IR 05000282/2005002

IR 05000282-05-002 (Drs); IR 05000306-05-002 (Drs); on 06/13/2005 - 07/01/2005; for Prairie Island Nuclear Generating Plant, Units 1 and 2; Safety System Design and Performance Capability Inspection

ML052270068
Person / Time
Site:	Prairie Island
Issue date:	08/12/2005
From:	Ann Marie Stone Division of Reactor Safety III
To:	Thomas J. Palmisano Nuclear Management Co
References
IR-05-002
Download: ML052270068 (40)

Text

ust 12, 2005

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY INSPECTION REPORT 05000282/2005002(DRS); 05000306/2005002(DRS)

Dear Mr. Palmisano:

On July 1, 2005, the U. S. Nuclear Regulatory Commission (NRC) completed a safety system design and performance capability inspection at your Prairie Island Nuclear Generating Plant, Units 1 and 2. The enclosed report documents the inspection findings which were discussed on July 1, 2005, with Mr. L. Clewett and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and to 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 auxiliary feedwater system and support systems to ensure that they were capable of performing their required safety related functions.

Based on the results of this inspection, six NRC-identified findings of very low safety significance, all of which involved violations of NRC requirements were identified. However, because these violations were of very low safety significance and because the findings were entered into the licensees corrective action program, the NRC is treating these findings as Non-Cited Violations in accordance with Section VI.A.1 of the NRCs Enforcement Policy.

Additionally, a licensee identified violation is listed in Section 4OA7 of this report.

If you contest the subject or severity of a Non-Cited Violation, 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 a copy to the Regional Administrator, U. S. Nuclear Regulatory Commission -

Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Prairie Island Nuclear Generating Plant facility.

In accordance with 10 CFR 2.390 of the NRC's 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 Publically 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/

Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-282; 50-306 License Nos. DPR-42; DPR-60 Enclosure: Inspection Report 05000282/2005002(DRS); 05000306/2005002(DRS)

w/Attachment: Supplemental Information cc w/encl: C. Anderson, Senior Vice President, Group Operations J. Cowan, Executive Vice President and Chief Nuclear Officer Regulatory Affairs Manager J. Rogoff, Vice President, Counsel & Secretary Nuclear Asset Manager Tribal Council, Prairie Island Indian Community Administrator, Goodhue County Courthouse Commissioner, Minnesota Department of Commerce Manager, Environmental Protection Division Office of the Attorney General of Minnesota

SUMMARY OF FINDINGS

IR 05000282/2005002(DRS); 05000306/2005002(DRS); 06/13/2005 - 07/01/2005; Prairie

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

This report covers a three-week period of announced baseline inspection on the design and performance capability of the auxiliary feedwater system and support systems. The inspection was conducted by Region III inspectors and a mechanical engineering consultant. Six Green findings associated with six Non-Cited Violations 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.

A. Inspector-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, requirements.

The licensee failed to recognize an increased pressure drop in the hydraulic characteristics between the new replacement steam generators (RSGs) and associated main steam safety valves. Specifically, Calculation ENG-ME-454, Pressure Drop Between SG [steam generator] and Safety Valve, Revision 0, was not updated (i.e., revised) to evaluate the affects of the increased pressure drop associated with the RSGs. Once identified, the licensee entered the finding into their corrective action program (CAP) as CAP043077 to revise the affected calculations.

The finding was more than minor because the failure to evaluate a change in pressure drop through the RSGs could have caused an adverse effect on the auxiliary feedwater (AFW) pumps flow delivery to the RSGs and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate design margin remained for the increased pressure drop on the AFW system and did not represent an actual loss of a safety function. (Section 1R21.1b.1)

Green.

A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, requirements.

The licensee failed to select an appropriate method for calculating the onset of vortexing at the intake of the AFW suction lines from the condensate storage tank (CST).

Specifically, Calculation ENG-ME-293, Safety Related Tank Usable Volume Evaluation, Revision 3, used a method to determine the minimum height of water above the auxiliary feedwater (AFW) pumps intake to preclude vortex formation that was not appropriate. Once identified, the licensee entered the finding into their corrective action program (CAP) as CAP043276 to revise the affected calculations.

The finding was more than minor because the failure to prevent the formation of vortexing at the intake of the AFW suction lines would result in air entrapment causing pulsating pump flow and/or reduction in pump performance and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate CST capacity remained for the AFW system and did not represent an actual loss of a safety function.

