IR 05000261/1997201

Insp Rept 50-261/97-201 on 970407-0523.No Violations Noted. Major Areas Inspected:Si & AFW Sys & Support Sys

ML14181A944
Person / Time
Site:	Robinson
Issue date:	08/14/1997
From:	NRC (Affiliation Not Assigned)
To:
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ML14181A943	List: IR 05000261/1997201
References
50-261-97-201, NUDOCS 9708280104
Download: ML14181A944 (59)

Text

U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Docket N License N DPR-23 Report N /97-201 Licensee:

Carolina Power and Light Company Facility:

H. B. Robinson Steam Electric Plant, Unit No. 2 Location:

Hartsville, Date:

April 7 through May 23, 1997 Inspectors:

E.A. Kleeh, Team Leader, Special Inspection Branch C.J. Baron, Contractor*

P.J. Bienick, Contractor*

R.B. Bradbury, Contractor*

M. Huq, Contractor*

D. Schuler, Contractor*

(* Contractors from Stone & Webster Engineering Corporation)

Approved By:

Donald P. Norkin, Section Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation 9708280104 970814 PDR ADOCK 05000261 G

PDR

SUMMARY From April 7 through May 23, 1997, the staff of the U.S. Nuclear Regulatory Commission (NRC),

.

Office of Nuclear Reactor Regulation (NRR), Special Inspection Branch, conducted a design inspection at H.B. Robinson Steam Electric Plant, Unit No. 2 (HBR) operated by Carolina Power &

Light Co. (CP&L). This inspection included onsite inspections during April 21-May 2 and May 12 23, 1997. The inspection team consisted of an NRR team leader and five engineers from Stone &

Webster Engineering Corporation (SWEC) under contract to the NR 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 updated final safety analysis report (UFSAR). The team selected the safety injection (SI) and auxiliary feedwater (AFW) systems, and their support systems, for this inspection because of the importance of these systems in mitigating design basis accidents at HB For guidance in performing the inspection, the team followed the applicable engineering design and configuration control sections of Inspection Procedure 93801, "Safety System Functional Inspection" (SSFI). The team reviewed relevant portions of the UFSAR, design-basis documents, Technical Specifications (TS), drawings, calculations, modification packages, procedures, and other associated plant document In general, the selected systems were capable of performing their safety functions, and design and licensing bases were adequately adhered to. Safety evaluations for plant modifications reviewed by the team were generally adequate and reached appropriate conclusions. In general, as-built configurations of the systems were consistent with design drawings, and system design documents adequately supported the design. However, a number of issues were identified by the team as discussed belo A Siemens Power Corporation calculation indicated that the maximum fuel peak clad temperature (PCT) could be 21020 F during the SI system transfer from injection to recirculation following a large break loss of coolant accident (LBLOCA). This temperature was more than 500F above the previous analysis and higher than the PCT stated in the UFSAR from June 28, 1995, to October 14, 1996 for the blowdown phase of an LBLOCA. The licensee did not report this significant PCT change to the NRC as required by 10 CFR 50.46. (E1.3.2.2(a))

The PCT rise (second PCT peak in the overall LBLOCA evaluation) during the switchover phase of the LBLOCA may not have been evaluated by the NRC. The licensee reviewed NRC inspection reports and discovered references where the second peak had apparently been evaluated and found acceptable. The inspection reports did not completely clarify the issue because of the wide margin between the PCT value of 1400 OF agreed upon in 1992 on the basis of a Westinghouse analysis and the PCT value of 2102 OF supported by the 1994 Siemens analysis of the switchover phase for the LBLOCA. The team referred this matter to the NRR staff for review and evaluation. (E1.3.2.2(a))

The 'icensee discovered that redundant cables for two SI pumps were routed in the same raceway. The licensee declared one pump inoperable, immediately placed the installed spare pump in service, and implemented a modification to provide correct separation subsequent to this inspection. The licensee notified the NRC of this discrepancy in accordance with 10 CFR 50.72 on May 21, 1997. (E1.3.3.2(a))

