IR 05000255/1997201

Insp Rept 50-255/97-201 on 970916-1114.Deficiencies Identified in Control & Performance of Calculations.Purpose of Insp Was to Evaluate Capability of Selected Sys to Perform Safety Functions Required by Design Bases

ML18065B138
Person / Time
Site:	Palisades
Issue date:	12/30/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML18065B137	List: IR 05000255/1997201 ML18065B136
References
50-255-97-201, NUDOCS 9801130395
Download: ML18065B138 (39)

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OFFICE OF NUCLEAR REACTOR REGULATION Docket No.-:-*-50-255 License No.:

DPR-20 Report No.:

50-255/97-201 Licensee:

Consumers Energy Company Facility:

Palisades Nuclear Power Station Location:

27780 Blue Star Highway Covert, Ml 49043 Dates:

September 16 through November 14, 1997 Inspectors:

James Isom, Team Leader, PECB, NRR Craig J. Baron, Mechanical Engineer*

Robert B. Bradbury, Lead Engineer*

Paul J. Bieniek, Electrical Engineer*

Manzural Huq, Mechanical Engineer*

Douglas M. Schuler, l&C Engineer*

• Contractors from Stone & Webster Engineering Corporation Approved by: Donald P. Norkin, Section Chief Special Inspection Section Events Assessment, Generic Communications, and Special Inspection Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation 9801130395 971230 PDR ADOCK 05000255 G

PDR

• SUMMARY From September 16 through November 14, 1997, the staff of the U.S. Nuclear Regulatory Commission (NRC}, Office of Nuclear Reactor Regulation (NRR}, Events Assessments, Generic Communications and Special Inspection Branch, conducted a design inspection at the Palisades Nuclear Plant. This inspection included on-site inspections during October 6-10, 20-31, and November 10-14, 1997. The inspection team consisted of a team leader from NRR and five engineers from Stone & Webster Engineering Corporatio The purpose of the inspection was to evaluate the capability of the selected systems to perform safety functions required by their design bases, the adherence of the systems to their design and licensing bases, and the consistency of the as-built configuration with the Final Safety Analysis Report (FSAR}. The team selected the safety injection (SI} and component cooling water (CCW}

systems, and their support systems, for this inspection because of their importance in mitigating design-basis accidents at Palisades. The engineering design and configuration control section of Inspection Procedure 93801 was followed for this inspection. The team selected and reviewed relevant portions of the FSAR, design-basis documents (DBDs}, Technical Specifications (TS},

drawings, calculations, modification packages, procedures, and other associated plant document Overall, the team found that the selected systems were capable of performing their design-basis safety functions, although some discrepancies were identified regarding adherence of the systems to their design and licen*sing bases. The DBDs reviewed provided comprehensive information for personnel involved in plant modifications and evaluations. Operability assessments performed during the course of the inspection were comprehensiv The team identified a modification that resulted in the capability for an automatic transfer between redundant safety-related electrical busses. This capability was outside the licensing basi Additionally, the 125-V de system electrical fault protection design implementation was not in accordance with the licensing basis in that the effects of short-circuit fa ult currents were not evaluated at the correct location The team identified many inconsistencies between the installed configurations of instrument tubing and the design basis in the CCW and SI systems. For example, the high-and low-head SI flow transmitters were installed about 8 feet above the flow elements and the team believed that potential air entrapment in the sensing lines could cause significant and unquantifiable errors in the instruments. Information from these flow elements was used in postaccident monitoring and control activitie The team identified numerous deficiencies in the control and performance of calculations. Several calculations were not updated when analytical inputs changed, such as SI pump horsepower inputs to the emergency diesel generator loading calculation and load changes, which affected the main electrical load analysis. Errors in calculations included failure to consider specific uncertainty values in instrument setpoint calculations and a non-conservative initial air temperature in a room heatup calculation. Also, several instrumentation calculations failed to adequately identify the source of inputs; the calculation evaluating the effects of a high energy line break (HELB} on CCW piping did not contain adequate analysis to support the conclusion; and the 125-V de short-circuit calculation was issued without verifying all input parameters or providing any conclusion on the acceptability of the de system. Failure to maintain design-basis calculations current was apparently due, in part, to a weakness in the transfer of information between engineering groups.

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• The licensee had no evidence that the CCW pumps met the vendor-recommended minimum flow requirement under all operating conditions. A preliminary analysis showed that the flow was adequate to meet a revised vendor recommendation. The team identified other instances in which the design basis was not adequately documented. No analysis was available to show that the de loads would operate at the minimum battery voltage stated in the FSAR; there was no analysis to show adequate ac voltage at the 120-V safety-related loads; there was no analysis to show that the battery could carry all required de loads during a design-basis accident with the battery chargers cross-connecte The TS required one battery charger on each bus during normal operation; however, both chargers were disconnected during a monthly evolution of switching battery chargers, and a limiting condition for operation (LCO) was not entered. The team determined that the licensee's failure to enter an LCO during the battery switching evolution had minimal safety impact on the plant. Another TS concern identified was that the 2-hour battery test duration required by the TS appeared non-conservative compared to the 4-hour battery duration required by the design basi The team identified several valves that performed a safety function which were not included in the in-service testing (IST) program. Check valves in the high pressure SI (HPSI) pump minimum-flow recirculation lines, which prevented overpressurization of HPSI pump suction lines, were not tested to verify closure. There were requirements for closure of the safety injection tank vent valves and operation of the relief valve inside containment on the CCW return line in the event of an acciden However, these valves had not been evaluated for inclusion in the IST progra Analyses had been performed, which identified that the CCW system could operate at temperatures in excess of the design-basis temperature, yet a complete evaluation of CCW system performance at these higher temperatures had not been done and the maximum postaccident CCW temperature had not been determine In several instances, the team observed that maintenance and operations support activities were not performed in accordance with plant procedures. Two scaffolds were erected in the vicinity of safety-related equipment without Engineering review, a storage cabinet was improperly located adjacent to safety-related piping and valves, a chainfall was stored adjacent to the shutdown cooling heat exchangers, and a ladder was improperly store Other discrepancies included a potential path for debris to bypass the containment sump screens, installation of incorrectly rated solenoid coils, incorrect implementation of the design-basis lifetime of Agastat time delay relays, incomplete evaluation of a 10 CFR Part 21 notification concerning Agastat relays, and a missed surveillance for safety-related overcurrent relays. The team also identified a number of discrepancies in the FSAR, DBDs, and other plant document The licensee took appropriate actions to resolve the immediate concerns identified by the team. For other issues, the licensee initiated appropriate reviews and evaluations using the corrective action process or initiated changes to documents.

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Ill. Engineering E1 Conduct ofEngineering E1.1 Inspection Objectives and Methodology The primary objective of the design inspection at the Palisades Nuclear Plant was to evaluate the capability of selected systems to perform their safety functions, the adherence of the systems to the design and licensing bases, and the consistency of the as-built configuration and system operation with the Final Safety Analysis Report (FSAR). The systems selected for inspection were the component cooling water (CCW) system and the safety injection (SI) system. These systems were selected on the basis of their importance in mitigating design-basis accidents at Palisade The inspection consisted of reviews at various system levels, including the engineered safeguards functions of the selected systems, component level reviews of selected components within those systems, and interfaces with other systems. The inspection was performed in accordance with the applicable portions of Inspection Procedure (IP) 93801, "Safety System Functional Inspection." The engineering design and configuration control section of the IP was the primary focus of the inspectio The open items resulting from this inspection are* listed in Appendix A and a list of abbreviations used is in Appendix E 1.2 <;;omponent Cooling Water (CCW) System E1.2.1 Mechanical Design Review E 1.2.1.1 Scope of Review The team evaluated the mechanical aspects of the CCW system for its ability to perform the design duty and safety functions during normal power operation and accident conditions. The evaluation included review of the FSAR, Technical Specifications (TS), design basis documents (DBDs), *

standard operating procedures (SOPs), emergency operating procedures (EOPs), testing procedures, piping and instrumentation diagrams (P&IDs), equipment specifications, equipment *

drawings, manufacturer's information, plant modifications, applicable analyses, and calculation The team also walked down the accessible system piping and components, and witnessed simulated plant operation following a loss of coolant accident (LOCA).

E1.2.1.2 Inspection Findings E1.2.1.2(a) CCW System Performance The team verified that the CCW system could provide the flow required for various plant operating modes. The CCW system consisted of three pumps in parallel. Each pump was rated for 6000 gallons per minute (gpm) at 160 ft of head. FSAR Table 9-7 stated that the minimum CCW flow required following a LOCA was 1647 gpm initially and 5578 gpm after the recirculation actuation signal (RAS). Only one CCW pump would be available, assuming a limiting single failure. The team reviewed the results of testing performed in accordance with Q0-15, "lnservice Test Procedure -

Component Cooling Water Pumps," Revision 14, and verified that one pump could provide the required post,,accident flow.

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The highest flow required in any condition was 7,076 gpm during normal shutdown cooling as stated in FSAR Ta~l.:t 9-7. The team reviewed the results of testing performed in accordance with T-223,

"Component Cooling Water Flow Verification,* Revision 7, and verified that two CCW pumps could provide this flo The team also verified the adequacy of the available net positive suction head (NPSH) for the CCW pumps. The pump manufacturer did not include the NPSH requirements on the pump characteristic curves. The team observed that the surge tank was located at an elevation that provided a pressure head of more than 65 ft at the CCW pump suction. The CCW pumps were single-stage, double sided, horizontal, centrifugal pumps. Based on the team's experience with this type of pump in cold water service at similar flow rates, an NPSH of approximately 20 ft would be required. Therefore, the team concluded that the available NPSH of greater than 90 ft was adequat The team verified that the CCW system could accommodate a single failure in CQnjunction with postulated accidents. CCW pumps had adequate pumping capacity and the CCW system contained adequate number of valves to enable the CCW system to meet the design flow requirements in the event of a single failure of an active componen The team questioned whether the CCW system design met the vendor-recommended minimum flow of 2000 gpm for the CCW pumps under all operating conditions. The team was concerned that small differences in the pump operating characteristics could cause significant differences in flow through each pump during parallel pump operation due to the flatness of the pump operating curves at low flows. The licensee had no analysis available to demonstrate that the CCW pumps met the minimum flow requirements. During the inspection, the licensee developed a preliminary system flow model, which showed that, when all three pumps were started upon receiving a safety injection system (SIS) signal, the minimum pump flow was through CCW pump P-52A at 1768 gpm. The licensee received a revised minimum flow requirement of 1600 gpm from the pump manufacture The team's review of the licensee's completed flow model calculation will be an Inspection Followup Item 50-255/97-201-0 The team verified the heat removal capability of the CCW heat exchangers by reviewing the results of various accident analyses. The licensee had performed the following LOCA analyses:

EA-D-PAL-93-207-01, "LOCA Containment Response Analysis With Reduced LPSI Flow Using CONTEMPT El-28 Code, n Revision 0,

EA-D-PAL-93-272-03, "LOCA Containment Response Analysis with Degraded Heat Removal System Using CONTEMPT El-28A Computer Code,* Revision 0, and

EA-GEJ-96-01, *A-PAL-94-324 Containment Spray System (CSS) Sensitivity on the Containment Heat Removal During Recirculation (Post-RAS),n Revision The team verified that the input assumptions relating to the CCW system for the above analyses were correct. The above LOCA analyses demonstrated that the heat exchangers could remove sufficient heat from containment following a LOCA to keep the containment pressure and temperature within the de~ign limits. In each case, the analysis documented a CCW temperature exiting the shutdown coolers exceeding the system design temperature of 140 degrees Fahrenheit (140 °F) as stated in FSAR table 9-6 and DBD 1.01, "Component Cooling Water,n Revision The team noted that the licensee accepted the maximum CCW temperature that resulted from the scenarios analyzed in EA-D-PAL-207-01 and EA-D-PAL-93-272-03 by Corrective Action D-PAL-93-272G, based primarily on an evaluation of the effects on pipe stress. However, the licensee had not

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considered the other negative effects, such as any detrimental effects from elevated CCW temperature on pump seals. Also, the licensee had not determined the maximum possible CCW temperature under worst case conditions and had not identified that a change to the FSAR could be require The team reviewed the latest LOCA analysis, EA-GEJ-96-01, and determined that it documented a CCW temperature exiting the shutdown cooling heat exchanger was 184 °F. The licensee determined the system was operable under this condition and issued Condition Report (CR) C-PAL-97-1363F to determine the most limiting CCW temperature for any condition and to evaluate all the effects resulting from that limiting temperature on the CCW syste It appeared that the requirements of 10 CFR 50, Appendix B, Criterion Ill, "Design Control," were not met in this case in that the design basis for the CCW system, as defined in 10 CFR 50:2, did not encompass the entire range of bounding temperatures. The team identified this item as Unresolved Item 50-255/97-201-0 E1.2.1.2(b) CCW Surge Tank The team verified that the CCW surge tank was equipped with an appropriate pressure relief valve, which was connected to the plant radwaste system. The team verified that the tank was seismically supported and installed at an elevation that provided approximately 28 pounds per squa're inch gauge (psig) suction pressure to the CCW pumps. The team concluded that the surge tank provided adequate water volume and pressure head for the system and was adequately protected against pressure outside the design basi E1.2.1.2(c) CCW In-Service Testing (IST)

The team reviewed EM-09-04, "In Service Testing of Selected Safety-Related Pumps," Revision 18, and verified that the CCW pump testing met the re'quirements of the appropriate American Society of Mechanical Engineers (ASME)/American National Standards Institute (ANSI) operating and maintenance (OM) standard as implemented by the ASME Boiler and Pressure Vessel Code,Section XI. The team also witnessed an in-service test of pump P-52C, performed in accordance with Procedure No. Q0-15, "Surveillance Test Procedure," Revision 7, and verified that the pump performance met the acceptance criteria of the test procedur The team reviewed the CCW portion of IST program documents for the CCW system valves: EM-09-02, "In Service Testing of Valves," Revision 18; EGAD-EP-0, "In Service Test Program Valve Test Table and Valve Reference Flow Rate," Revision 11; QQ-6, "Cold Shutdown Valve Test Procedure (Includes Containment Isolation Valves)," Revision 28; and the "Palisades lnservice Testing Program Basis Document,* dated October 30, 1997. With the exception of valve RV-0939, the team verified that the CCW valves were tested to meet the requirements of the appropriate ASME/ANSI OM standard as implemented by the ASME Boiler and Pressure Vessel Code, Section.Xl(IST).