(Section 1R21.1b.2)

Green.

A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, requirements.

The licensee failed to correctly specify the minimum pump operability limits to be used in auxiliary feedwater (AFW) surveillance testing. Specifically, Calculation ENG-ME-576,

AFW Pump Minimum Acceptance Criteria - Proto Power Calculation 96-076,

Revision B, Revision 0, did not include the bypass cooling flow to the turbine driven auxiliary feedwater pump (TDAFWP) turbine bearings and governor nor include the potential variability in the speed of the TDAFWP. This resulted in an AFW system hydraulic calculation that was non-conservative when determining the minimum acceptance criteria for the TDAFWP full flow test. Once identified, the licensee verified operability and entered the finding into their corrective action program (CAP) as CAP043273 to revise the tests acceptance criteria.

The finding was more than minor because the failure to account for bypass cooling flow and pump speed variation in the surveillance test acceptance criteria would result in over-predicting the AFW pumps performance (i.e., creating design margin capability that would not exist) and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate design margin existed for the AFW system and did not represent an actual loss of a safety function. (Section 1R21.2b.1)

Green.

A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, requirements.

The licensee failed to include the affects of increased initial room temperature and heat load addition due to turbine driven auxiliary feedwater pump (TDAFWP) steam leaks when evaluating the auxiliary feedwater (AFW) pump rooms temperature on a loss of ventilation. Specifically, Calculation ENG-ME-182, AFW Pump Room Ventilation System Design, Revision 0, assumed an initial nominal AFW pump room temperature that was not consistent with actual environmental conditions which resulted in a non-conservative heat-up transient design analysis. Once identified, the licensee entered the finding into their corrective action program (CAP) as CAP043301 to revise the affected calculations.

The finding was more than minor because the failure to account for a higher initial room temperature and the potential steam leaks would result in a higher room temperature on a loss of ventilation causing equipment degradation due to the higher than anticipated ambient temperature and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees heat-up transient design analysis showed that adequate design margin remained for the increased temperature on the AFW system and did not represent an actual loss of a safety function. (Section 1R21.2b.2)

Green.

A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, requirements.

The licensee failed to recognize that the calculated design value for cooling water inlet temperature was higher than that assumed by the auxiliary feedwater (AFW) pumps lube oil cooler thermal performance analysis. Specifically, Calculation MECH-0268.4,

Verification of Heat Removal Capability of the American Standard Heat Exchanger,

Model 02030-EF, Revision 0, used an assumed value for cooling water inlet temperature that did not include the AFW pumps heat energy transferred to the cooling water when calculating the lube oil coolers operating temperature. This resulted in the lube oil coolers thermal performance analysis being non-conservative. Once identified, the licensee entered the finding into their corrective action program (CAP) as CAP043239 to revise the affected calculations.

The finding was more than minor because the failure to account for the AFW pumps heat energy transferred to the cooling water would result in a higher lube oil cooler operating temperature causing increased turbine bearing and governor degradation and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate design margin remained for the AFW system and did not represent an actual loss of a safety function. (Section 1R21.3b.1)

Green.

A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, requirements.

The licensee failed to maintain the auxiliary feedwater (AFW) instrumentation tubing suction lines in a water solid condition to pressure switch 17704. The pressure switch performed a safety related function to sense low suction pressure and trip the 11 turbine driven auxiliary feedwater pump (TDAFWP) upon a low level condition in the condensate storage tank (CST). Specifically, a void was discovered in the safety related instrumentation tubing which lowered the effective setpoint for the 11 TDAFW pumps low suction pressure trip. Once identified, the licensee entered the finding into their corrective action program (CAP) as CAP043298 to take corrective actions.

The finding was more than minor because the failure to prevent the formation of a void in the TDAFW pumps instrumentation tubing suction lines would result in air entrapment causing erroneous pressure switch performance and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate design margin remained for the trip setpoint on the AFW system and did not represent an actual loss of a safety function.

(Section 1R21.3b.2)

Licensee-Identified Violations

A violation of very low safety significance, which was identified by the licensee has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and the licensees corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 and Unit 2 operated at or near full power throughout the inspection period.

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 system and components selected were from the auxiliary feedwater (AFW) system.

This system was 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; and 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; and Testing - level, flow rate, pressure, temperature, voltage, and current.