The UFSAR discussion of NPSH for SI pumps was based on the original design configuration of three SI pumps operating at 600 gpm each. The SI system operating configuration had been changed to use only two inservice pumps, each capable of delivering approximately 640 gp The licensee had not established a new design basis analysis for this operating mode chang The team recognized a similar lack of analysis for the residual heat removal (RHR) pump's NPSH requirements, as documented in the UFSAR, when supplied from the containment building sum In each case, the design basis analyses were the original Westinghouse calculations, maintained by Westinghouse, and were not under the licensee's control. The licensee initiated a hydraulic analysis to verify that the SI and RHR pumps had adequate NPSH under the most limiting conditions. The hydraulIc analysis was completed in July 1997 and the licensee concluded (as verified in LER 97-08) that prior to raising the RWST level setpoint (subsequent to the inspection)

that SI pumps B and C had insufficient NPSH while pump A had adequate NPSH. The team has not had the opportunity to review this analysis along with the potential impact on the past and present operability of the SI and RHR pumps. (E1.3.2.2(c))

Several sensing lines for post-accident monitoring instruments were not sloped per their design basis. Other post-accident monitoring instruments did not have a seismic uncertainty term included in their setpoint calculations as required by their design basis. The RWST level instrument, which dictated when operators shifted the SI and RHR pumps' suction to the

containment sump, also lacked a seismic uncertainty term in its setpoint analysis. The team questioned the licensee's position that this uncertainty term was not required for the RWST instrument because the plant's licensing basis did not postulate a seismic event in conjunction with a LOCA. The team referred this issue to the NRR staff for review and evaluatio (El.2.4.2(c); E1.3.4.2(b); El.3.4.2(e))

The team believes that the redundant condensate storage tank (CST) level instruments are not able to withstand an emergency diesel generator (EDG) A failure coincident with a loss of offsite power (LOOP) for longer than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The team referred this issue to the NRR staff for review and evaluation. In addition, the alarm setpoints for these instruments do not consider the uncertainty introduced by a seismic event. (El.2.4.2(a)

Several valves that performed accident mitigating functions were not included in the in-service testing program. These included SI valves used to realign the SI system in the event of an accident while filling the accumulators, and the component cooling water (CCW) and service water (SW) system valves used to isolate leaks in the room containing the RHR pumps. (E1.3.2.2(e);

E 1.3.2.2(f))

The team identified several discrepancies in the plant procedures. For example, the containment floor water level setpoint used in an emergency plant procedure to initiate post-accident recirculation did not agree with the UFSAR value and may not have included the

appropriate instrument uncertainty. The procedure for transfer of SI to cold leg recirculation did not provide direction should an RHR heat exchanger discharge valve fail to close during a series of procedural steps that must be completed in a short time. Additionally, the procedures for performance testing of the station batteries had several deficiencies. (El.3.2.2(d); E1.3.4.2(c);

E1.4.2.3)

Other issues concerned differences between the design basis and the as-built plant, lack of testing for certain batteries, station battery test control, lack of timely resolution of cracking of relays in the Reactor Protection System, lack of design basis calculations for some field flash batteries, and a 10 CFR 50.59 screening error. (E1.4.2.1; E1.4.2.2; E1.4.2.3; E1.3.2.2(f))

The team identified discrepancies in the control, documentation, and preparation of calculation Several voided or superseded calculations were not marked as such and numerous calculations contained incorrect inputs, incorrect methods, incomplete analysis, and undocumented assumptions. For example, a calculation that concerned the effects of high energy line breaks on the CCW system inside containment did not consider all potential breaks, contained an error in methodology, and did not adequately document the analysis. However, none of the deficiencies affected the final conclusion of any calculation. (E1.5.2.1; E1.5.2.2)

The team also identified discrepancies in the UFSAR, design basis documents, system descriptions, and drawings. (E1.2.6; E1.3.6; E1.4.2.6)

The licensee implemented appropriate measures to resolve the immediate concerns identified by the team. For other issues, the licensee initiated appropriate reviews, evaluations, or took corrective actions such as revising design documents and changing procedure III

Report Details Ill. Engineering El Conduct of Engineering E1.1 Inspection Objectives and Methodology The primary objective of the design inspection at HBR was to evaluate selected systems'

capability to perform their safety functions, adherence to design and licensing bases, and consistency of their as-built configurations and system operations with the UFSAR. The systems selected for inspection were the AFW System and the SI System. These systems were selected with consideration of their importance in mitigating design-basis accidents (DBAs) at HBR. The inspection focused on the engineered safeguards functions of these systems and the interfaces with other systems. The inspection was performed in accordance with the applicable engineering design and configuration-control portions of Inspection Procedure (IP) 93801, "Safety System Functional Inspection."