The team reviewed C-PAL-96-1-63-01, "120 day response to GL 96-06, Assurance of Equipment Operability and Containment Integrity during Design Basis Accident Condition," Revision 0, which was the licensee's response to Nuclear Regulatory Commission (NRC) Generic Letter 96-06,

"Assurance of Containment Operability and Containment. Integrity During Design-Basis Accident Conditions," and observed that the licensee took credit for relief valve RV-0939 to protect the CCW piping inside containment from overpressurization in the event of a LOCA. RV-0939 was not included in the IST program. The team questioned whether RV-0939 performed a safety function and if it should have been included in the IST program. The licensee issued CR C-PAL-97-1686 to evaluate this discrepancy. 10 CFR 50.55a requires IST in accordance with ASME Section XI of

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valves that perform a safety function. It appeared that the licensee did not fully implement these requireme.n,!~.for RV- 0939. The team identified this item as part of Unresolved Item 50-255/97-201-0 E1.2.1.2(d) CCW Piping Inside Containment FSAR Section 9.3.2.3 stated that the CCW piping within containment was not vulnerable to failure caused by a high energy line break (HELB) and referred to Deviation Report (DR) D-PAL-89-061,

"Post Accident Operation of CCW System,* dated March 23, 1989, for the evaluation. This DR referred to Engineering Analysis (EA) EA-GW0-7793-01, "CCW Piping Inside Containment HELSA,"

Revision 0. This EA was reviewed by the team, and it concluded that the CCW piping inside containment was not affected by HELBs, but did not contain the analysis performed or a reference to the analysis. The EA contained an outline of the methodology, listed the drawings and walkdowns used, and referenced the source of the postulated HELBs. Palisades Administrative Procedure No. 9.11, "Engineering Analysis,* Revision 9, stated that an EA shall present an argument which substantiates the conclusion of the EA. The EA also contained an error in the identification of the Systematic Evaluation Program (SEP) topic number for evaluation of the effects of internally generated missiles. The licensee initiated Engineering Assistance Request (EAR) EAR-97-0632 to revise EA-GW0-7793-0 During the inspection, the licensee issued Revision 1 of EA-GW0-7793-01, which included a discussion of the walkdown analysis used and corrected the SEP references. This revised EA was acceptable to the team. It appeared that the requirements of 10 CFR Part 50, Appendix 8, Criterion Ill, "Design Control," regarding verifying the adequacy of designs were not adhered to in this case. Also, the requirements of the licensee's Administrative Procedure 9.11 were not fully met in that EA-GW0-7793-01, Revision 0, did not contain full substantiation of the conclusion. The team identified this item as Unresolved Item 50-255/97-201-0 E1.2.1.2(e) CCW Containment Isolation The team reviewed the two CCW containment piping penetrations. Both the supply and return lines had double isolation valves, and the containment isolation for each penetration was designed to accommodate a loss of air as well as a single active failure. The team concluded that the CCW containment penetrations were consistent with the design and licensing base E1.2.1.2(f) CCW Interfacing Systems The team reviewed the capability of the service water (SW) system to provide sufficient SW to the CCW heat exchangers during accident conditions. The SW system consisted of three vertical pumps, one of which was adequate to provide the required postaccident flow. A non-critical header and two critical headers were used to provide water to various components. The team observed that a single air-operated valve, CV-1359, isolated SW to the non-critical header in the event of a LOC Should this valve fail to close because of a single failure, three SW pumps would be operating and the licensee stated that adequate flow would be provided. The team agreed with the licensee because of the available high SW flow capacity.

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E1.. 2.1.2(g) CCW System Modifications The.team reviewed two modifications and one temporary modificatio One modification involved replacement of differential pressure indicators (DPls) DPl-0918A and DPl-09188 in the CCW pump suction and discharge piping. The replacement DPls were similar to the old DPls, but were from a different manufacturer. The licensee performed a functional equivalent substitution (FES) evaluation (FES-97-020) of the replacement indicators and concluded that the replacement DPls were acceptable. The other modification reviewed by the team changed the shaft of CCW pump P-52A to improve the impeller fit. The licensee performed the CCW pump shaft change using their modification process (FES-96-216) to ensure that the modification was technically acceptabl The temporary modification "gagged* closed relief valve RV-2108 to prevent the valve from leaking by RV-2018 had failed to reseat after a spurious actuatio The team found the safety evaluations performed for the modifications were satisfactory and that the two FESs were consistent with the design basi E1.2.1.2(h) CCW System Walkdown The team walked down the accessible portion of the CCW piping and components and found the arrangement was in conformance with the P&ID. The team observed that the pumps were running smoothly without any discernible vibration and that the CCW component area was clean and general housekeeping was goo E1.2.1.3 Conclusion The team concluded that the mechanical aspects of the CCW system could perform the design functions of cooling the safety-related equipment during the normal operating mode and postaccident *

condition The team identified that the maximum CCW temperature had not been determined and the effects of this temperature on various CCW components were not fully analyzed, and a relief valve was possibly not tested as required. However, these discrepancies did not change the conclusion that the system could provide adequate safety-related coolin E1.2.2 Electrical Design Review E1.2.2.1 Scope of Review The team evaluated the electrical loads required for the CCW and interfacing systems to perform their design-basis functions under normal and accident conditions. This evaluation addressed alternating current (ac) bus loading, direct current (de) battery loading and distribution, protective device coordination, and modification E1.2.2.2 Inspection Findings The team reviewed EA-ELEC-LDTAB-005, "Emergency Diesel Generator 1-1 & 1-2 Steady State Loading," Revision 4, and EA-ELEC-VOLT-13, "Palisades Loss of Coolant Accident With Offsite Power Available," Revision 0. The team verified that all major CCW electrical loads were accounted for in the electrical load analyses for both normal and accident conditions and that the motors were

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sized to accelerate the CCW pumps and drive them for long-term continuous operation. The team also reviewed the Class 1 E 125-V de battery and 120-V ac preferred bus loadings for the CCW loads and determined. that adequate capacity was available for the required loads. Protective relaying for the 2400-V ac motors was determined to be properly set and calibrated so as to protect the electrical distribution system and the CCW load E1.2.2.3 Conclusions The electrical design for components that performed the normal and accident functions of the CCW system supported the design-basis functions of the system. The electrical system provided adequate redundant and safety-related power to the CCW load The team identified several discrepancies conceming the electrical design related to the CCW system. These discrepancies are discussed in Section E1.4, "Electrical Interface Systems," of this repor E1.2.3 Instrumentation and Controls Review E1.2.3.1 Scope of Review The team evaluated the ability of the instrumentation and controls for the CCW system to perform the design safety functions. The team reviewed sections of the FSAR, applicable TS sections, DBDs, SI actuation signals, P&IDs, plant procedures, and calibration data. System walkdowns were also conducte E1.2.3.2 Inspection Findings E1.2.3.2(a) lnstrumentati.on and Controls Design Review The team reviewed the* implementation of the licensee's commitment to NRC Regulatory Guide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident," Revision 3, as described in FSAR Appendix 7C. The RG stated a range for CCW flow instrumentation of 0-110 percent. Since there was no instrument to directly measure CCW flow, the licensee used a combination of instruments, including TE-0912 and TE-0913, which measure shutdown cooling heat exchanger outlet temperature, to indicate flow. Use of instruments (other than flow indicators) to monitor for CCW flow was determined as acceptable by the NRC (a letter from NRC to Consumers Power Company, dated July 19, 1988, entitled "Palisades Plant - Response to Generic Letter 82-33 Conformance to Regulatory Guide 1.97 "Instrumentation for Light-Water-Cooled_Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident"). The required range for these TEs in FSAR Appendix 7C was 0-180 °F. This range did not encompass the temperature determined in EA-GEJ-96-01, "A-PAL-94-324 Containment Spray System (CSS) Sensitivity on Containment Heat Removal During Recirculation (Post-RAS)," Revision 1. This analysis determined an outlet temperature of the CCW from the shutdown cooling heat exchanger of 184 °F. The licensee issued CR C-PAL-97-1363E to evaluate the process instrumentation and controls associated with the CCW 3ystem for the effects of the higher temperature predicted by the analysi The licensee did not appear to meet their commitment to NRC RG 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident," in that the installed CCW temperature indicators were not capable of monitoring the full temperature range expected to be observed in the CCW system. The team identified this item as part of Unresolved Item 50-255/97-201-0 **

E1.2.3.2(b) Procedure Review The team reviewed the following five proeedures:

Q0-14, "lnservice Test Procedure-Service Water Pumps,* Revision 7,

RT-71F, "Component Cooling System Class 2 and 3 lnservice Test,* Revision 0,

RT-71K, "Class 2 System Functional Test for Shutdown Cooling System,* Revision 1,

RT-213, "CCW Flow Test of the CCW, LPSI, Containment Spray Pumps, and CCW Heat Exchangers," Revision 3, and

T-216, "Service Water Flow Verification," Revision Each procedure was adequate and consistent with the desig E1.2.3.3 Conclusion The team found the instrumentation and controls portion of the CCW system capable of providing adequate control and monitoring and capable of performing its design safety function. Two instruments (TE-0912 & TE-0913) which could not measure the entire range of possible temperatures were identified

E1.3 Safety Injection (SI) System E1.3.1 Mechanical Design Review E1.3.1.1 Scope of Review The mechanical design review of the SI system included a review of the applicable FSAR sections, TS sections, licensee event reports (LERs), DBDs, P&IDs, calculations, design modifications, equipment specifications, the operations and testing procedures required to assess consistency with the system design and licensing basis, and the evaluation of several generic items. In addition, the

. team performed several walkdowns of the accessible portions of the SI system and observed a LOCA scenario on the simulato E1.3.1.2 Inspection Findings E1.3.1.2(a) SI System Performance The team reviewed the available licensing, design, and operations documents related to the capability of the SI system to provide adequate emergency core cooling flow under accident conditions. The team reviewed EA-SDW-95-001, "Generation of Minimum and Maximum HPSl/LPSI System Performance Curves Using Pipe-Flo," Revision 2, and EA-A-PAL-96-003, "ECCS Evaluation in Post-RAS Recirculation Modes Using Pipe-Flo," Revision 1, which addressed the SI system performance capability under various accident conditions. These analyses considered degraded pump conditions and various system single failures. The results of the analyses were used as input to the LOCA analysis. The team found these analyses satisfactory and found the SI system performance information presented in Section 14 of the FSAR consistent with the applicable licensing, design, and operations document E1.3.1.2(b) Safety Injection and Refueling Water Tank (SIRWT)

The team reviewed the available licensing, design, and operations documents related to the capability of the SIRWT to provide an adequate water supply to the SI system under accident

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conditions. The team reviewed EA-C-PAL-95-877D, *evaluation of the Potential for Excessive Air Entrainment Caused by Vortexing in the SIRWT During a LOCA,° Revision 0. The EA concluded that the emergency core cooling system (ECCS) pump performance would not be significantly affected by air entrainment due to vortexing. The team found that the SIRWT design was consistent with the applicable licensing, design, and operations documents, and that the tank was capable of performing its function under accident condition E1.3.1.2(c) Transfer to SI Recirculation Mode The team reviewed the available licensing, design, and operations documents related to the transfer of the SI pumps' suction supply from the SIRWT to the containment sump under accident condition EA-SC-88-185-0, "Calculation of SIRW Tank Volume Discharge Subsequent to RAS,n Revision 0, addressed the automatic transfer and determined that approximately 20,000 gallons of water would remain in the SIRWT after the transfer was completed. The team found the EA satisfactory and found that the portions of the SI system associated with the transfer of the SI pumps' suction supply from the SIRWT to the containment sump under accident conditions were consistent with the plant design and licensing base E1.3.1.2(d) Net Positive Suction Head (NPSH)