Component Level Component Degradation potential degradation monitored or prevented and component replacement consistent with inservice/equipment qualified life; Equipment/Environmental Qualification temperature, humidity, radiation, pressure, voltage and vibration; Equipment Protection fire, flood, missile, high energy line breaks (HELBs), freezing, heating, ventilation and air conditioning; and Component Inputs/Outputs component inputs/outputs are suitable for application (e.g., inputs/outputs for proper component operation are provided and valves fail in safe configuration).

.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 AFW system. 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.

.1 Hydraulic Analysis Not Updated for RSGs (Replacement Steam Generators)

Introduction:

The inspectors identified a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green)involving the AFW systems hydraulic design analysis. Specifically, the inspectors determined that the licensee failed to recognize an increased pressure drop in the hydraulic characteristics between the new RSGs and associated main steam SVs. The new RSGs were installed during the 1R23 Refueling Outage in the Fall of 2004.

Description:

The inspectors reviewed Calculation ENG-ME-454, Pressure Drop Between SG [steam generator] and Safety Valve [SV], Revision 0. The purpose of the calculation was to evaluate the pressure drop between the SGs and the associated main steam SVs to determine the effect of hydraulic resistance on the AFW systems flow delivery to the SGs, since flow delivery is affected by pressure drop. This calculation was used as a design input to Calculation ENG-ME-576, AFW Pump Minimum Acceptance Criteria-Proto Power Calculation 96-076, Revision B, Revision 0.

The inspectors noted that Calculation ENG-ME-454 had not been updated (i.e., revised)to evaluate the affects on pressure drop due to the new RSGs. In response, the licensee performed an analysis which indicated that with the new SGs, the pressure drop between the SG and the SV was higher for the new RSGs when compared with the old SGs. This increased pressure drop required additional evaluation by the licensee to ensure that adequate design margin existed, such that the increased pressure drop did not have an adverse affect on the AFW system.

The licensee subsequently evaluated the effect of the increased pressure drop on the AFW systems operability and concluded that Calculation ENG-ME-576, which used the ENG-ME-454 calculation results as design input, was conservative. As a result, the inspectors review concluded that there was no affect on the AFW systems operability since adequate design margin existed with the back-pressure value used in Calculation ENG-ME-576.

Analysis:

The inspectors determined that failure to recognize that an increased pressure drop in the hydraulic characteristics between the new RSGs and associated main steam SVs was a performance deficiency warranting a significance evaluation.

The inspectors concluded that the finding was greater than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002. The finding involved the attribute of design control, where failure to evaluate a change in pressure drop through the RSGs could have caused an adverse effect on the AFW pumps flow delivery to the RSGs, and could have affected the mitigating systems objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. The inspectors agreed with the licensee's position that, despite the loss of design margin in the AFW pumps flow delivery to the RSGs, the AFW system would have performed its safety function.

Therefore, the inspectors concluded that the finding did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design changes shall be subject to design control measures commensurate with those applied to the original design and that design control measures shall provide for verifying or checking the adequacy of design.

Contrary to the above, as of July 1, 2005, the licensees design control measures failed to recognize and provide for verifying or checking the adequacy of design to account for an increased pressure drop in the hydraulic characteristics between the new RSGs and associated main steam SVs. Specifically, Calculation ENG-ME-454, Pressure Drop Between SG and Safety Valve, Revision 0, was not updated (i.e., revised) to evaluate the affects of the increased pressure drop associated with the RSGs. Once identified, the licensee entered the finding into their corrective action program (CAP) as CAP043077 to revise the affected calculations. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000282/2005002-01(DRS);05000306/2005002-01(DRS)).

.2 Vortex Analysis Methodology Not Appropriate

Introduction:

The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) involving the condensate storage tank (CST) volumes design analysis. Specifically, the inspectors identified that the licensee failed to select an appropriate method for calculating the onset of vortexing at the intake of the AFW suction lines from the CST.

Description:

The inspectors reviewed Calculation ENG-ME-293, Safety Related Tank Usable Volume Evaluation, Revision 3. The purpose of the calculation was to determine the usable volume for each of the plants specified safety related tanks, then compare the usable volume to the minimum tank volume identified in the plants Technical Specifications to ensure that plant procedures specified an adequate tank minimum volume.