The open items resulting from this inspection are identified in Appendix A. Appendices B and C list the exit meeting attendees and the acronyms used, respectivel E1.2 Auxiliary Feedwater (AFW) System E1.2.1 System Description and Safety Functions O

The AFW system consists of two parallel trains. One train has two parallel motor-driven (MDAFW) pumps, and the other a single steam-driven (SDAFW) pump. Each pump is capable of supplying high pressure water into all three steam generators (SG's) independently via the feedwater headers. Each MDAFW pump is sized to supply the SGs with 100 percent of the required feedwater flow for a normal reactor coolant system (RCS) cooldown. The SDAFW pump is sized to supply 200 percent of the required feedwater flow for a normal RCS cooldown. The system is designed to supply sufficient feedwater to at least two SGs despite a single failure. The MDAFW pumps can be used to fill the SGs under any plant condition, but the SDAFW pump can only operate when RCS temperature is above 3500 F. Under accident conditions, only a single MDAFW pump or the SDAFW pump is required to maintain level in the SG During normal power operation, the AFW system is aligned to receive water from the CS Backup sources of water are the SW System and the Primary and Demineralized Water Makeup System, which uses the deep wells. Realignment of the AFW system from the CST to either backup source is a manual operatio The AFW system supplies high pressure feedwater to the SGs during plant startup and cooldown and, as an engineered safety feature, removes decay and sensible heat from the RCS via the SGs during an accident. The AFW system also provides sufficient AFW flow to prevent loss of the pressurizer vapor space during a feedwater line break accident coincident with a loss of offsite power. The SDAFW pump is fully functional during a station blackout conditio E1.2.2 Mechanical Design Review E1.2.2.1 Inspection Scope The team evaluated the AFW system's capability to perform its mechanical design safety functions. The evaluation included review of the UFSAR, TS, design basis documents (DBDs),

system descriptions (SDs), flow diagrams, equipment specifications, equipment drawings, manufacturer's information, design modifications, operating procedures, and applicable analysis and calculations. In addition, the team performed several system walkdowns and observed surveillance test E1.2.2.2 Observations and Findings E1.2.2.2(a) Condensate Storage Tank (CST)

The CST is the primary source of water for the AFW system for all transient and accident conditions. The total volume of the CST is approximately 200,000 gallons. The CST is administratively controlled at least 50 percent full. TS 3.4.1.c. requires a minimum of 35,000 gallons (28% level, allowing 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of AFW operation) of water in the CST and a water supply from the lake via either one of the two legs of the SW system. The CST is a seismic Category 1 component. The pH of the CST water is maintained at 9.4, which reduces the effects of corrosion in the system. Additionally, the inside of the CST is fitted with a rubber membrane to prevent air from coming in contact with the water surfac The team reviewed the CST manufacturer's drawings 5379-1581, Revision 4, and HBR2-9671, Sheet 1 through 4, and verified that the volume was adequate. The team also verified that the appropriate ground acceleration and stress limits were used in the design basis for the seismic analysis of the CS The CST is located outside and exposed to tornado-generated missiles. The original design basis was that the plant could achieve safe shutdown in the event of a tornado, since the SW System was the designated backup to the CST. The licensee had completed a tornado-missile probablistic risk assessment (PRA), NUS-4396, with a concurrent single failure in the AFW and SW systems in 1985 and concluded that both systems' availability was acceptable. The final results were transmitted to the NRC on June 13, 1985, and were approved by the NRC with no exception E1.2.2.2(b) AFW System Performance The team reviewed the multistage MDAFW pumps and their associated electric motors. Each pump-motor combination was sized adequately to supply the required flow to the SGs. Similarly, the single-stage SDAFW pump and its single-stage turbine were sized adequately for their duty conditions. The team also reviewed calculation RNP-M/MECH-1460, "NPSH vs. CST Level for SDAFW Pump," Revision 0, which was the limiting condition for all the pumps, and verified that the available NPSH was adequat Operating Procedure OP-402, "Auxiliary Feedwater System," Revision 41, provides instructions to the operator for shifting the suction of the AFW pumps from the CST to the SW syste Operators stop the AFW pumps when the CST water level reaches 10 percent of the CST's