The team reviewed the available licensing, design, and operations documents related to the required and available NPSH of the low pressure safety injection (LPSI) and high pressure safety injection (HPSI) pumps operating under accident conditions for both the injection phase of the SI system from the SIRWT and the recirculation phase of the SI system from the containment sum EA-C-PAL-95-877D, "Evaluation of the Potential for Excessive Air Entrainment Caused by Vortexing in the SIRWT During a LOCA," Revision 0, addressed the available NPSH during the injection mode of SI operation, and determined that the LPSI and HPSI pumps would have adequate NPS The HPSI pump NPSH available from the containment sump during recirculation was calculated by EA-A-PAL-96-003, "ECCS Evaluation in Post-RAS Recirculation Modes Using Pipe-Flo," Revision This EA determined that HPSI pump operation from the containment sump would not provide sufficient NPSH under all conditions, and that subcooling flow from the containment spray pumps would be required to ensure adequate NPSH. The team verified that the plant EOPs contained the applicable criteria for establishing subcooling flow. The team noted that the LPSI pumps would not be operated during the recirculation mod E1.3.1.2(e) SI System Valve Operation The team reviewed the available licensing, design, and operations documents related to the capability of SI system valves to perform their required functions under accident conditions. This review included EA-AOVSYS-ESS-01, "System Design Basis Review for Air Operated Valves (AOV)

in the Engineered Safeguards System (ESS),n Revision 3; EA-NL-92-185-01, "Worst Case Operating Conditions for the LPSl/(shutdown cooling)SDC System MOVs,n Revision 1; EA-NL-92-185-08,

"Worst Case Operating Conditions for the HPSI, (redundant high pressure safety injection) RHPSI, and Hot Leg Injection (motor-operated valves)MOVs," Revision O; and EA-CPCo/PAL-DSS-94-01,

"Worst Case Pressures and Flows for MOVs in the (Generic Letter) GL 89-10 Program," Revision The team found that the system valve design was consistent with the applicable licensing, design, and operations documents and that the valves were capable of performing their functions under accident condition E1.3.1.2(f) SI System Testing The team reviewed the available licensing, design, and operations documents related to testing of SI system mechanical components. This review included the applicable TS and the applicable surveillance test procedures. The team compared acceptance criteria in the surveillance test procedures with the design-basis requirements for the equipment and found the two consisten With the exception of the following IST issues, the team found that the testing of SI system mechanical components was consistent with the applicable licensing, design, and operations documents, and that the testing was sufficient to verify that the mechanical equipment was capable of performing its required functions under accident condition *

The team identified a lack of closure verification testing on SI system check valves that could potentially result in an overpressure condition affecting the low-pressure piping on the suction of the HPSI pumps. The minimum flow recirculation lines associated with the two HPSI pumps and the two LPSI pumps were interconnected upstream of the air-operated minimum flow recirculation isolation valves. In the event that only one HPSI pump was operating under postaccident conditions with the *

minimum flow recirculation isolation valves closed, back leakage through the minimum flow piping associated with the idle HPSI pump could over pressurize the idle HPSI pump suction pipin Backflow between the HPSI minimum flow lines should be prevented by check valves CK-ES3339 or CK-ES3331, and CK-ES3340 or CK-ES3332. However, EGAD-EP-01, "lnservice Testing Program -

Valve Test Program,* Revision 10, indicated that closure verification testing of these check valves was not included in the IST program. The team asked the licensee if closure of these check valves was considered a safety function requiring IS The licensee initiated CR C-PAL-97-1660 to evaluate the testing requirements of these check valve On November 10, 1997, the operability determination concluded that these system check valves had not been subject to closure verification testing as required, and both HPSI pumps were declared inoperable. In accordance with TS sections 3.0.3, 3.3, and 4.0.3, the licensee entered a Limiting Condition for Operation (LCO) action statement; performed closure verification testing of check valves CK-ES3339 and CK-ES3340, and verified the operability of these valves. The licensee stated that closure verification testing of these check valves would be added to the IST progra The team also identified a lack of closure verification testing on SI system valves that could potentially result in a safety injection tank (Sin being degraded under postaccident conditions. The normally closed SIT vent valves, CV-3051, 3063, 3065, and 3067, could be opened in accordance with SOP-3, "Safety Injection and Shutdown Cooling System," Revision 28, to reduce SIT pressur SOP-3 did not require the affected SIT to be declared inoperable when a vent was opened. When a vent valve was opened, the SIT pressure boundary (250 psig design pressure) was exposed to the SIT vent header piping (100 psig design pressure). SOP-3 did not include directions to isolate an open vent valve in the event of an accident. EGAD-EP-01, "lnservice Testing Program - Valve Test Program,* Revision 10, indicated that closure verification testing of these valves was not included in the IST program. The team asked the licensee if the failure of a valve to close could result in a SIT being degraded under accident conditions, and if closure of these valves was considered a safety function requiring IST testin The licensee initiated CR C-PAL-97-1592 to evaluate this item and placed caution tags on the control room switches for vent valves CV-3051, 3063, 3065, and 3067 to prevent the valves from being opened without entering an LCO for the SITs. The licensee also stated that these valves had been opened rarely during plant operatio O CFR 50.55a requires in-service inspection in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. This code requires testing of valves which perform a safety function. It

appeared that the licensee did not implement these requirements with regard to valves CK-ES3339, CK-ES3340, CV-3051, CV-3063, CV-3065, and CV-3067. The team identified this item as part of Unresolved Item 50-255/97-201-0 E1.3.1.2(g) Interfacing System Review The instrument air penetration through the containment consisted of a check valve and a normally open air-operated valve which did not automatically close on a containment isolation signal in order to provide a supply of non-safety-related instrument air to controls inside containment. The instrument air system piping inside containment was not missile protected and, therefore, was considered connected to the containment atmosphere. The team observed that, should this piping be ruptured by the dynamic effects of a LOCA or main steam line break (MSLB), compressed air would be supplied to the containment atmosphere and would add to the amount of air in containment at accident initiation. The licensee stated that the containment pressure analysis for a MSLB, which yielded the maximum containment pressure, did not account for any additional air from an instrument air line ruptur The licensee performed a conservative assessment during the inspection and determined that the maximum containment pressure with an air line rupture would result in a small pressure increase and would not exceed the design pressure of 55 psig. Tl°'e licensee issued Action Item Record (AIR)

A-PAL-97-103 to determine the actual expected pressure increase from this air line rupture and to incorporate the effects of this rupture in the upcoming containment analysis for Cycle 14 as necessar The team reviewed the portion of the auxiliary building heating, ventilation, and air conditioning (HVAC) system which served the engineered safeguards (ESG) rooms. These rooms contained the HPSI and LPSI pumps and associated equipment. The team reviewed EA-D-PAL-93-272F-01,

"Engineered Safeguards Room Heatup Following LOCA in Conjunction With a (Loss of Offsite Power)LOOP, D Revision 0, which was the current analysis of ESG room temperature. This EA determined a maximum post-LOCA temperature of 117 °F. This EA used 90 °F as the initial room temperature. The team questioned this assumption because DBD-1.07, "Auxiliary Building HVAC Systems, D Revision 1, stated that the normal design temperature of the ESG rooms was 104 °F. The licensee stated that 104 °1; was a design specification number and that the temperature was controlled by the room coolers at a maximum of 95 °F, including instrument erro The team reviewed Equipment Qualification (EQ) File Report E-48 EMA-3 for the HPSI pump motors (P-66A and B). This EQ file reported a qualified life of the motors of 379 years using the 117 °F post-LOCA room temperature from EA-D-PAL-272F-01. The team questioned the qualification of the motors since the 117 °F temperature appeared nonconservative. The licensee stated that the motors were documented as qualified for a 135 °F maximum post-LOCA temperature before the EQ file was revised to reflect 117 °F, and, even if an initial room temperature of 100 °F were assumed, the maximum post-LOCA room temperature would be less than 127 °F (117 + 10), which was below the 135 °F qualification temperature previously documented. The team agreed with the license The licensee issued CR C-PAL-97-1558 to document this discrepancy and added an action to existing AIR A-PAL-94-257 to revise EA-D-PAL-93-272F to use a corrected initial room temperatur The team reviewed the HVAC system serving the cable spreading room. The team observed that DR F-CG-91-072 was prepared in May 1991 when it was discovered that the assumptions in calculation EA-FC-573-2, "Calculated Required Air Flow for Inverter/Charger Cabinet Cooling Fan,"

dated October 3, 1982, used an ambient temperature of 94 °F instead of the correct design-basis temperature of 104 °F. The Safety System Design Confirmation (SSDC) Team that found this discrepancy recommended that the EA be updated. Procedure 9.11, "Engineering Analysis,"

Revision 9, required all EAs to be revised if analytical inputs or major assumptions change. The licensee decided not to revise the EA, and the discrepancy was recorded in DB~.02 (125-V de system) and DB~.03 (preferred ac system). The fans were installed in 1983 and were not safety related. DR F-CG-91-072 was closed in October 1994, when the decision was made not to revise the calculation. The licensee stated that specifications were being developed for replacing the inverters and chargers during the time the discrepancy was being evaluated and that this knowledge contributed to the decision not to update the EA. The inverters and chargers were scheduled to be replaced in the near future by Specification Change (SC) SC-96-033. The new equipment would have internal cooling fans designed for a 104 °F maximum ambient and SC-96-033 would supersede EA-FC-573-2 upon installation. The team had no other concerns about the cable spreading room HVAC system. It appeared that the requirements of 10 CFR Part 50, Appendix B, Criterion II,

"Quality Assurance Program,* were not followed in this case in that the requirements of Procedure 9.11 regarding revising EAs were not fully implemented. The team identified this item as part of Unresolved Item 50-255/97-201~ E1.3.1.2(h) SI System Calculations The team identified the following discrepancies in SI system mechanical calculations:

EA-DBD-2.01-004, "Electrical and Mechanical Failure Analysis for the Low Presswe Safety Injection System," Revision 0, pages 10 and 25, identified a situation in which a loss of an emergency diesel generator (EDG) during a large-break LOCA would result in only one LPSI pump and two LPSI injection valves being operable. The EA stated: "The acceptability of this situation could not be verified.n The team asked if this statement was correct. The licensee replied that the statement was not current, and that the statement appeared to be based on superseded calculation ANF-88-107, *palisades Large Break LOCA/ECCS Analysis With *

Increased Radial Peaking," Revision 1. Calculation ANF-88-107 was superseded by Seimens calculation EMF-96-172, "Palisades Large Break LOCA/ECCS Analysis," Revision 0. The licensee initiated Engineering Assistance Request (EAR) 97-0635 to revise EA-DBD-2.01-00 *

EA-A-NL-92-185-01, "Worst Case Operating Conditions for the LPSl/SDC System MOVs,"

Revision 1, addressed the most limiting conditions under which the system motor-operated valves (MOVs) were required to open and close. This analysis included MOVs M0-3015 an M0-3016. These valves were the isolation valves installed in the shutdown cooling inlet piping from primary coolant system (PCS) loop 2. For all normal operations - other than shutdown cooling being in service, - the valves were electrically locked closed. Page 19 of EA-A-NL-92-185-01 stated that the scenario that could produce the most limiting differential pressure was that these valves would be required to close in the event of a downstream pipe break. The EA addressed a potential 12-in. downstream pipe break and determined that complete depressurization and blowdown of the PCS to the hot-leg elevation would occur before operators could enter the EOPs and attempt to isolate the break. Therefore, the analysis then established a maximum flow rate of 4120 gpm through valves M0-3015 and M0-3016, based on a normal system flow rate of 3000 gpm and a calculated leakage of 1120 gpm through a break ofa 1-1/2-inch branch line downstream of the valves. The team asked !he licensee to provide the basis of the postulated 1-1/2-inch branch line failure, since it did not appear to be consistent with the postulated pipe crack used in the internal flooding analysis of the safeguards areas (EA-C-PAL-95-1526-01, "Internal Flooding Evaluation for Plant Areas Outside of Containment," Revision 0). The licensee verified that the flooding analysis break flow was different and that this difference would not affect the conclusions of EA-A-NL-92-185-0 Assumptions 5.9 and 5.10 of EA-A-NL-92-18~1 stated that the HPSI and LPSI injection flows to the loops were approximately equal under post-accident conditions. These assumptions did not appear consistent with the flow values calculated in EA-SDW-95-001,

"Generation of Minimum and Maximum HPSl/LPSI System Performance Curves Using Pipe-Flo,* Revision 2. The team asked the licensee to provide the bases of these value The licensee stated that the values were not current and verified that the difference between these values and the current values would not affect the EA result The licensee initiated CR C-PAL-97-1670 to resolve the discrepancies in EA-A-NL-92-185-0 *