The inspectors noted that the methodology used in Calculation ENG-ME-293 to determine the minimum height of water above the AFW pumps intake to preclude vortex formation was not appropriate. The calculations methodology did not account for the actual fluid configuration where air ingestion into the AFW pumps intake would potential occur. The onset of vortexing was calculated using a methodology developed by Harleman, which is based on selective fluid withdrawal from a stratified fluid consisting of an upper and lower liquid layer differing slightly in density and similar in viscosity (emphasis added). This methodology was described in a paper by Harleman, D. R. F., et. al, Selective Withdrawal From A Vertically Stratified Fluid, Intl. Association for Hydraulic Research, 8th Congress - Montreal, August 24, 1959. The term stratified fluid implies a variation in the density of the fluid in the vertical direction.

The inspectors asked the licensee to provide justification for using the Harleman method since the fluid in the CST (e.g., this configuration also applied to the refueling water storage tank (RWST)) was air over water and not a stratified fluid consisting of an upper and lower liquid layer differing slightly in density and similar in viscosity. The licensee was unable to provide adequate technical justification for the methodology used and stated they would consider other methods applicable to this configuration that were more readily accepted by the industry.

The inspectors independently calculated (i.e., using the analysis methodology recommended by the Hydraulics Institute) that the onset of AFW pump inlet vortexing would occur at almost twice the height determined by the Harleman method. The licensee performed a similar calculation using an alternate method and reached the same conclusion - that the usable CST tank capacity was correspondingly reduced by approximately 2500 gallons per tank. Although the usable CST tank capacity was reduced, the inspectors concluded that there was adequate CST capacity and that no safety concern existed for the AFW system.

Although not reviewed by the inspectors, the licensee re-evaluated the potential for vortexing in the RWST by using a more appropriate analysis method and determined that switch-over of residual heat removal pump suction to the sump would occur prior to the level where vortexing in the tank would be a concern.

Analysis:

The inspectors determined that failure to select an appropriate method for calculating the onset of vortexing at the intake of the AFW suction lines from the CST was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002. The finding involved the attribute of design control, where failure to prevent the formation of vortexing at the intake of the AFW suction lines would result in air entrapment causing pulsating pump flow and/or reduction in pump performance, and could have affected the mitigating systems objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. The inspectors agreed with the licensee's position that, despite the loss of design margin in available CST volume, the AFW system would have performed its safety function. Therefore, the inspectors concluded that the finding did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established for the selection and review for suitability of application of processes that are essential to the safety-related functions of the structures, systems and components.

Contrary to the above, as of July 1, 2005, the licensee failed to select and review for suitability an appropriate method for calculating the onset of vortexing at the intake of the AFW suction lines from the CST. Specifically, Calculation ENG-ME-293, Safety Related Tank Usable Volume Evaluation, Revision 3, used a method to determine the minimum height of water above the AFW pumps intake to preclude vortex formation that was not appropriate. The calculations methodology did not account for the actual fluid configuration where air ingestion into the AFW pumps intake would potential occur.

Once identified, the licensee entered the finding into their corrective action program as CAP043276 to revise the affected calculations. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000282/2005002-02(DRS);05000306/2005002-02(DRS)).

.2 System Condition and Capability

a. Inspection Scope

The inspectors reviewed design basis documents and plant drawings, abnormal and EOP, 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 AFW system 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 AFW system 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 Non-Conservative Acceptance Criteria

Introduction:

The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) involving the AFW systems hydraulic design analysis. Specifically, the inspectors identified that the licensee failed to correctly specify the minimum pump operability limits to be used in AFW surveillance testing.

Description:

The inspectors reviewed Calculation ENG-ME-576, AFW Pump Minimum Acceptance Criteria - Proto Power Calculation 96-076, Revision B, Revision 0. The purpose of the calculation was to develop AFW pump curves to be used in IST procedures when testing the AFW pump. The inspectors identified that the hydraulic analysis, which established the minimum acceptance criteria for the AFW pump, did not include the effect of the flow diversion due to the bypass flow to the turbine driven auxiliary feedwater pumps (TDAFWP) turbine bearings and governor cooling lines. In addition, the analysis did not include the affect on the pump curve due to potential variability in the speed of the turbine. A change in turbine speed would result in a different pump curve. These issues did not affect the motor driven auxiliary feedwater pumps (MDAFWP). However, by not accounting for the bypass flow, the calculation assumed more flow would be delivered to the SGs. Secondly, without correcting for the allowable minimum turbine speed, the calculation was non-conservative when calculating the allowable degradation of the pump curve.

