basis Configuration (Section 1R21.2b.3)

Closed

NONE

Discussed

NONE

A3