EA-E-PAL-93-004E-01, "IST Check Valve Minimum Flow Rate Requirements to Support Chapter 14 Events,* Revision 0, identified 1601 gpm as the required test flow for the LPSI injection check valves. The team observed that this value appeared to be less limiting than the values calculated in EA-SDW-95-001, "Generation of Minimum and Maximum HPSl/LPSI System Performance Curves Using Pip8-Flo,* Revision 2. The licensee initiated CR C-PAL-97-1603 to address this discrepancy. The licensee determined that the LPSI test flow presented in EA-E-PAL-93-004E-01 was less than the current calculated requiremen However, the actual LPSI check valve flow acceptance criterion in IST Procedure Q0-88,

"ESS Check Valve Operability Test (Cold Shutdown),* Revision 17, was verified to be 1690 gpm, which was greater than the current calculated requirement. The licensee stated that the affected documentation will be correcte Administrative Procedure 9.11,* Engineering Analysis,~ Revision 9, Section 6.1.5.c stated that an analysis shall be revised if analytical inputs changed. In the above instances, engineering analyses were not updated to reflect analytical*input change. The licensee initiated C-PAL-97-1636 to evaluate the overall issue of calculation control. The team identified this item as part of Unresolved Item 50-255/97-201-0 E1.3.1.2(i) SI System Modifications The team reviewed two significant design modifications to the mechanical portions of the SI syste Facility Change (FC) FC-929, "Engineered Safeguards Recirculation Line Modification," Revision 1, installed a spectacle flange and a ball valve in the CSS recirculation line to prevent leakage from the CSS to the SIRWT under post-accident conditions. FC-756, "HPSI Pump Miniflow Bypass," dated December 9, 1988, installed piping and throttle valves to bypass flow around the HPSI pump minimum flow restricting orifiees during surveillance testing. The team found the problem identification and justification for the modifications were clearly stated, the 10 CFR 50.59 evaluations were correct, post-modification testing was appropriate, and the required plant documentation was updated to reflect the modificatio E1.3.1.2(j) SI System Walkdown The team performed several walkdowns of the accessible portions of the SI system to verify that the system configuration was consistent with the design basis. These walkdowns included the SIRWT area, east ESG room, west ESG room, and the control room. In addition, the team utilized the

"surrogate tour" to view the SI system inside containment, including the containment sump area. the SI injection valve area, and the SIT area. The system configuration was consistent with the SI system P&IDs in the areas walked down. With the exception of the deficiencies described in the following paragraphs, the system configuration was consistent with the design basi During an SI system walkdown on October 6, 1997, the team observed scaffolding installed adjacent to the SIRWT on the roof of the auxiliary building. The team questioned how the installation of

scaffolding in the vicinity of safety-related equipment was controlled to prevent damage to the safety-related equipment during a seismic event. The licensee provided Procedure MSM-M-43,

"Scaffolding,8 Revision 2, for the team's review. Section 5.3 of this procedure required an engineering review of scaffolding installed in the vicinity of safety-related equipment. However, the licensee determined that the scaffolding observed during the walkdown had not received engineering review in accordance with the procedure. The licensee initiated CR C-PAL-97-1417 to address the scaffolding installation, and the scaffolding was removed on October 8, 199 EA-C-PAL-97-1417A-01, "Operability Reassessment of SIRWT Scaffolding,,; Revision 0, was completed during the inspection. Based on a structural analysis of the maximum loading on the SIRWT due to seismic interaction with the scaffolding during a safe shutdown earthquake, this analysis concluded that the SIRWT was not inoperable due to this nonconforming conditio During another SI system walkdown on October 30, 1997, the team observed additional scaffolding installed in the east ESG room adjacent to safety-related piping. An evaluation by the licensee determined that this scaffolding had not been installed in accordance with Procedure MSM-M-43,

"Scaffolding,* Revision 2. The licensee initiated CR C-PAL-97-1585 to address this scaffolding installation and, based on a visual inspection, concluded that this nonconforming scaffolding would not render any safety-related piping or components inoperable. The licensee removed the

scaffolding. In addition, the licensee performed a walkdown of all plant scaffolding during the inspection and verified that there were no additional nonconforming conditions. The licensee stated that all scaffolding erections would cease until appropriate personnel underwent remedial trainin The team observed the following three separate conditions in the west ESG room involving potential seismic interactions with safety-related equipment. The team noted that, during a seismic event, unrestrained items could potentially damage safety-related piping and equipment. The safety-related piping and equipment in the west ESG room were required for operation of the HPSI, LPSI, and containment spray systems in the event of an acciden *

The team observed an unsecured operations storage cabinet located adjacent to safety-related piping and valves. The team asked the licensee if the condition was in accordance with plant procedures. The licensee initiated CR C-PAL-97-1587, which determined that the cabinet was not placed in accordance with the spacing requirements of Administrative Procedure 1.01," Material Condition Standards and Housekeeping Responsibilities,"

Revision 11. The operability evaluation concluded that the nonconforming condition did not result in any safety-related equipment being inoperable. The cabinet was laid on its side to eliminate the toppling concern. The licensee stated that the cabinet would be removed from the are *

The team observed an unsecured chainfall located adjacent to and above the shutdown cooling heat exchangers. A similar chainfall in the east ESG room was secured. The team asked the licensee if the condition was in accordance with plant procedures. The licensee determined that the chainfall location was not in accordance with Administrative

Procedure 1.01, and initiated CR C-PAL-97-1586. The operability evaluation concluded that the nonconforming condition did not result in any safety-related equipment being inoperabl The licensee stated that the chainfall chains would be moved away from the heat exchange The team observed a ladder in the west ESG room that appeared to be improperly stored.

The ladder was lying on the floor under the installed ladder rack. The team asked the licensee if the condition was in accordance with plant procedures. The licensee initiated CR C-PAL-97-1601 and determined that the ladder location was not in accordance with the

"Palisades Ladder Control Policy for Operating Spaces," dated May 14, 1997. The CR

• concluded that, although the ladder storage did not meet the ladder control policy, the nonconforming condition did not result in any safety-related equipment being inoperable. The licensee stated that the ladder was removed from the are Procedure MSM-M-43 required an engineering review of scaffolding installed in the vicinity of safety-related equipment. Procedure 1.01 and the "Palisades Ladder Control Policy for Operating Spaces,"

dated May 14, 1997, contain requirements for storing items in the vicinity of safety-related equipment. In these cases, the licensee did not comply with the procedural requirements for activities affecting quality as required by 10 CFR Part 50; Appendix 8, Criterion V, "Instructions, Procedures, and Drawings: The team identified this item as Unresolved Item 50-255/97-201-0 During the surrogate tour, the team observed the ends of two vent pipes that connected the containment sump to the 590-ft elevation of the containment. Tt)e team asked the licensee to explain the design of these vent lines. During a review of the vent lines, the licensee determined that the top of the vents were located inside the containment at an elevation of approximately 595 ft. The maximum calculated post-accident water elevation was at elevation 597 ft. The vent pipes did not have screens on their inlets. The licensee also determined that the two vent lines entered the containment sump inside the sump screens, creating a potential path for debris to enter the ECCS pump suction piping under post-accident conditions. The licensee initiated CR C-PAL-97-1571, on October 29, 1997, to evaluate this condition and determined that the postulated type and quantity of debris that could enter the vent pipes under' post-accident conditions would not prevent the SI and containment spray systems from performing their safety function, and that these systems were operable under this condition. The licensee also installed Temporary Modification TM-97-046, on October 29, 1997, to add screens to the top of the vent pipes during the inspection. These screens would prevent debris from entering the ECCS pump suctions in the event of an acciden It appeared that the requirements of 10 CFR Part 50, Appendix 8, Criterion Ill, "Design Control,"

were not met in this instance in that the design basis of the containment sump to exclude debris from the ECCS pump suction piping was not fully implemented. The team identified this item as part of Unresolved Item 50-255/97-201-1 The team also observed several piping penetrations between the east and west ESG rooms which included rubber piping expansion joints used as penetration seals. The team questioned the design of these piping penetration seals. The licensee stated that the engineering analyses that demonstrated that these penetrations met the design basis did not specifically address the use of rubber piping expansion joints in the penetration seals. The team reviewed EA-RJC-92-0508,

"Analysis of the Effect of a Fire on the Fire Barrier Penetration Seal Number FZ-0508," Revision O, and verified that the rubber piping expansion joints were not addressed. The licensee initiated CR C-PAL-97-1627 and determined that the failure to specifically justify the presence of rubber expansion joints did not invalidate the conclusions of the original engineering analyses and that the penetration seals were adequate. The licensee also stated that the affected documentation would be corrected, and that an "extent of conditionn review would be performed. The team identified this item as Inspection Followup Item 50-255/97-201-1 E1.3.1.3 Conclusions The team found that the mechanical portion of the SI system was capable of providing adequate emergency core cooling flow under accident conditions. The SI system performance information presented in the FSAR was consistent with the system design and operations documents.

The team identified a weakness in the scope of the IST program in that several SI system valves were not tested in accordance with ASME Section XI to verify their capability to perform the required safety functions as required by 10 CFR 50.55a(f). -

The team identified a weakness in the control of scaffolding and other items located in the vicinity of safety-related equipment, in that the plant failed to follow existing procedures regarding location and evaluation of these item The team identified a weakness in control of engineering analyses, in that several analyses were not updated as a result of changes to analytical inputs in accordance with plant procedure E1.3.2 Electrical Design Review E1.3.2.1 Scope The team evaluated the ability of the electrical loads required for the SI and interfacing systems to perform their functions under normal and accident conditions. This evaluation addressed ac bus loading, de battery loading and distribution, protective coordination, relaying, and modification E1.3.2.2 Inspection Findings The team reviewed EA-ELEC-LDTAB-005, "Emergency Diesel Generator 1-1 & 1-2 Steady State Loading,* Revision 4, and EA-ELEC-VOLT-13, "Palisades Loss of Coolant Accident With Offsite Power Available,* Revision 1, and determined that incorrect motor loadings were utilized for the LPSI pump motors. The licensee had not utilized the most conservative pump flows within the electrical analyses. During the inspection, the licensee determined that adequate electrical capacity was available to operate the LPSI and HPSI pump motors within their design parameters. This discrepancy is discussed in more detail in Section E1.4, "Electrical Interface Systems,* of this repor The LPSI and HPSI pump motors were sized to accelerate the SI pumps within the required time periods and to drive them as required for long-term continuous operation. The team reviewed the Class 1 E 125-V de battery and 120-V ac preferred bus loadings for the SI loads and determined that adequate capacity was available for the required loads. Protective relaying for the 2400-V'ac motors was determined to be properly set and calibrated so as to protect the SI loads and the electrical syste E1.3.2.3 Conclusions The electrical design for equipment that performed the normal and accident functions of the SI system supported the design-basis functions of the system. The electrical system provided redundant and safety-related power to the SI load The team identified discrepancies concerning the electrical design related to the SI system. These discrepancies are discussed in Section E1.4, "Electrical Interface Systems,n of this repor E1.3.3 Instrumentation and Controls Review E1.3.3.1 Scope of Review The team evaluated the ability of the instrumentation and controls for the SI system to perform the design safety functions. The team reviewed sections of the FSAR, applicable TS sections, DBDs, P&IDs, design modifications, plant procedures, and calibration data and also conducted system walkdowns. An in-depth evaluation was performed of the SIRwr level instrumentation, SIT

instrumentation, containment sump level instrumentation, SI actuation circuitry, sequencer logic, the licensee's commitment to RG 1.97, ulnstrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident,* environmental qualificationanalyses, and redundancy and independence of instrumentatio E1.3.3.2 Inspection Findings E1.3.3.2(a) Uncertainty Calculations The team reviewed 10 SI system calculations and 1 pressurizer pressure uncertainty calculation; these were identified as ubasis documents.*