total volume, manually realign AFW from the CST to the SW system, and then restart the AFW pumps. However, the procedure did not provide any time limit for restarting the AFW pump The team was concerned that the SG tubes could uncover and that the reactor coolant pressure boundary could be adversely affected when cold AFW flow was restored. The licensee stated that the SGs will boil dry in about 30 minutes following a reactor trip from full power with no AFW flo The SG boil-dry time is a dynamic parameter that increases with elapsed time following the reactor trip because reactor decay heat steadily drops. The licensee did not want to include a time limit because operators were trained to closely monitor the swapover (a critical safety maneuver), because the 2-hour drawdown time of the CST permits the plant to be stabilized prior to the swapover, and because the swapover from the CST to the SW system can be completed in less than 30 minutes. The team agreed that the procedure did not need to specify a time limit for restarting the AFW pumps following the swapover to the SW syste The team reviewed the capability of the AFW system to accommodate a design overspeed of the SDAFW pump's turbine with the resulting increased pressure on the piping and fittings. The licensee's review of NRC 'nformation Notice (IN) 90-45, "Overspeed of the Turbine-Driven Auxiliary Feedwater Pumps and Overpressurization of the Associated Piping System," primarily concentrated on the failure of the overspeed governor. The licensee did not address the potential overpressurization of the AFW piping and components during a turbine overspeed until prompted by the team. The licensee's subsequent review concluded that the piping system would meet the applicable design code requirements at maximum SDAFW pump overspeed by assuming the piping operated at 2000F instead of 5000F and allowing the potential overpressurization condition to exist for only a short time period (15% of the operating time). The 2000F temperature was conservative relative to the 11 50F temperature for AFW stated in document EMF-96-049, "Plant Parameter Document, H.B. Robinson Unit 2 Cycle 18" (May 1996). The affected discharge piping was checked every shift by the outside auxiliary operator for increased temperature due to steam inleakage. The licensee also took ultrasonic measurements at two locations and determined that the pipe wall thickness was consistent with the design basis and that excessive corrosion had not occurre The team questioned the licensee's followup action for NRC Information Notice 91-38, "Thermal Stratification in Feedwater System Piping." Although a release for engineering task (RET) was drafted to investigate the applicability of this IN, the licensee could not locate the final RET. The licensee assigned Operating Events (OE) Review 97-01071 to evaluate this IN. The team identified this item as Inspector Followup Item (IFI) 50-261/97-201-0 E1.2.2.2(c) AFW System Testing The team reviewed Technical Support Management Manual procedure TMM-004, "In-Service Inspection Testing," Revision 44, and determined that all the appropriate AFW valves and the AFW pumps were included in the inservice testing progra The team reviewed Operations Surveillance Test procedure OST-202, "Steam Driven Auxiliary Feedwater System Component Test," Revision 40, for the monthly surveillance of the SDAFW pump and system components, and witnessed the monthly surveillance pump test on April 30, 1997. The licensee performed a comprehensive briefing in the control room and later

established clear and interference-free wireless communication between test personnel. The testing was performed in accordance with the procedure and the results were verified by the team as acceptable and correctly documented E1.2.2.2(d) AFW System Valve Operation The team reviewed the capability of the AFW system valves to perform their required functions under normal and accident conditions. The review included calculation RPN-MN/MECH-1056,

"AFW SD Pump FCV Setpoint Revision Calibration Data for RPN-2", Revision 2, and calculation E9124-1, "Maximum and Required Thrust Analysis, Carolina Power and Light Company," Revision A, for the motor-operated valves. The team found the inputs, assumptions, methodology, and results of these calculations appropriat E1.2.2.2(e) AFW System Modifications and Engineering Evaluations (EE)

The team reviewed two modifications and two engineering evaluations: nos. 988, "V2-14, V2-16, V2-18 Valve Operator Upgrade"; 1018, "Replacement of AFW Pump Suction Piping"; EE88-0083,