The affect of not evaluating these issues in the calculation was addressed by the licensee to determine the effect on the pump acceptance criteria in the systems test procedures. The licensee determined that the acceptance criteria for the minimum flow tests were still appropriate. However, the acceptance criteria for the full flow test was non-conservative. The most recent pump tests were reviewed by the inspectors. The inspectors determined that adequate design margin remained between the higher minimum test points and current operating points. As a result, the inspectors concluded the AFW system was operable.

The licensee determined that Calculation ENG-ME-576 required revision to include the effects of unaccounted bypass flow and turbine speed variations. In addition, because the calculation determined the acceptance criteria for AFW pump surveillance testing, the procedures for AFW pump testing required revision as well.

Analysis:

The inspectors determined that failure to correctly specify the minimum pump operability limits to be used in AFW surveillance testing was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002. The finding involved the attribute of design control, where failure to account for bypass cooling flow and pump speed variation in the surveillance test acceptance criteria would result in over-predicting the AFW pumps performance (i.e., creating design margin capability that would not exist), and could have affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. The inspectors agreed with the licensee's position that, despite the loss of design margin in the AFW pump flow delivery to the SGs, the AFW system would have performed its safety function.

Therefore, the inspectors concluded that the finding did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, 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, as of July 1, 2005, the licensee failed to assure that the minimum pump operability limits to be used in AFW surveillance testing were correctly translated into specifications, drawings, procedures, and instructions. Specifically, Calculation ENG-ME-576, AFW Pump Minimum Acceptance Criteria - Proto Power Calculation 96-076, Revision B, Revision 0, did not include the bypass cooling flow to the TDAFW pumps turbine bearings and governor and did not include the potential variability in the speed of the TDAFW pump. Once identified, the licensee entered the finding into their corrective action program as CAP043273 to revise the affected documents. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000282/2005002-03(DRS);05000306/2005002-03(DRS)).

.2 AFW Room Heat-Up Analysis Deficiencies

Introduction:

The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) involving the AFW pump rooms heat-up transient design analysis. Specifically, the inspectors identified that the licensee failed to include the affects of increased initial room temperature and heat load addition due to TDAFW pump steam leaks when evaluating the AFW pump rooms temperature on a loss of ventilation.

Description:

The inspectors reviewed Calculation ENG-ME-182, AFW Pump Room Ventilation System Design, Revision 0, and supporting Calculation 194001-2.5-001, Unit Cooler Downgrade Study, Revision 0. The purpose of the calculations was to determine the temperature versus time characteristics of the AFW pump room on a loss of room cooling function, which was based on the transient temperature behavior of the room.

The inspectors noted that the calculations assumed the nominal room temperature in the AFW pump room area was 80 degrees Fahrenheit (EF) and that no steam leaks existed that might add heat to the room. On June 14, 2005, during the inspectors walkdown of the AFW pump room area, the inspectors noted that the room temperature was significantly higher than 80 EF. On June 16, 2005, during the 11 TDAFW pump testing, the inspectors observed a small steam leak below the turbines trip throttle valve. The licensee initiated CAP043301 to document the elevated room temperature and steam leak conditions. The inspectors concluded that since the AFW pump rooms heat-up transient design analysis did not consider the rooms higher initial temperature and the heat load addition due to the steam leaks, the heat-up transient design analysis was regarded as non-conservative.

The licensee evaluated the affects of not assuming a higher initial room temperature and the additional heat load due to steam leaks on the AFW pump rooms heat-up transient design analysis. A draft analysis was performed that showed the predicted room air temperatures would be less than those used for evaluation of the equipment in the AFW pump room as part of the Unit Cooler Downgrade Study. The licensees review of the completed draft analysis concluded that there was no impact on operability of the AFW pump. The inspectors concurred with this determination. The licensee stated that there was current action to reperform the AFW pump rooms heat-up transient design analysis and that the specific items discussed above would be considered.