Basis Document Rl-38, uSIRW Tank Level Instrument Calibration,* Revision 6, was reviewed for adequacy. It provided the basis for calibration of SIRWT level indicators LT-0332A and LT-0332B to enable their use to monitor the TS requirement that the tank contain at least 250,000 gallons of borated water. Rl-38 used a tank boron concentration of 1720 parts per million (ppm) and did not consider the range of 1720 to 2500 ppm allowed by TS Section 3.3. Rl-38 was the basis document for the calibration of the level indicator that supported manual actuation of post-accident recirculation operation. The team was concerned that the increased density of the tank water at higher boron concentrations would increase the instrument uncertainty. The calculation also did not account for variation in boron concentration density caused by temperature changes; an effect which could also affect the total uncertainty. The licensee recalculated the total instrument uncertainty using the most conservative boron concentrations and temperature, and the resulting change to the total uncertainty remained bounded by the original uncertainty valu Bases Document Rl-69, usubcooled Margin Monitor Surveillance," Revision 6, was reviewed for adequacy. The subcooled margin monitor (SMM) provided the operator indication of the PCS margin to saturation conditions. Rl-69 evaluated possible errors induced in the SMM. The team found that Rl-69 did not account for seismic uncertainty.. This was inconsistent with RG 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident," May 1983. This RG identifies subcooled margin as a Category I, Type A variable, which must continue to read within the required accuracy following, but not necessarily during, a safe-shutdown earthquake event. The team was concerned that the calculated error was nonconservative because it did not consider seismic uncertainty, and could provide misleading information to the operators. The licensee reanalyzed the potential error in the SMM, including seismic uncertainty, and the resulting total uncertainty remained bounded by the original uncertainty value. The licensee assigned Procedure Change Request (PCR) 5569 to revise Rl-6 EA-RSW-94-001, "Fl-0404 Instrumentation Uncertainty Calculation," Revision 2, was also reviewed for adequacy. The analysis established the recommended uncertainties of Fl-0404, which was used in flow testing of the SI pumps. The instrument was installed in 1989, and.has been calibrated five times since then. Drift error was determined using historical calibration data. For the first 4 years, the instrument was calibrated once a year. The team found that 24 months had transpired between the fourth and fifth calibrations. The licensee stated that the interval was changed in 1993 from 11 months to 24 months. The team asked if the drift analysis was revised to account for this change in the calibration interval. The team was concerned that increasing the calibration interval to 24 months would increase the drift error and consequently increase the total uncertainty of the instrument. The licensee reanalyzed the Fl-0404 uncertainty using appropriate drift performance data for the longer calibration interval, and the resulting change to the total uncertainty remained bounded by the original uncertainty value. The licensee issued EAR-97-0658 to revise EA-RSW-94-00 The team also reviewed Basis Document Rl-15A, "Safety Injection Tank Pressure Channel Calibration,* Revision 7, for adequacy. Rl-15A formed the bases for the pressure channel setpoints for PIA-0363, 0367, 0369, and 0371, which defined low-and high-pressure alarms for the SITs. The low-pressufe~alarms warned the operators of decreasing nitrogen pressure in the tanks. The channel alarms were set to annunciate eartier than the pressure limits of TS Section 3.3.1 (b) so appropriate action could be taken before pressure reached the setpoints of pressure switches PS-03408, 03448, 03738, and 03508, which were set to alarm at the TS limits. The team was concerned that Rl-15A did not conside*r uncertainties such as stability and temperature effects and that the current total uncertainty was not adequate. Considering the low alarm point of 207 psig, the calculated uncertainty allowance of+/- 6.85 psig could result in an alarm at close to 200 psig, which was the TS limit. If additional uncertainties were added, the channel pressure switches could alarm after the TS pressure switches. The licensee reanalyzed the setpoint for PIA-0363, 0367, 0369, and 0371 using additional appropriate uncertainty inputs and determined that the resulting instrument uncertainty was bounded by Rl-15 * The team observed that the results of these basis documents were determined to encompass specific additional uncertainties due to the assumed margins used in the documents to account for unquantified effects. The licensee had a guide entitled "Design & Maintenance Guide on Instrument Setpoint Methodology," EGAD-PROJ-16, Revision 0, and the team concluded that it provided a satisfactory methodology for setpoint calculations and was consistent with industry standard ISA-867-04, Part I, "Setpoints for Nuclear Safety-Related Instrumentation." The licensee stated that EGAD-PROJ-16 provided identical guidance as EGAD-PROJ-08, Revision 0, of the same title, which was the current designation of the guide. The instruments that were re-analyzed during the inspection used the guidance of EGAD-PROJ-08. This methodology affirmed that margins remained bounded. The licensee stated that use of this guide was not required by plant procedures. * However, the licensee has previously recognized from past assessments that its basis documents were not as rigorous as required by the current ISA standards. The licensee stated that EGAD-PROJ-08 was being revised and that the appropriate procedures would be revised to require its use. The team identified this item as Inspection Followup Item 50-255/97-201-1 The team reviewed the remaining calculations and found them technically adequat E1.3.3.2(b) Single Failure and Redundancy Review The team evaluated the design of each SI instrument loop for single-failure vulnerability, redundancy, and independence and found that all the designs reviewed were adequate and consistent with the licensing base E1.3.3.2(c) SI Actuation Logic Review The team evaluated two SI actuation logic circuits for consistency with the design bases, the containment high pressure signals and the pressurizer low-low pressure trip signals, along with their associated coincident logic and initiating circuits. The team found that these circuits would perform their safety function as designed. Furthermore, the team found the circuits which initiated SI concurrently with a loss of offsite power were consistent with the licensing base E1.3.3.2(d) SI System Modifications Review Modification FC-737 replaced the original mechanical sequencers with programmable logic controllers. The team reviewed the safety evaluation for the modification and was concerned about the apparent lack of documentation for the conclusion that an unreviewed safety question (USQ) was

• not created by the modification. Although the supporting documentation was not organized and summarized sufficiently, the licensee demonstrated to the team that documentation existed to support that no USQ resulted from modification FC-73 E1.3.3.2 (e) SI Walkdown During a walkdown of the SI system, the team observed that transmitters for containment spray flow, FT-0301 and FT-0302, and shutdown cooling heat exchanger flow, FT-0306, were properly mounted below their flow elements, but the process tubing was observed to be inadequately sloped back to the transmitters. Additionally, a walkdown performed by the licensee at the team's request during an in-containment inspection revealed that the process lines to the HPSI cold-leg flow transmitters FT-0308, FT-0310, FT-0312, and FT-0313, and the LPSI flow transmitters, FT-0307, FT-0309, FT-0311,.

and FT-0314, were also installed with inadequate slope. The team was concerned that inadequate slope in instrument tubing could contribute to significant instrument uncertainty by entraining unequal amounts of air in either leg of the transmitter, causing erroneous readings. This was shown to be a valid concern when an operator observed an erroneous reading in the left channel containment spray loop indicator, Fl-0301A. The "below zero* reading was caused by air trapped in one of the process lines. The licensee issued CR C-PAL-97-1561 to vent the lin The lack of tubing slope was inconsistent with original plant installation specification J-F020, Revision 0. This specification stated: "Flow instruments (differential type) in liquid and condensable vapor service shall preferably be mounted below the main line connection so that the. impulse lines will slope down to the instrument." The specification also stated: "Impulse lines to flow instruments shall slope (up or down) a minimum of one inch per foot." Plant drawings J-F133; Revision 1; J-F134, Revision O; J-F140, Revision O; and J-F141, Revision 0, depict various acceptable installation configurations for a differential transmitter. The current installations of the flow instruments identified above were not consistent with these drawings. A later specification, J-465 (Q), "The Technical Specification for Installation of Instrumentation for Nuclear Service for CPCo Palisades,* Revision 0, dated 1981, stated: "The installation shall be neat in appearance, properly supported, and shall provide for proper slope for adequate drainage or venting of the instrument lines." This specification has since been incorporated into specification 20557-J-59 (Q) under the same title, which requires that a "horizontal tubing run is continually sloped in accordance with design drawings." The licensee issued CR C-PAL-97-1561 to evaluate these instrument tubing sloping discrepancies. According to the operability determination of the CR, the instruments have never shown any adverse effects of trapped air during the last 20 years of operatio The HPSI and LPSI flow transmitters were mounted as much as 8 ft above their flow elements. To accommodate instruments mounted above flow elements, specification J-F020 stated: "5 foot minimum "drop legs (equivalent of a loop sea1r may be required before the tubing is sloped up the meter." Plant drawings J-F152, Revision 1, and J-F153,.Revision 0, depict these mounting configurations. The licensee stated that the bottom and side tap locations for the tubing would tend to limit the amount of air getting into the transmitters and that air entrainment would be minimal due to the ratio of the volume of the HPSI and LPSI pump suction piping to the tubing volume. EA-C-PAL-95-0877D, "Evaluation of the Potential for Excessive Air Entrainment Caused by Vortexing SIRWT During a LOCA," Revision 0, evaluated the potential for excessive air entrainment in the lines of the pumps caused by vcrtexing in the SIRWT during a LOCA, and determined that the air entrainment would be a small percentage of the flow volume. The licensee also stated that technicians are required to vent the transmitters during every 18 month surveillance. However, the team was concerned that, since the transmitters sense low static pressure during normal standby operation, air may accumulate between calibration intervals and between system tests. Additionally, the water circulated through the SI lines from the containment sump could contain significant amounts of dissolved gasses, which could enter the tubing up to the flow transmitters. The team

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was concerned that the effed of air entrapped in the instrument tubing could cause large and unquantifiable errors in the flow indication EOP Supplement 4, "loss of Coolant Accident Recovery Safety Fundion Status Check Sheet,"

contained curves presenting total SI flow ranges intended to help ensure that the minimum values utilized in the accident analyses (LOCA, MSLB, steam generator tube rupture (SGTR}) were me There was also a minimum total flow criterion for the operators to meet, which ensured the containment sump check valves remained in a stable condition in EOP-4.0, "loss of Coolant Accident Recovery,* Revision 9. The operators would use the HPSI and LPSI flow indication from FT-0308, 0310, 0312, 0313, 0307, 0309, 0311, and 0314 to compare SI system performance against the EOP requirements. The team was concerned that the potentially large errors could confuse the operator and impair decision making. The licensee stated that the operators are trained to use all available indications and that alternate/additional instrumentation could be used to confirm trending of PCS conditions such as that for pressurizer level, subcooling margin, reador vessel level, and charging pump flows. The licensee issued EAR-97-0699 to evaluate this ite It appeared. that the design basis for instrument tubing installation was not implemented in the plant installation as required by 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control." The team identified this item as Unresolved Item 50-255/97-201-1 E1.3.3.2(f) Equipment Qualification The team reviewed EA-GAW-89-EQ-01, "Instrument Loop Error Evaluation for EEQ Listed Transmitter Loops," Revision 1, for technical adequacy and consistency with the SI instrumentation-basis documents. The analysis provided a basis for instrument loop errors caused by the harsh environment of a design-basis accident (OBA). It provided loop error calculation methodology and error magnitude The team reviewed the methodology used in the analysis of one instrument (FT-0308, HPSI Flow)

and how it was applied to other basis documents. The team found that all the appropriate source inputs were used and the methodology was adequate and consistent with industry standards. The analysis was technically adequate and the various calculated uncertainties for OBA conditions were appropriately included in other basis document E1.3.3.3 Conclusion The team found the instrumentation and controls portion of the SI system and interfacing SI actuation was capable of providing adequate control and monitoring. The SI system initiation logic was capable of performing its design safety fundion. The team identified instrument tubing slope deficiencies which could potentially cause inaccurate indication E1.4 Eledrical Interface Systems E1.4.1 Scope of Review In this portion 0f the design review, the team evaluated the EOG static loading calculation, ac and de system calculations, electrical plant modifications, and testing. In particular, the team focused its evaluation on assessing the consistency between the electrical systems and their design and licensing base *

E1.4.2 Inspection Findings E1.4.2(a) AC System The team reviewed EA-ELEC-LDTAB-005, "Emergency Diesel Generator 1-1 & 1-2 Steady State Loading,* Revision 4, and verified that the analysis was consistent with the design-basis information in the FSAR. All required accident loads for a LOCA and a LOOP were identified and tabulated. The electrical loads exceeded the continuous rating of the EOG during the first 32 minutes of operation but were below the EOG maximum 2-hour rating. One of the inputs to this analysis was the electrical load estimate for LPSI pumps P-67 A and P-678. These electrical load estimates were based on the minimum hydraulic LPSI pump performance used in EA-A-PAL-92-037, *Emergency Diesel Generator Loadings-First Two Hours,* Revision 1, which determined that LPSI pump flow would be 3600 gpm. Although the LPSI pump flow was conservative for evaluating LOCA mitigation, it was not conservative for determining the maximum load the EOG could experience during a LOC The team determined that the LPSI pumps could pump 4500 gpm with one LPSI pump discharging into all four injection loops as identified in EA-SDW-95-001, "Generation of Minimum and Maximum HPSl/LPSI System Performance Curves Using Pipe-Flo,* Revision 2. The team was concerned that the licensee had not analyzed for the worst-case electrical load demand on the EDGs. Preliminary evaluations by the licensee using the correct maximum loads indicated that the electrical loading on one EOG could be higher than that determined in EA-ELEC-LDTAB-005. The licensee issued CR C-PAL-97-1650 to review and correct all necessary electrical analyses and determined the EDGs to be operabl The team reviewed EA-ELEC-VOLT-13, "Palisades Loss of Coolant Accident With Offsite Power Available,n Revision 0, which evaluated the ac voltage available during normal operating, refueling, and accident conditions. The team noted that the calculation had not been revised since 1993 and that the load magnitudes identified in EA-ELEC-LDTAB-005, Revision 4, and EA-SDW-95-001, Revision 2, had not been included. The licensee reviewed the impact of the revised loads on EA-ELEC-VOL T-13 and determined that the changes had minimal effect on the analysis. The team also noted that FSAR Section 8.3 stated that backfeeding via*the main and station power transformers could be utilized; however, EA-ELEC-VOLT-13 had not anal}'zed this particular operating mode. The licensee stated that it had recognized that an analysis for backfeeding needed to be performed in 1994 and had issued AIR A-PAL-94-223 to create an analysis in orderto bound this condition of operation. The licensee initiated C-PAL-97-1619 to review and update EA-ELEC-VOLT-13 for load change It appeared that the requirements of 10 CFR Part 50, Appendix 8, Criterion Ill, "Design Control," had not been met for EA-ELEC-LDTAB-005 and EA-ELEC-VOLT-13 in that the design basis had not been updated to document the actual plant parameters. The team identified this item as part of Unresolved Item 50-255/97-201-1 FSAR Section 8.5.2 stated that cables would be sized in accordance with the National Electric Code (NEC) or Insulated Power Cable Engineers Association (IPCEA/ICEA) ampacity values and the cable ampacities would be adjusted on the basis of actual field conditions when possible. The adjustments included conductor operating temperature, ambient temperature, cable overall diameter, raceway fill, and fire stops. The licensee had recently initiated a program to verify the adequacy of its cable ampacity sizing. EA-ELEC-AMP-032, "Ampacity Evaluation for Open Air Cable Trays With a Percent Fill Greater Than 30% of the Usable Cross Sectional Area, n Revision 1, was issued in 1997 to address cable sizing. While reviewing the EA, the team noted the absence of fire stop derating and increased cable temperatures due to thermal radiation from hot pipes. The licensee had initiated AIR A-PAL-97-062 to evaluate the effects of local heat sources on fire stops; however, evaluation of the effects on cable degradation due to the close proximity of hot piping systems had