"Rearrange the AFW Valve Controllers Position Lights on Reactor Turbine Generator Board (RTGB)"; and EE91-146, "Replacement of AFW-9 Check Valve." The team assessed each modification to verify that the problem identification and the justification for the modification were clearly stated; to determine if the problem was generic, potentially affecting other systems; to determine the impact on the design and licensing basis; to verify that the safety evaluation, when required, was correct and consistent with the applicable procedures; to verify that the post modification testing was complete and appropriate; and to verify that the required plant documentation was updated to reflect the modificatio E1.2.2.2(f) AFW System Walkdown The team conducted a walkdown of the accessible portion of the AFW system and identified one concer The team noted that the operators for valves V2-16A, B, and C were not oriented vertically; however, stress isometric drawing FW-2, sheets 3A and 3B, Revision 1, show the operators oriented vertically. The team questioned the acceptability of this installation. An earlier stress analysis of the installed configurations for the valve operators demonstrated their acceptabilit The licensee issued Condition Report (CR) 97-01137 which recommended revising the isometric drawing E1.2.2.3 Conclusions The team concluded that the CST was capable of providing sufficient water to AFW during both normal and accident conditions and that mechanically, it was consistent with the design and licensing bases. AFW pumps and their motors including associated piping were sized correctly and would respond in time to prevent the SGs from boiling dry for a reactor trip at full power. The SDAFW pump's discharge piping was sized adequately for a turbine overspeed. All the appropriate AFW pumps and valves were included in the inservice testing program. The functional analysis for all MOVs appeared to have the correct design inputs and methodolog All mechanical modifications for the AFW system reviewed by the team were adequate. The team was concerned that the licensee had not evaluated the applicability of IN 91-38 to the AFW syste E1.2.3 Electrical Design Review E1.2.3.1 Inspection Scope The team evaluated the electrical loads required for the AFW system to perform its design functions under both normal and accident conditions. This evaluation addressed electrical alternating current (AC) bus loading, direct current (DC) battery loading, protective coordination, cable sizing, and modification package E1.2.3.2 Observations and Findings E1.2.3.2(a) AFW Electrical Distribution To verify AFW electrical loads on safety buses, the team reviewed Robinson Nuclear Plant (RNP)

calculations RNP-E-7.002, "Emergency Equipment Load Factor Study," Revision 5; RNP-E-8.016,

"Emergency Diesel Generator Static and Dynamic Analysis," Revision 5; and RNP-E-8.002, "AC Auxiliary Electrical Distribution System Voltage/Load Flow/Fault Current," Revision 3. All major AFW electrical loads were accounted for in those calculations for both normal and accident conditions. The motors were sized to accelerate the MDAFW pumps within the required accident time lines and to drive them for long-term continuous operation. The team determined that the methodology and assumptions used were appropriat *

The team also reviewed calculations RNP-E-5.004, "Ampacity Evaluation of Safety Related Power Cables on 480V and 208V AC MCC's and Buses," Revision 4; RNP-E-2.001, "Overcurrent Protection for Aux. Feedwater Pumps A and B Motors," Revision 2; and RNP-E-8.045, "AFW, SI, RHR, and CCW Pump Motor Characteristics Impacts due to High Ambient Temperatures,"

Revision 0. The team determined that the sizes of installed AFW cables, the overcurrent protection for AFW electrical loads, including cables, and the operating characteristics of major AFW motors conformed to industry standard In reviewing calculation RNP-E-8.016, the team noted that the AFW pumps would produce 380 gpm for the first 30 minutes after a LBLOCA and 100 gpm for the next 30 minutes. The team questioned this operating scenario, and the licensee said that the actual operation of the AFW pumps might not be as stated in the calculation. The AFW flow could, instead, gradually reduce over the entire time period to match SG losses. The licensee stated that the conservative approach utilized in the calculation bounded either of these two cases and that emergency operating procedures required operators to maintain EDGs' loading. However, the procedures did not specifically state the EDG emergency loading limits since operators were trained to know them. The licensee stated that Operations Management Manual procedure OMM-22, "Emergency Operation Procedures Users Guide," Revision 11, would be updated to provide guidance concerning allowable EDG loads and for monitoring EDG loading in excess of its continuous rating (2500 kW). The licensee issued CR 97-01074 to track this concer E1.2;3.2(b) Electrical AFW Modifications The team reviewed Modification M-1 145, "FW-V2-6A Cable Replacement," Revision 9, for its

.

technical adequacy. This modification replaced the existing cables for the feedwater block valves (FW-V2-6AB,C) with larger cables in order to reduce the voltage drop in the cables. The modification was performed because the valves could not develop enough torque to ensure valve closure. This effect was attributed to excessive feeder cable voltage drop under degraded voltage conditions. The team determined that this modification was technically adequate and that the safety evaluation was properly performe E1.2.3.2(c) AFW System Testing Section 4.6.7 of DBD/R87038/SD16, "Electrical Power Distribution System," Revision 0, committed to testing instantaneous and thermal overload trip elements for the molded case circuit breakers associated with valves V2-16A and V6-16C. The team reviewed test procedure MST-925,