Analysis:

The inspectors determined that failure to include the affects of increased initial room temperature and heat load addition due to TDAFW pump steam leaks when evaluating the AFW pump rooms temperature on a loss of ventilation was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002. The finding involved the attribute of design control, where failure to account for a higher initial room temperature and the potential steam leaks would result in a higher room temperature on a loss of ventilation causing equipment degradation due to the higher than anticipated ambient temperature, and could have affected the mitigating systems objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. The inspectors agreed with the licensee's position that, despite the loss of design margin in the AFW pump rooms heat-up transient design analysis, the AFW system would have performed its safety function. Therefore, the inspectors concluded that the finding did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, as of July 1, 2005, the licensee failed to provide design control measures for verifying or checking the adequacy of design to evaluate the initial design assumptions assumed in the AFW pump rooms heat-up transient design analysis.

Specifically, Calculation ENG-ME-182, AFW Pump Room Ventilation System Design, Revision 0, assumed an initial nominal AFW pump room temperature that was not consistent with actual environmental conditions which resulted in a non-conservative heat-up transient design analysis. Once identified, the licensee entered the finding into their corrective action program as CAP043301 to revise the affected calculations.

Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000282/2005002-04(DRS);05000306/2005002-04(DRS)).

.3 Components

a. Inspection Scope

The inspectors examined the AFW systems associated pumps, heat exchangers and instrumentation to ensure that component level attributes were satisfied.

Component Degradation: This attribute verifies that potential degradation was monitored or prevented and component replacement was consistent with inservice and/or equipment qualification life. The inspectors examined existing system programs to ensure that components were adequately maintained.

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 AFW system was qualified to operate within the environmental conditions specified in the environmental qualification documentation.

Equipment Protection: This attribute verifies that the AFW system was 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.

Component Inputs/Outputs: This attribute verifies that the components inputs and outputs were suitable for the application and would be acceptable under accident conditions. For example, the valve fails in a safe configuration and the required inputs to components, such as coolant flow, electrical voltage, and control air necessary for proper component operation were provided. The inspectors reviewed design information, specifications, and documentation and ensured that selected system components were provided inputs and/or outputs suitable for the application.

.1 Lube Oil Cooler Analysis Deficiencies

Introduction:

The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) involving the AFW pumps lube oil cooler thermal performance analysis. Specifically, the licensee failed to recognize that the calculated design value for cooling water inlet temperature was higher than that assumed by the AFW pumps lube oil cooler thermal performance analysis.

Description:

The inspectors reviewed Calculation MECH-0268.4, Verification of Heat Removal Capability of the American Standard Heat Exchanger, Model 02030-EF, Revision 0. The purpose of the calculation was to confirm the heat removal capability of the AFW pumps lube oil cooler. The cooling waters flow path was configured, such that prior to entering the heat exchanger the cooling water passed through the AFW pump. During AFW pump operation, due to inefficiencies of the pump, the pump transfers energy in the form of heat to the water passing through the pump. This heat energy transfer (i.e., pump heat energy transferred to the cooling water) raises the temperature of the cooling water several degrees before the cooling water enters the lube oil cooler.

The inspectors noted that in Calculation MECH-0268.4, the licensee failed to recognize that the assumed value for cooling water inlet temperature did not include the pumps heat energy transferred to the cooling water when calculating the AFW pumps lube oil coolers operating temperature. By not including the pumps heat energy transfer to the cooling water, the calculation was non-conservative by several degrees Fahrenheit when predicting the AFW pumps lube oil coolers operating temperature.

The licensee subsequently evaluated the effect of not including the pumps heat energy transfer in the heat exchangers thermal performance analysis. The licensee determined that there was no impact on operability of the AFW pumps lube oil coolers because the limit for lube oil temperature out of the pump bearing was 160 EF and preliminary calculations indicated that the predicted temperature out of the lube oil cooler would be 154 EF when accounting for the AFW pumps heat energy transfer. The inspectors reviewed the licensees evaluation and concluded that there would be no affect on operability of the lube oil coolers when accounting for the higher cooling water inlet temperature. The licensee determined that Calculation MECH-0268.4 needed to be revised and issued CAP043239, AFW Lube Oil Cooler Calculation, dated June 27, 2005, to revise the subject calculation.