not been included. The licensee stated that evaluation of the effects of hot piping would be included under A-PAL-97-062. The team Identified this as Inspection Followup Item 50-255/97-201-1 The 120-V ac.safety-related and non-safety-related loads were powered from instrument ac bus Y~01. Bus Y-01 was powered from either motor control center {MCC) 1 or 2 via automatic transfer switch Y-50. MCCs 1 and 2 were redundant safety-related busses. The licensee stated in a January 24, 1978, letter to the NRC that it would implement the recommendation of RG 1.6 in that no provision would exist for automatically transferring loads between redundant power sources. The NRC issued a safety evaluation report, dated April 7, 1978, confirming the licensee's commitmen FC 364, "Feeder Change for Instrument Bus Y-01," Revision 0, implemented this commitment and powered bus Y-01 from MCC 1 and non-safety-related MCC 3. However, FC 854, "Y-01 Power Supply Feed Modification,* Revision 0, moved the backup power source from MCC 3 to the safety-related MCC 2, and resulted in a departure from the plant's licensing basis. The modification installed fuses in series with the existing breakers, which provided an additional level of protection for the two safety-related busses. The team observed that the safety evaluation performed for FC 854 did not identify that prior NRC approval was required. The licensee issued CR C-PAL-97-1678 to document this deviation from the licensing basis. It appeared that this modification was a USO in that the possibility of a common-mode failure of the redundant safety-related busses was created, which was not previously evaluated in the FSAR and, thus, the criterion of 10 CFR 50.59(a)(2){ii) was satisfied. The team identified this item as Unresolved Item 50-255/97-201-1 The team observed that no system analysis existed to show that all the Class 1 E 120-V ac loads had adequate voltages. The licensee demonstrated during the inspection that adequate voltages did exist for selected loads. For example, EA-ELEC-VOL T-24, "Voltage Drop From Preferred AC Power Source Y10 Breaker 2 and Y40 Breaker 2 Out to the 5U12 Relays," Revision 0, showed that adequate ac voltage for those selected components was available at the minimum inverter voltag The licensee initiated CR C-PAL-97-1621 to evaluate and resolve this concern. The team identified this item as part of Inspection Followup Item 50-255/97-201-1 The team reviewed relay settings for protective relays associated with LPSI pump P-67 A, HPSI *

pump P-66A, SW pump P-7A, CCW pump P-52A, EOG 1-1 differential protection, bus 1C undervoltage protection, and Bus 1C second-level undervoltage protection. The settings were consistent with the design parameters of the devices being protected. However, during the review, the licensee determined that the overcurrent relays for supply breakers 152-105 and 152-106 to bus 1C had not been calibration tested during the last refueling outage {1995) as required by Periodic and Predetermined Activity (PPAC) $PS025, "Bus 1C Relay Testing." The licensee stated that these relays would be calibrated during the 1998 refueling outage. The licensee reviewed past calibration data for this type of relay and determined that negligible drift had previously been documented. The licensee initiated CR C-PAL-97-1568 to resolve this discrepancy. It appeared that the requirements of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," had not been implemented in this case in that certain relays had not been tested as required by the test program. The team identified this item as Unresolved Item 50-255/97-201-1 The team questioned the replacement schedule for Agastat E7000 series relays. The team was aware that the manufacturer, in correspondence to other utilities, had recommended a 10-year replacement schedule for these relays. The licensee stated that 52 E7000 series relays were installed and that 7000 series Agastats were also installed in Class 1 E applications. Some circuits containing 7000 series relays included the 2400-V bus 1 C and 1 D supply breakers, time delay relays associated with charging pumps P-55A, B, and C, and auto transfer failure alarms for 2400-V busses 1 C and 1 D. The manufacturer's stated qualified life for the E7000 relays was 10 years. The licensee stated that the 10-year qualified life applied if the relays were located in a harsh

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environment and, since the E7000 relays were located in a mild environment, no qualified life determination was required. Based upon this justification, the licensee issued PPAC Deletion Form MSE 034, dated Marcil 3, 1995, which stated that the relays-would not require replacement at 10-year intervals. The team believed that the qualified life stated by the manufacturer applied to any environment. The team verified with the manufacturer that the projected qualified life of 1 O years was the operating life of the E7000 series relay as long as the device did not exceed the equipment ratings, and that the life of 10 years was applicable to either a mild or harsh environment. The licensee had not evaluated the qualified life of the 7000 series relay The manufacturer of Agastat relays issued a 10 CFR Part 21 notification concerning the inability of the E7000 series relays to switch a 1-amp load at rated voltage. The licensee evaluated the installed E7000 series relays and identified no concerns. The team observed that this evaluation did not review those 7000 series relays dedicated by the licensee to safety-related us The licensee issued CR C-PAL-97-1663 to resolve the issues concerning Agastat relays and determined that all the relays were operable. It appeared that the requirements of 10 CFR Part 50, Appendix 8, Criterion Ill, "Design Control,* had not been met in this instance in that the design-basis lifetime for Agastat relays as stated by the manufacturer had not been correctly implemented in the facility. The team identified this item as Unresolved Item 50-255/97-201-1 E1.4.2(b) DC System The 125-V de system was divided into two independent systems. Each system consisted of a battery, switchgear, distribution panel, and two chargers. Station battery 1, battery charger 1, and battery charger 3 supplied 125-V de bus 1. Battery charger 1 was supplied from MCC 1 and battery charger 3 was supplied from MCC 2. Administrative controls limited the operation so that only one charger per battery was in service. This prevented a common-mode failure from affecting both emergency busses. The supply to 125-V de bus 2 was similar, with battery charger 2 fed from MCC 2 and battery charger 4 fed from MCC 1. Operating Procedure SOP-30, "Station Power,"

Revision 20, required the battery chargers to be operated in pairs (1 and 2 or 3 and 4). The licensee stated that the battery chargers were swapped monthly to provide equal operating time for each battery charger. During swapping of the battery chargers in accordance with Section 7.7.2 of SOP30, the 125-V de breaker on the in-service battery charger was opened and then the 125-V de breaker for the battery charger to be placed in service was closed.. During this evolution, both battery chargers were disconnected from the station battery and 125-V de switchgear bus. Although temporary disconnecting the battery charger from the de bus had minimal safety impact on the plant, the team observed that TS 3. 7.1 h required two station batteries and the de systems (including at least one battery charger on each bus) to be operable when the PCS was above 300 °F. The licensee stated that an LCO was not entered when no battery chargers were connected to the de busses. The licensee initiated CR C-PAL-97-1537 to resolve this discrepancy. The team identified the licensee's failure to enter an LCO during battery charger switching evolution as Unresolved Item 50-255/97-201-2 The team reviewed the 125-V de battery loading during the normal and alternate battery charger alignment. During the normal battery charger alignment, battery charger 1 was powered from EDG 1-1 and battery charger 2 was powered from EDG 1-2. During a LOCA combined with a LOOP in this normal alignment, the batteries would be without ac power for approximately 10 seconds until the EDGs restored power. The team reviewed EA-ELEC-LDTAB-009, "Battery Sizing for the Palisades Class 1 E Station Batteries ED-01 and ED-02, * Revision 2, which verified that the battery was.sized to provide adequate power during the.10-second interval until the EDGs provided ac power to battery chargers 1 and 2. During the alternate battery charger alignment with battery charger 3 powered from EDG 1-2 and battery charger 4 powered from EDG 1-1, the station batteries

• would be required to carry the de loads for more than 10 seconds in the event of a LOCA combined with a LOOP and a single failure of ac power. EA-ELEC-LDT AB-009 did not analyze the battery loading for station batteries ED-01 and ED-02 during this condition. When questioned by the team, the licensee stated that the de loading during this scenario would be greater than the worst-case loading assumed in EA-ELEC-LDTAB-009. The licensee issued CR C-PAL-97-1596 to resolve this discrepancy. Additionally, the team had concerns on whether the licensee met the single failure criterion when the alternate battery charger alignment was in effect. The team identified the question with respect to the single failure criterion and the additional loading on the battery as an Inspection Followup Item 50-255/97-201-2 The team identified that TS Section 4.7.2c required that each station battery be demonstrated operable by verifying that the battery capacity was adequate to supply and maintain in an operable status all of the actual emergency loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the battery was subjected to a battery service test. The battery service tests performed on station batteries ED-01 and ED-02 were performed for a duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4-hour duration and loading was based on the design-basis station blackout {SBO) coping time. The team noted that the 2-hour requirement of TS 4. 7.2c was non-conservative with respect to the design basis, which required the station batteries to be available for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The design-basis duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> was included in FSAR section 8.4.2; DBD 4.01, "Station Batteries, 0 Revision 3; RE-83A, "Service/Modified Performance Test-Battery N E0-01: Revision 9, and RE-83B, "Service/Modified Performance Test-Battery No. ED-02,"

Revision 9. Testing the batteries in accordance with RE-83A and B has ensured that batteries ED-01 and 02 have met the 4-hour design-basis requirement. The licensee has submitted TS changes to correct the non-conservative TS Section 4.7.2c and issued CR C-PAL-97-1551 to resolve this discrepancy. The team identified this item as Inspection Followup Item 50-255/97-201-2 EA-ELEC-FL T-005, "Short Circuit for the Palisades Class 1 E Station Batteries ED-01 and ED-02,"

Revision 0, was submitted to the team as the short-circuit analysis for the Class 1 E 125-V de system. The following discrepancies with the assumptions, methodology, and conclusions were identified:

Section 4.4 and 4.5 assumed various breaker and fuse impedances, which had not been verified against the installed facilit *

Section 5.2 utilized the battery charger current limit of 220 amps as the maximum short-circuit contribution without supporting documentatio *

Section 5.2 stated that the open-circuit voltage was 2.06 V per cell, whereas the EA utilized an open-circuit voltage of 2.0 V per cel *

Section 8.0 stated that the results were to be further reviewed by the licensee; however, the team found no evidence of this review. Section 8.0 also contained no conclusion about the de system acceptabilit The licensee issued AIRs A-PAL-97-108, 109, and 110 to resolve these discrepancies. The licensee stated that the analyses would be reviewed and the conclusions revise During the 1995 refueling outage, FES95-206 replaced existing batteries ED-01 and ED-02. The team questioned if the short-circuit current provided by the new battery was analyzed and if there were any effects on the de distribution panel breakers, since the team noted that EA-ELEC-FL T-005 had not been updated since 1994. The team also noted that the design basis for the evaluation of fault current contributions on de circuits was in FSAR Section 8.5.2, which stated "The 125 volt de

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protection design considers the fault current available at the source side of the feeder protective device.* However, the licensee stated that the short-circuit contribution value for de circuits was taken at the electrical load tenninals and not at the breaker load tenninals {de short-circuit current value would-be less when calculated at the load tenninal vice the source side of the feeder protection device because voltage available at the load tenninal would be less than at the source breaker). The licensee detennined that the short-circuit contribution at 8 breakers {breakers 72-1O1, 72-105, 72-106, 72-121, 72-127, 72-133, and 72-135) on distribution panels D11-1 and D11-2 could exceed the short circuit interrupting ratings when evaluated in accordance with the design basis method in the FSAR. Also, when the team questioned the assumed breaker fault ratings on de busses D10, D20, D11-1, and D11-2 of 13,000 amps in EA-ELEC-FLT-005, the licensee was unable

_to show manufacturer-or testing documentation to support this assumption. The team believed that

• this assumption-was inconsistent with its experience. The licensee perfonned an operability review and issued CR C-PAL-97-1652 to resolve these discrepancie The maximum short-circuit current of the battery installed by FES95-206, as provided by the manufacturer, was 17,094 amps. Calculation EA-ELEC-FLT-005 did not reflect this new short-circuit current. Upon questioning by the team, the licensee stated that an evaluation was perfonned to