'Westinghouse Molded Case Circuit Breaker Thermal & Instantaneous Trips Testing," Revision 6, which was performed on September 14, 1996, and concluded that the testing was performed for valves V2-16A and V6-16C as prescribed in DBD/R87038/SD16. During the review of the test package, the team noted that MST-925 provided no guidance for torquing the line side connections. The licensee demonstrated that adequate torquing was produced for these devices and initiated ESR 97-00266 to incorporate the manufacturer's recommendations for torque values in the Maintenance Surveillance Test (MST).

E1.2.3.2(d) AFW System Walkdown The team performed a walkdown of the MDAFW and SDAFW pump areas. The areas were clean and well maintained. The team questioned the lack of grounding connections across flexible conduits. The team determined that this design was consistent with the design basis and industry standards when the plant was constructed. The licensee updated its installation practices in 1990 to provide ground connections across flexible conduits that are greater than 1 1/2 inch in diamete A review of the MDAFW motors showed that they conform with the original motor outline drawing 5379-1127, Revision 3. The team identified no concerns during this walkdow E1.2.3.3 Conclusions The electrical design for components that perform the normal and accident functions of the AFW system supported the design basis functions of the system. The electrical system provides independent, redundant, safety-related power to the electrical AFW loads. The team verified that the MDAFW pump's motors were sized in accordance with system demands and industry guidelines. The basis for determining the ampacity and overcurrent protection of installed AFW cables seemed appropriate. The team identified a concern regarding lack of procedural guidance for torquing line-side connections for some circuit breaker overload trip element E1.2.4 Instrumentation & Controls (l&C) Design Review E1.2.4.1 Inspection Scope The team evaluated the ability of the instrumentation & controls for the AFW system to perform the design safety functions. The team reviewed sections of the UFSAR, applicable TS sections, modification packages, the AFW system DBD, flow diagrams, control wiring diagrams, plant procedures and calculations. System walkdowns were also conducted. An in-depth evaluation was performed of the CST level instrumentation, AFW pump shutdown circuitry, AFW actuation logic, and instrument loop uncertainty calculation E1.2.4.2 Observations and Findings E1.2.4.2(a) Condensate Storage Tank The design review includcd a detailed evaluation of the CST level transmitters AFW-LT-1454A and 1454B, which provide redundant low level alarms and continuous, real-time indication in the main control room. The alarms actuate at a CST level of 10 percent to notify the operator to transfer the AFW pump suction from the CST to an alternate source, normally the SW syste AFW-LT-1454A is powered from instrument bus #1, which is a EDG A-backed source. AFW-LT 1454B is powered from instrument bus #2, which is a battery A-backed sourc During a loss of offsite power (LOOP), an EDG A failure could cause the loss of both redundant level instruments. AFW-LT-1454A is lost immediately due to the LOOP and the EDG A failure, and AFW-LT-1454B is lost 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> later when battery A is completely discharged as described in UFSAR Section 8.3.2. TS section 3.4 stated that the CST should have at least a 2-hour supply of water at hot standby condition The licensee stated that the loss of primary CST level indication after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was not a problem because the licensee could rely upon the non-safety CST level indication provided by AFW-LT 1454C and also because an operator would have transferred the AFW suction from the CST to the SW system before 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> had elapse Revision 2 of the licensee's RG 1.97 submittal, (NLS-86-267 letter to the NRC, dated July 28, 1986) listed CST level as a Type A, Category 1 variable. Thus, the primary CST level indication circuits must be redundant and be powered from onsite (standby) power sources. Section 3.6 of Generic Issues Document GID/R87038/0008, "Regulatory Guide 1.97," Revision 0, states that the single failure of a RG 1.97 Category 1 instrument, its auxiliary supporting features, or its power sources, concurrent with failures resulting from an accident, should not prevent operators from being able to determine the plant's safety status and to maintain it in a safe conditio Furthermore, UFSAR Section 3.1.1.2 states that items whose failure might cause, or increase the severity of, an accident are designated Class. The team believes the current design did not provide true redundant power supplies for the CST level instrumentation circuits because both are powered from train A. This issue was referred to the NRR staff for resolution. The team identified this item as Unresolved Item (URI) 50-261/97-201-0 E1.2.4.2(b) MDAFW and SDAFW Pump Trip Discharge Pressure Switches The team reviewed the design of the SDAFW pump pressure switches PSL-1476-1 and PSL

.