Analysis:

The inspectors determined that failure to account for the pumps heat energy transfer when calculating cooling water inlet temperature to the AFW pumps lube oil coolers was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002. The finding involved the attribute of design control, where failure to account for the AFW pumps heat energy transferred to the cooling water would result in a higher lube oil cooler operating temperature causing increased turbine bearing and governor degradation, and could have affected the mitigating systems objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. The inspectors agreed with the licensee's position that, despite the loss of design margin in the AFW pumps lube oil cooler thermal performance analysis, the AFW system would have performed its safety function. Therefore, the inspectors concluded that the finding did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, as of July 1, 2005, the licensees design control measures failed to recognize and provide for verifying or checking the adequacy of design by validating that the calculated design value for cooling water inlet temperature was higher than that assumed by the AFW pumps lube oil cooler thermal performance analysis. Specifically, Calculation MECH-0268.4, Verification of Heat Removal Capability of the American Standard Heat Exchanger, Model 02030-EF, Revision 0, used an assumed value for cooling water inlet temperature that did not include the AFW pumps heat energy transferred to the cooling water when calculating the lube oil coolers operating temperature. This resulted in the lube oil coolers thermal performance analysis being non-conservative. Once identified, the licensee entered the finding into their corrective action program (CAP) as CAP043239 to revise the affected calculations. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000282/2005002-05(DRS);05000306/2005002-05(DRS)).

.2 Void in TDAFW Pump Instrumentation Line

Introduction:

The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) involving the AFW pumps suction pressure instrumentation. Specifically, the inspectors identified that the licensee failed to assure that the design bases requirement to maintain the AFW instrumentation tubing suction lines in a water solid condition was not correctly translated into specifications, drawings, procedures, and instructions.

Description:

On June 14, 2005, the inspectors conducted a walkdown of the AFW system. During the walkdown, the inspectors observed that some instrument tubing for the 11 TDAFW pump was installed with a large inverted U-shaped loop. The instrumentation tubing was attached to the suction pressure switch 17704, which performed a safety related function to sense low suction pressure and trip the TDAFW pump upon a low level condition in the CST. All AFW pumps were installed with a similar configuration, although not as pronounced as that on the 11 TDAFW pump.

When the inspectors asked the licensee how the instrumentation tubing lines were assured to be water solid, the licensee responded that there was no periodic procedure to vent these lines. Data from a surveillance (SP 1102) conducted on June 15, 2005, recorded local suction pressure 1.1-psi higher than expected from the recorded height of the water in the CST. The licensee walked down the systems on June 29, 2005, and local suction pressure indication was approximately 1.7 psi higher than expected from the height of the water in the CST. This data indicated that there was a void in the instrumentation tubing. On June 30, 2005, the licensee vented the instrumentation line and observed a change of 1.5 psi in the local indicated pressure. The change in pressure corresponded to an approximate 42-inch long void in the instrumentation tubing. The inspectors noted that this void would also expand as the pressure dropped in the system during operation. The licensee calculated that, if called upon, the pump would not have tripped until the CST level was approximately 55-inches lower than expected. Although the CST would have emptied, sufficient net positive suction head was available due to the large suction header piping. The licensee subsequently evaluated the effect of the setpoint bias on operability and concluded that the AFW system would have performed its safety function.

Analysis:

The inspectors determined that failure to maintain the AFW instrumentation tubing suction lines in a water solid condition to pressure switch 17704 was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002. The finding involved the attribute of design control, where failure to prevent the formation of a void in the TDAFW pumps instrumentation tubing suction lines would result in air entrapment causing erroneous pressure switch performance and could have affected the mitigating systems objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. The inspectors agreed with the licensee's position that, despite the significant loss of design margin in the trip setpoint, the AFW system would have performed its safety function. Therefore, the inspectors concluded that the finding did not represent an actual loss of a safety function and the finding screened out as having very low safety significance or Green.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, 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, as of July 1, 2005, the licensee failed to assure that the design bases requirement to maintain the AFW instrumentation tubing water solid was not correctly translated into specifications, drawings, procedures, and instructions.

Specifically, a void was discovered in the safety related instrumentation tubing which lowered the effective setpoint for the 11 TDAFW pumps low suction pressure trip.

Once identified, the licensee entered the finding into their corrective action program as CAP043298 to take corrective actions. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000282/2005002-06(DRS);05000306/2005002-06(DRS)).

4OA2 Identification and Resolution of Problems

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed a sample of problems associated with the AFW system 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

.1 Exit Meeting

The inspectors presented the inspection results to Mr. L. Clewett and other members of licensee management at the conclusion of the inspection on July 1, 2005. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

.2 Interim Exit Meetings

None.

4OA7 Licensee-Identified Violations

The following violation of very low significance was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Manual, NUREG-1600, for being dispositioned as an NCV.