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ensure that the system short circuits were acceptable. During the team's review of this evaluation it was detennined that the maximum battery short circuit current was not utilized. The licensee stated that the short-circuit current utilized, 12,821 amps, was provided by the manufacturer as a more

. realistic value than 17,094 amps. However, the licensee could not document a basis for the 12,821 amps and stated that they would verify it with t_he manufacture The team identified these discrepancies concerning EA-ELEC-FL T --005 as part of Inspection Followup Item 50-255/97-201-2 FSAR Section 8.4.3.3 stated that the batteries were designed to furnish their maximum load down to an operating temperature of 70 °F without dropping below 105 V de, and that the equipment supplied by the batteries was capable of oper~ting satisfactorily at this voltage rating. EA-ELEC-VOL T-026,

"Voltage Drop Model of the Palisades Class 1 E Station Batteries 001 and D02," Revision 0, evaluated the de voltages at the distribution panels based upon a battery voltage of 105 V de, but did not evaluate the voltages that would be available at the load device tenninals. The team was concerned that the additional voltage drop from the distribution panel to the loads could result in voltages less than the design basis of the loads, and that no analysis was perfonned to evaluate this situation. For example, the design-basis minimum input voltage for the inverters was 105 V de and the licensee could not show any vendor documentation to support operating at a value less than 105 V de. The team noted that the inverters could be subjected to an input voltage of approximately 102 V de if the battery voltage were 105Vdc. The licensee stated that battery surveillance testing has shown that battery voltage, when subjected to an SBO duty cycle, did not decrease below 108 V de. During the inspection, the licensee evaluated several safety-related loads and verified that adequate voltages would exist at 105 V de battery voltage. The licensee issued CR C-PAL-97-1620 to evaluate the lack of an EA to ensure that adequate voltages would exist at the load terminals. The team identified this item as part of Inspection Followup Item 50-255/97-201-2 The team also questioned the capability of solenoid valves to operate at voltages of 87 V de as stated in DBD 1.01, "Component Cooling Water System,* Revision 4. The licensee determined that the_DBD was incorrectly worded and that the correct solenoid capability was90-140 V de. Upon further review, the licensee identified that improper1y rated coils, rated 102-126 V de, were installed in solenoid valves SV-09.18 and SV-09778. The licensee initiated Engineering Assistance Request (EAR) 97-0652 to replace the coils. It appeared that the requirements of 10 CFR Pat 50, Appendix B, Criterion Ill, "Design Control," were not followed in that the design basis for the solenoid

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valve coils was not implemented in the plant. The team identified this item as Unresolved Item 50-255/97-201-2 The team-identified other discrepancies in calculations as follows:

Assumptions 4.6 and 4.7 of EA-ELEC-VOLT-26, Revision 0, and assumptions 4.8 and 4.9 of EA-ELEC-MISC-022, *Electrical Systems Model of the Palisades Class 1 E Safety Related 125Vdc System,* Revision 1, assumed various fuse and breaker impedances which had not been verified against the installed equipmen *

Section 7.0 of EA-ELEC-VOLT-26, Revision 0, "Conclusion,* stated that the results were to be further reviewed by the licensee; however, the team found no indication that this review had been performed. The "Conclusion" section also contained no statement concerning the de system acceptabilit *

EA-ELEC-VOLT-26, Revision 0, utilized a correction factor for battery temperature of 77 °F instead of the correction factor for 70 °F, which was the minimum design-basis temperature for the battery. The number utilized is less conservative and the licensee evaluated that the overall effect on voltages in the calculation would be less than 0.5 percen *

EA-ELEC-LDTAB-029, Revision 2, stated the type of battery constant as 1.0 in Attachment A and 1.4 on Sheet 4. The constant to be utilized depended on the type of battery. 1.0 referred to a lead acid battery; 1.4 referred to a nickel-cadmium battery. The licensee reviewed the EA and determined that the correct constant was utilized in the EA and that the reference to 1.4 was an editorial erro The licensee issued CR C-PAL-97-1656 to address the battery temperature correction factor and stated that the other discrepancies would be corrected in future revisions to the calculations. The team identified this item as part of Inspection Followup Item 50-255/97-201-2 E1.4.2(c) Testing The team noted that TS Section 4. 7.1.b required testing to be performed at every refueling to demonstrate the overall automatic operation of the emergency power system. Proper operation was verified by bus load shedding and automatic starting of selected motors and equipment to establish that emergency power had been restored within 30 seconds. FSAR Tables 8-6 and 8-7 stated that sequencing would occur in 65 seconds.* Technical Surveillance Procedure RT-BC, "Engineered Safeguards System - Left Channel," Revision 8, and RT-8D, "Engineered Safeguards System - Right Channel,* Revision 8, required performance testing to be within the 65-second requirement. The team questioned the use of a 30-second test duration in the TS instead of a 65-second duration, which would demonstrate that all required equipment would start. The licensee stated that the TS did not specifically require full testing of the entire diesel load sequence but only required testing of selected loads. The team noted that the licensee was testing the diesel loading to the full accident 0 loading sequence and has submitted a proposed TS change which would be more consistent with the.current desig The team reviewed Test Procedures R0-128-1, "Diesel Generator 1-1 24 Hour Load Run,"

Revision 2, and R0-128-2, "Diesel Generator 1-2 24 Hour Load Run," Revision 2. The team noted that Section 3.0 of the Acceptance Criteria and Operability Sheet for Procedure R0-128-2 referred to TS Sections 3.7.1 and 4.7.1.11, and that these references would only be correct when the proposed improved TS, which have been submitted to NRC for approval, became effective. The licensee

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issued CR C-PAL-97-1566 to resolve these discrepancies. The team identified this item as Inspection Followup Item 50-255/97-201-2 The team identified the following discrepancies when reviewing station battery Test Procedures RE-83A, uservice/Modified Performance Test-Battery No. ED-01,* Revision 9, and RE-838,

"Service/Modified Performance Test-Battery No. ED-02,* Revision 9:

The tests evaluated whether the final discharge voltage (105 V de) of station batteries ED-01 and 02 was met at the end of the test (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). Load parameters (amps) at 1 and 239 minutes were not verified during the test. These load parameters were design requirements of EA-ELEC-LDTAB-009, Revision 2. The licensee demonstrated that the 1-and 239-minute data were recorded elsewhere and that the duty cycle was tested in accordance with the design requirements. The licensee stated that the battery testing procedures would be revised to include verification of these design parameter *

The procedures did not require any calibration tolerances for the discharge testing shunt and control unit. The licensee stated that the tolerance was removed from the procedure before testing during the 1996 refueling outage and issued PCRs 5422 and 5423 to change the procedures to include these tolerance *

The battery charging data in Procedure RE-83B for the 1996 refueling outage did not meet Step 5.2.2, which required the battery charging rate to be decreasing and to remain within 5 percent over the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before stopping the equalization process, in that the process was stopped before the end of the 8-hour period. The licensee stated that the nearly steady-state voltage operation of the charger gave adequate assurance that the battery was operable before exiting the test and issued CR C-PAL-97-1460 to resolve this discrepanc *

During the performance of procedure RE-838 at the 1996 refueling outage, the elapsed time recorded manually did not agree with the testing control unit time. The licensee stated that because the testing unit did not have the capability to record the time, the test start and stop times were recorded manually. The inconsistencies were minor and had no effect on the test results. The licensee issued C-PAL-97-1460 to evaluate this discrepanc The team identified this item as Inspection Followup Item 50-255/97-201-2 E1.4.2(d) Modification Review The team reviewed the following electrical modification packages and found them consistent with the plant design basis:

Temporary Modification TM-96-027, "Install 152-Spare #5 Breaker in 152-113 Cubicle," dated April 10, 1996

FES-95-206, "ED-01 and ED-02 Station Battery Replacement," Revision 0

FC 364, "Feeder Change for Instrument Bus Y-01," Revision 0

FC 854, "Y-01 Power Supply Feed Modification," Revision O

FC 638, "Add Component Cooling Water Pumps to the Normal Shutdown Sequencer,"

Revision O

FC 798, "Battery Room Temperature Indication and Alarm," Revision O

FC 683, "Removal of Pressurizer Heaters From SIS Trip," Revision O Except as previously discussed, all these modifications were adequately prepared, provided the necessary technical basis for the changes, and contained adequate installation instructions and

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testing requirements. The 10 CFR50.59 safety evaluations were adequate, except for the two listed below:

Safety Review$ 95-1431 and 95-1432, dated July 7, 1995, for FES-95-206 stated that the batte-ry duty cycle service test duration for station batteries ED-01 and ED-02 was changed from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The licensee noted that TS Section 4.7.2.c was affected by this design change. However, the USQ evaluation, Question 2 of Section II, was not checked

"Yes" for a TS change. TS 4.7.2.c required that a 2-hour battery test be performed; while design analysis ELEC-LDTAB-009 and FSAR Section 8.4.2 required a 4-hour battery duty cycle. The licensee has submitted a proposed TS change to reflect the proper battery test duration and issued CR C-PAL-97-1551 to address this discrepanc *

The safety review documentation for TM-96-027 stated that the FSAR was not reviewe Administrative Procedure 3.07, "Safety Evaluations," page.12, required that the FSAR be reviewed and that those sections reviewed be noted on the safety review sheet. The licensee initiated C-PAL-97-1439 to evaluate this discrepanc The team identified these safety review discrepancies as Inspection Followup Item 50-255/97-201-2 E1.4.3 Conclusions The team concluded that adequate ac supply was available for both normal and accident condition The station battery de system was adequate to supply the required power during normal and accident condition *

The team identified a weakness in the control and performance of EAs in that several EAs were not updated to incorporate the latest information and several contained errors. There was also a lack of complete documentation to demonstrate that the design basis was correctly implemented for the de and ac system *

The team identified an instance where the licensee failed to recognize a potential USQ (modification which resulted in the capability for an automatic transfer between redundant safety-related busses),

and an example of licensee's failure to enter an LCO during a battery charger switching operatio E1.5 FSAR and Design Documentation Review E1.5.1 Scope of Review The team reviewed the FSAR, DBDs, and various drawings for consistency with the design and licensing basi E1.5.2 Inspection Findings The team identified the following discrepancies in the FSAR:

• Page 6.7-4 stated that containment isolation valves fail closed with loss of voltage or control air except for the CCW return isolation valves. However, the CCW supply isolation valve (CV-0910) is also a fail-open valve and should have been noted as an exception to fail-closed containment isolation valves. The licensee issued FSAR Change Request 6-142-R20-1426 to correct the FSA *'

Section 6:7 classified the CCW penetrations as Class C-2, which was defined as penetrations with lines not missile protected. However, EA-GW0-7793-01 stated that the entire CCW system (both inside and outside cont~inment) was missile protected. The licensee issued FSAR Change Request 6-143-R20-1427 to state that the CCW penetrations were not vulnerable to intemally generated missile *

Table 9-10 stated that valves 3029 and 3030, containment sump suction valves, failed closed upon loss of air and were equipped with an accumulator. The valves actually failed as is and had no accumulator. The licensee issued FSAR Change Request 9-293-R20-1431 to correct the FSAR and CR C-PAL-97-1559 to evaluate and trend the FSAR discrepancies being identified at the plan *

Table 9-9 correctly stated that the high-pressure air piping was seismic Class I from the receivers to the valve operators. However, FSAR Table 5.2-3 stated that the entire system was seismic Class I. The licensee issued FSAR Change Request 5-155-R20-1432 to correct the FSAR 5.2- *

Section 8.4.2.2 stated that the station batteries would be tested to Institute of Electrical and Electronics Engineers (IEEE) 450-1975. However, battery test.ing procedures RE-83A, Revision 9, and RE-838, Revision 9, referred to IEEE 450-1995. FSAR Change Request 8-126-R20-1249 had been initiated, but the licensee did not intend to act on this change until approval was received from NRC of a related proposed TS chang *

Table 5.7-8 listed the seismic design value for the station batteries and racks as "later" instead of including the actual values of the batteries installed by FES95-206. The licensee issued EAR 97-0636 to evaluate this discrepancy and revise the FSA *

Section 8.2.3 stated that "The de battery system is designed to supply the required shutdown loads, with a total loss of ac power, for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />." This statement did not reflect the fact that load stripping was required during the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the battery to perform its intended function during a loss of ac powe *

Section 8.3.5.2 stated that "Operation of all circuit breakers in the de and the preferred ac systems is manual with automatic trip for fault isolation." The battery shunt trip breakers and the de bus tie breakers do not comply with this statemen *

Section 8.3.5.3 stated that "Each of the two battery chargers provided on the de bus is capable of supplying the normal de loads on the bus and simultaneously recharging the battery in a reasonable time. A fully discharged battery can be recharged in less than nine hours." Contrary to the statement, one battery charger could not supply the normal loads and recharge a fully discharged battery in less than 9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> *