1476-2 and the MDAFW pump switches PSL-1474A/B-1 and PSL-1474A/B-2. Each switch provides an alarm and trips the respective pump on low discharge pressure possibly caused by for example, loss of suction pressure or significant pump degradatio Section 3.1.5.1 of DBD/R87038/SD32, "Auxiliary Feedwater System," Revision 0, states that, in accordance with a CP&L letter to the NRC (NO-80-746 dated May 15, 1980), each AFW pump should be protected from damage due to low suction pressure or loss of water supply by two redundant pressure detectors. The pressure switches were installed in the pumps' discharge piping. The team was concerned that the location of the switches was not consistent with the DBD. However, SD-042, "Auxiliary Feedwater System," Revision 0, noted the correct location of the pressure switches. The licensee stated that mounting the pressure switches in a pump's suction header lines would unnecessarily actuate the switches for normal operating perturbation The team agreed that the pumps were adequately protected with the switches in their present discharge piping locations and the 650 psig trip setpoint was adequate for the pumps. The licensee issued Document Change Form (DCF) 97-D-0005 to clarify the DBD by describing the actual mounting configuration and to correct other documentatio E1.2.4.2(c) Instrument Loop Uncertainty Calculations The team reviewed calculation RNP-1/INST-1015, "Condensate Storage Tank Level Alarm Setpoints", Revision 0, for adequacy and consistency with the design basis. The CST level instruments were required for safe shutdown. The calculation did not consider seismic effects as an error term. As stated previously, CST level was classified as a RG 1.97 Type A, Category 1 variable, which required it to have continuous indication and redundant seismically qualified channels. Engineering procedure, EGR-NGGC-0153, "Engineering Instrument Setpoints,"

Revision 2, Section 9.4.10, requires the uncertainty term, specified by the vendor, to be included in an uncertainty calculation when the instrument is used in an application requiring seismic qualification. The team was concerned that, after a seismic event, these instruments' outputs could be outside their design basis since the calculation did not include seismic uncertaint Seismic qualification of an instrument verifies that it will be functional during and after a seismic event, but not that its accuracy is unaffected by the seismic event. Inclusion of the seismic uncertainty term takes into account the effects of the seismic event on the overall accuracy of an instrument. The calculation did not refer to licensee's procedure, EGR-NGGC-0153, for scope and methodology of instrument uncertainty and scaling calculations. The licensee's Quality Assurance Program in USFAR 17.3.1.7 commits to compliance with the Regulatory Guides and Standards listed in USFAR, Section 1.8. The licensee commits to American National Standards Institute (ANSI) standard N45.2.11-1974 for design and modification of the plant in USFAR Section 1.8, page 1.8.0-12, Amendment No. 3. The Corporate Quality Assurance Manual, Revision 18, Section 3.4.3 requires design verification be performed to substantiate that the final design document meets the appropriate design inputs. Criterion III, "Design Control" of Title 10 of Code of Federal Regulations (CFR) Part 50, Appendix B states that the design control measures shall provide for verifying or checking the adequacy of the design. The licensee issued CR 97 01211 to evaluate inserting the seismic uncertainty term into the calculation. The licensee issued Revision 1 of the calculation after the inspection. The team reviewed the calculation and verified that the associated setpoints were not affected. The team considered the calculation's design inputs as nonconservative because of the exclusion of the seismic uncertainty term. This issue

represents a weakness with respect to Criterion III, "Design Control" of 10 CFR Part 50, Appendix B. The team identified this item as part of Unresolved Item 50-261/97-201-0 *

E1.2.4.2(d) Other Calculations The team reviewed calculation RNP-I/INST-1039, "Steam Generator Narrow Range Level Accuracy and Scaling Calculation," Revision 0. The team found that this calculation was consistent with the design basis and the setpoint program methodology. Seismic uncertainty was appropriately considered in the loop uncertaint E1.2.4.2(e) AFW Actuation Logic The steam generator narrow range signal inputs to the AMSAC [ATWS (Anticipated Transient Without Scram) Mitigation System Actuation Circuitry] was reviewed for consistency with the design basis. The team verified appropriate isolation devices were installed between non-Class 1 E AMSAC circuits and the Class 1 E AFW circuits. AFW automatic initiation/isolation logic design was also found to be consistent with the guidance provided in NUREG 0578, "TMI-1 Lesson Learned Task Force Status Report and Short Term Recommendations."