Cornerstone: Mitigating System

Criterion III, Design Control, of 10 CFR Part 50, Appendix B requires, in part, that measures be established to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled.

Design changes shall be subject to design control measures commensurate with those applied to the original design, including verifying or checking the adequacy of the design by the performance of design reviews, calculations, or testing. Inadequate design control measures for the AFW system resulted in the installation of non-safety related air receivers, check valves, and piping for the safety related TDAFW pump steam admission control valves during an inappropriate design change in 1981. The calculation for sizing the air receivers and the testing conducted were also inadequate to verify the modifications design requirements. The licensee did not have a clear understanding of the system design, nor was any periodic testing of the control valves air system conducted to ensure continued operability. This was identified in the licensees corrective action program (CAP) as CAP042775 and CAP043013. This finding is of very low safety significance because the licensee concluded the valves would function as required.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

L. Clewett, Plant Manager, Prairie Island Nuclear Generating Plant

J. Kivi, Regulator Compliance Engineer, Regulatory Affairs

C. Lambert, Vice President, Corporate Engineering

T. Lillehei, Engineer, Analysis/Design Configuration Engineering

S. Leingang, Engineer, Engineering Plant & Systems

S. McCall, Manager, Engineering Programs

C. Mundt, Manager, Engineering Plant & Systems

S. Myers, Supervisor, Analysis/Design Configuration Engineering

S. Northard, Manager, Business Support

E. Perry, Manager, Nuclear Oversight

K. Peterson, Engineer, Inspection & Material Engineering

M. Runion, Manager, Engineering Design

G. Salamon, Manager, Regulatory Affairs

T. Silverberg, Director, Site Engineering

D. Smith, Shift Manager/EOP Writer, Procedures

S. Thomas, Engineer, Analysis/Design Configuration Engineering

Nuclear Regulatory Commission

J. Adams, Senior Resident Inspector

D. Karjala, Resident Inspector

A. M. Stone, Chief, Engineering Branch 2

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000282/2005002-01(DRS); NCV Failed to Update Pressure Drop Calculation for

05000306/2005002-01(DRS) Replacement Steam Generators (Section 1R21.1b.1)

05000282/2005002-02(DRS); NCV Failed to Use Appropriate Vortex Methodology for CST

05000306/2005002-02(DRS) (Section 1R21.1b.2)

05000282/2005002-03(DRS); NCV Failed to Specify Correct Minimum Pump Operability Limits

05000306/2005002-03(DRS) for AFW Surveillance Testing (Section 1R21.2b.1)

05000282/2005002-04(DRS); NCV Failed to Validate Heat-Up Transient Design Analysis

05000306/2005002-04(DRS) Assumption for AFW Pump Rooms (Section 1R21.2b.2)

05000282/2005002-05(DRS); NCV Failed to Include AFWP Heat Energy Transfer in Lube Oil

05000306/2005002-05(DRS) Cooler Thermal Performance Analysis (Section 1R21.3b.1)

05000282/2005002-06(DRS); NCV Failed to Maintain Instrumentation Tubing Water Solid

05000306/2005002-06(DRS) (Section 1R21.3b.2)

Closed

05000282/2005002-01(DRS); NCV Failed to Update Pressure Drop Calculation for

05000306/2005002-01(DRS) Replacement Steam Generators (Section 1R21.1b.1)

05000282/2005002-02(DRS); NCV Failed to Use Appropriate Vortex Methodology for CST

05000306/2005002-02(DRS) (Section 1R21.1b.2)

05000282/2005002-03(DRS); NCV Failed to Specify Correct Minimum Pump Operability Limits

05000306/2005002-03(DRS) for AFW Surveillance Testing (Section 1R21.2b.1)

05000282/2005002-04(DRS); NCV Failed to Validate Heat-Up Transient Design Analysis

05000306/2005002-04(DRS) Assumption for AFWP Rooms (Section 1R21.2b.2)

05000282/2005002-05(DRS); NCV Failed to Include AFWP Heat Energy Transfer in Lube Oil

05000306/2005002-05(DRS) Cooler Thermal Performance Analysis (Section 1R21.3b.1)

05000282/2005002-06(DRS); NCV Failed to Maintain Instrumentation Tubing Water Solid

05000306/2005002-06(DRS) (Section 1R21.3b.2)

Discussed

None Attachment