Section 8.4.2.2 stated that "Emergency Operation-On loss of normal and standby ac power, the batteries will supply power to all preferred ac and de loads, until one of the (diesel generators) DGs has started and can supply power for the chargers." This statement was not correct if the battery chargers were in their alternate alignment and did not reflect load shedding during the 4-hour duration,

Section 8.5.2 stated that "The power source for the driven equipment and the control power for that system are supplied from the sources in one channel." This statement would not be correct if the battery chargers were cross-connecte *

Section 8.5.3.2 referred to *System 1, 2, 3, 4 Circuits" and separation requirements fo~ those circuits. The licensee was not able to identify these circuit *

Section 8.4.1.3 required clarification as to whether the reserve capability margin referred to the capability of. the overall EOG and engine or if it referred to the capability of the EOG to handle an increased loading due to a control circuit malfunction during the loading sequenc The licensee issued C-PAL-97-1309 to resolve this discrepanc *

Section 6.1.2.3 stated that "The RAS... provides a permissive to manually close the valves in the pump minimum flow lines: EOP-4.0, "Loss of Coolant Accident Recovery," Revision 9, Step 23, directed the operators to place the hand switches for these valves in the pump minimum flow lines (CV-3027 and CV-3056) to *cLOSE when SIRWT level lowered to between 25 percent and 15 percent. Per EOP-4.0, Step 51, the RAS occurred when the SIRWT level reached 2 percent. The FSAR appeared to conflict with EOP-4.0. The licensee initiated FSAR Change Request 6-141-R20-1425 to update the FSA *

The footnote for Table 14.17.1-1 implied that a containment building temperature of 90 °F was used as input to the large-break LOCA analysis because it is the limiting temperature during normal operation. The 90 °F value did not appear to be limiting. The licensee stated that the 90 °F value was the nominal containment building temperature, not the limiting temperature, and was used in the accident analysis in accordance with Seimens Power Corporation's large-break LOCA methodology guidelines. The licensee initiated FSAR Change Request 14-95-R20-1441 to update the FSA The above discrepancies had not been corrected and the FSAR had not been updated to ensure that the material in the FSAR contained the latest material as required by 10 CFR 50.71(e). The team identified this item as Unresolved Item 50-255/97-201-3 The team identified the following discrepancies in the OBOs:

080-1.07, "Auxiliary Building HVAC Systems,* Revision 1, Table 3.2.1, incorrectly stated that the design-basis temperature for Room 123, which contains the CCW pumps, was 125 ° The correct temperature was 104 °F as stated in 080-7.01, "Electrical Equipment Qualification Program," Revision 1, Appendix A The 125 °F temperature was a conservative assumption used to size the outside air supply fans. Table 3.2.1 also contained a typographical error in a reference number. The licensee issued 080 Change Requests 1.07-71-R1-0512 and 1.07-72-R1-0532 to correct the 08 *

OBD-1.07, Revision 1, Section 3.2.1.3, listed maximum room temperatures for the west ESF room from an outdated analysis. The latest analysis, EA-D-PAL-93-272F-01, "Engineered Safeguards Room Heatup Following LOCA in Conjunction With a LOOP," Revision 0, determined lower maximum room temperatures for various SW flows through the air cooler The 080 also required clarification of the normal design temperature of the ESG room. The licensee issued 080 Change Request 1.07-73-R1-0543 to correct the 08 *

080-7.08, "Plant Protection Against Flooding," Revision 1, incorrectly stated that the EOG would be inoperable before a flood reached the EOG windings because the lube oil heaters were located below the windings at 7 inches above the floor. EA-C-PAL-95-1526-01,

"Internal Flooding Evaluation for Plant Areas Outside of Containment," Revision 0, stated that the minimum flood level at which the EOG could become inoperable was 1 O inches due to the exciter cubicle bus bars and that the lube oil heaters were not needed for EOG

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operability. The licensee issued CR C-PAL-97-1557 to initiate a DBD change and evaluate the ite *

DBD-2.03, "Containment Spray System,* Revision 2, stated that the air supply to the sump-outlet valves, CV-3029 and 3030, was backed by an accumulator. There were no accumulators for these valves. The licensee identified this error while evaluating an FSAR statement that these valves had an accumulator backup that was questioned by the team, and issued DBD Change Request 2.03-22-R2-0531 to correct the DB *

DBD 1.01, "Component Cooling Water System," Revision 3, Section 3.3.7, incorrectly indicated that Class 1 E and non-Class 1 E breakers were installed in the same distribution panels. The licensee initiated DBD Change Request 1.01-14-R3-0518 to correct the DB Section 3.3. 7 of this DBD also stated that solenoid valves had been tested to operate at 87 V de instead of 90 V de. The licensee stated that the DBD would be correcte *

During the team's review of FES95-206, it was noted that the battery manufacturer had imposed a limit of 40 battery discharges for the 20-year life of the battery. This restriction had not been identified in any DBD. The licensee stated that the requirement would be added to DBD 4.0 *

Appendix A of DBD-7.02, "Palisades Design Basis Document EQ Master Equipment List,"

Revision 2, incorrectly listed the location for L T-0383; referred to E/P 0343 instead of E/P 0346; and did not include SV-3213B in Table A-1. The licensee issued DBD Change Requests 7.02-4-R2-0522, 7.02-6-R2-0527, and 7.02-4-R2-0523 to correct the DB *

OBO 2.01, "Low Pressure Safety Injection System," Revision 3, and DBD 2.02, "High Pressure Safety Injection System," Revision 3, both contained references to ANF-88-107,

"Palisades Large Break LOCA/ECCS Analysis With Increased Radial Peaking," Revision ANF-88-107 was superseded by Seimemi"Calculation EMF-96-172, "Palisades Large Break LOCA/ECCS Analysis," Revision 0. The licensee initiated OBD Change Requests 2.01-30-R3-0519 and 2.02-27-R3-0520 to update the OBD *

DBO 2.01, "Low Pressure Safety Injection System," Revision 3, Section 3.3.1.3, stated that the SIRWT must maintain a minimum of 20,000 gallons at the time of a RAS to limit the radiological consequences of an accident. The 080 reference for this statement was EA-TAM-95-05, "Radiological Consequences for the Palisades Maximum Hypothetical Accident & Loss of Coolant Accident," Revision 0. A review of EA-TAM-95-05 indicated that this analysis did not take credit.for the 20,000 gallons at the time of RAS to limit the radiological consequences of an accident. The licensee issued 080 Change Request 2.01-31-R3-0524 to update the OB The team also identified the following discrepancies in other documentation:

P&ID M-232, Sheet 2A, incorrectly identified LT-0383 as connected to penetration #54 instead of #56. The licensee issued Document Change Request (OCR) 97-0856 to correct the drawin *

Documents E-33, Revision 46, and E-37, Revision 46, were not revised to reflect the installed condition of the battery charger cabling that was rerouted by SC 89-284. The licensee issued CR C-PAL-97-1495 to resolve this discrepancy.

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P&ID M-209, Sheet 3 (Revision 34), incorrectly depicted valves SV-0918 and SV-09778 as nonnally deenergized. The licensee issued EAR 97-0652 to revise the drawin *

Vendor drawing E-12A, Sheet 39, Revision 0, indicated that the battery discharge characteristics were based upon battery cell specific gravities of 1.215 +/- 0.005. However, the batteries were being maintained to a criterion of 1.215 +/- 0.010. The licensee issued EAR 97-0669 to update the drawin These documentation discrepancies were not consistent with 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control,* which requires that the design basis be correctly translated into drawings. The team identified this item as Unresolved Item 50-255/97-201-3 E1.5.3 Conclusions The team identified several minor discrepancies in the FSAR, which indicated the need for improved control and updating of this document. The DBDs also contained discrepancies, as did several drawing..

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APPENDIX A OPEN ITEMS This report categorizes the inspection findings as unresolved items and inspection followup items in accordance with NRC Inspection Manual, Manual Chapter 0610. An unresolved item (URI) is a matter about which more infonnation is required to detennine whether the issue in question is an acceptable item, a deviation, a nonconfonnance, or a violation. The NRC Region Ill office will issue any enforcement action resulting from their review of the identified URls. An inspection followup item (IFI) is a matter that requires further inspection because of a potential problem, because specific licensee or NRC action is pending, or because additional infonnation is needed that was not available at the time of the inspection. The URls and I Fis found in this inspection are listed below:

Item Number 50-255/97-201-01 50-255/97-201-02 50-255/97-201-03 50-255/97-201-04 50-255/97-201-05 50-255/97-201-06 50-255/97-201-07 50-255/97-201-08 50-255/97-201-09 50-255/97-201-10 50-255/97-201-11 50-255/97-201-12 50-255/97-201-13 50-255/97-201-14 50-255/97.,201-15 50-255/97-201-16 50-255/97-201-17 Finding

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IFI URI URI URI URI URI URI URI URI URI IFI IFI URI URI IFI URI IFI Completed CCW Flow Model Calculation (Section E1.2.1.2(a))

CCW Design Temperature (Section E1.2.1.2(a))

RV-0939 Valve Testing (Sections E1.2.1.2(c))

Incomplete Analysis (Section E1.2.1.2(d))

Design-Basis -Implementation (Sections E1.2.3.2(a))

SI Check Valve Testing (Section E1.3.1.2(f))

Revising EAs (Sections E1.3.1.2(g))

Revising EAs (Section E1.3.1.2(h))

Interactions With Safety-Related Equipment (Section E1.3.1.20))

ECCS Suction Piping (Section E1.3.1.20))

Fire Barrier (Section E1.3.1.20))

Uncertainty Calculations (Section E1.3.3.2(a))

Instrument Tubing Slope (Section E1.3.3.2(e))

AC Load Update (Section E1.4.2(a))

Cable Degradation (Section E1.4.2(a))

Possible USQ (Section E1.4.2(a))

Documentation of Design Basis (Sections E1.4.2(a))

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50-255/97-201-18 50-255/97-201-19 50-255/97-201-20 50-255/97-201-21 50-255/97-201-22 50-255/97-201-23 50-255/97-201-24 50-255/97-201-25 50-255/97-201-26 50-255/97-201-27 50-255/97-201-28 50-255/97-201-29 50-255/97-201-30 50-255/97-201-31 URI URI URI IFI IFI IFI IFI URI IFI IFI IFI IFI URI URI Calibration Test (Section E1.4.2(a))

Agastat Relays (Section E1.4.2(a))

Battery Charger Operation (Section E1.4.2(b))

de Loading Analysis (Section E1.4.2(b))

Non-conservative TS (Section E1.4.2(b))

de Calculations (Section E1.4.2(b))

de Load Terminal Calculation (E1.4.2(b))

Solenoid Coils (Section E1.4.2(b))

Electrical Calculation Deficiencies (Section E1.4.2(b))

EDG Testing (Section E1.4.2(c))

Battery Testing (Section E1.4.2(c))

Safety Reviews (Section E1.4.2(d))

FSAR Discrepancies (Section E1.5.2))

Documentation Discrepancies (Section E1.5.2))

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ac AIR ANSI AOV ASME ccw CR css DBA DBD de DCR DG DPI DR EA EAR ECCS EDG EOP EQ ESG ESS FC FES FSAR FT GL gpm HELB HPSI HVAC ICEA IEEE IFI IN IP IPCEA IST KW LCO LER LOCA LOOP LPSI MCC MOV MSLB NEC APPENDIX B LIST OF ACRONYMS USED alternating current action item record American National Standards Institute air-operated valve American Society of Mechanical Engineers component cooling water condition report containment spray system design-basis accident design-basis document direct current document change request diesel generator differential pressure indicator deviation report engineering analysis engineering Assistance Request emergency core cooling* system emergency diesel generator emergency operating procedure equipment qualification engineered safeguards engineered safeguards system facility change functional equivalent substitution final safety analysis report flow transmitter generic letter gallons per minute high-energy line break high-pressure safety injection heating, ventilation, and air conditioning Insulated Cable Engineers Association Institute of Electrical and Electronics Engineers inspection followup item NRC information notice inspection procedure Insulated Power Cable Engineers Association in-service testing kilowatt limiting condition for operation lice11see event report loss-of-coolant accident loss of off site power low-pressure safety injection motor control center motor-operated valve main steam line break National Electric Code Bl

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NPSH NRC NRR OM P&ID PCR PCS PPAC ppm PS psig RAS RG RHPSI RV SBO SC soc SEP SGTR SI SIRWT SIS SIT SMM SOP SSDC sv SW TE TM TS URI USQ v

net positive suction head Nuclear Regulatory Commission NRC Office of Nuclear Reactor Regulation operating and maintenance piping and instrument diagram procedure change request primary coolant system periodic and predetermined activity parts per million pressure switch pounds per square inch gauge recirculation actuation signal NRC regulatory guide redundant high pressure safety injection relief valve station blackout specification change shutdown cooling *

Systematic Evaluation Program steam generator tube rupture safety injection safety injection and refueling water tank safety injection signal safety injection tank subcooling margin monitor standard operating procedure safety system design confirmation solenoid valve service water temperature element temporary modification technical specification unresolved item unreviewed safety question volt 82