E1.2.4.2(f) Modifications and Engineering Evaluations Two engineering evaluations and two modifications were reviewed: EE88-0083, "Rearrange the AFW Valve Controllers Position Lights on the RTGB"; EE88-0203, "Relocate Indicating Lights on DSS Panel for Motor Operated Valve (MOV) V1-8A"; 93-0072, "Westinghouse BFD Control Relay Replacement"; and 937, "Upgrade AFW Flow Indication." Each modification was reviewed for adequacy and consistency with the design bases and design configuration. The 10 CFR 50.59 evaluations were also reviewe Modification 937 installed annubar flow transmitters into the AFW system to give control room operators information on flow to the SGs. The team noted that the calibration data calculated in the modification documentation were not incorporated in the present calibration data. The licensee stated that the modification contained preliminary data and that a later design change notice, DCN-14, "Recalibrate FT's," dated January 27, 1989, revised the calibration by incorporating correct design parameters. The team reviewed the later calculation and determined that it was consistent with the present calibratio The team reviewed page AFW-Figure-5 (Revision 1), of SD-042, "Auxiliary Feedwater System,"

Revision 0, which referenced the annubar model as an ANR 75. The team confirmed that the annubars used in the AFW system were Type 73. The licensee stated that this error will be corrected in a SD-042 revisio The team had no other concerns with the modifications. The modifications reviewed were consistent with the design bases and did not negatively impact the ability of the system to perform its required functions during both normal and accident conditions. No concerns were identified with the 10 CFR 50.59 evaluation E1.2,4.2(g) Walkdown During a walkdown of the AFW system, the team inspected various instrument configuration The team observed that various instrument sensing line runs were not consistent with industry standards or licensee requirements for sloping. The process lines of several AFW flow instruments (AFW-FI-1424,.AFW-FT-1425A, and AFW-FT-1425C) appeared to have no slope. In response to the team's concern, the licensee confirmed the absence of slope by taking appropriate measurements. The inspection revealed that in several cases the absence of slope was due to original design deficiencies or damage from previous outages. Inadequate slope in the lines can have an adverse effect on flow readings by potentially trapping air between the process and the transmitter and inducing error Several of the instruments observed by the team (flow instruments AFW-FT-1425A and AFW-FT 1425C) were RG 1.97 Category 2, post-accident instruments required to make flow monitoring available to the operators. The configuration of the tubing was not in accordance with CP&L Standard Procedure EGR-NGGC-0151, "Evaluating Instrument Sensing Line Installations,"

Revision 0. This procedure requires a minimum of % inch per foot slope up to the fitting assemblies, unless otherwise approved. The licensee issued modification ESR 97-00257 to restore the tubing of various sensing lines to the plant design requirements, and issued CR 97 01005 to correct the other deficiencies. The team identified this item as part of Unresolved Item 50-261/97-201-04. The licensee has had past instances where sensing lines for SI system instruments were not correctly sloped (See Section E1.3.4.2(g) Walkdown), which permitted entrapment of gases and air resulting in an increased risk of faulty reading E1.2.4.3 Conclusions The team concluded that the majority of the AFW system design documents were adequate and consistent with the present design. The AFW system actuation logic, including AMSAC inputs and other instrumentation, were capable of performing the required safety functions. The plant design for various AFW instrumentation loops was found to be consistent with the design bases and RG 1.97 requirements with some exceptions, like the omission of the seismic uncertainty factor from the uncertainty calculation for CST level. The team determined that the slopes of the sensing lines for several AFW flow instruments were not correct and questioned the redundancy of the power supplies for the CST level instrumentation loops. Design changes were adequately evaluated and appropriately documented any changes to the original design base E1.2.5 System Interface Design Review E1.2. Inspection Scope The team evaluated the ability of the SW system to supply water to the AFW pumps and the ability of the Ventilation System (VS) to provide acceptable ambient conditions for operation of the MDAFW pumps. Applicable UFSAR sections, system flow diagrams, DBDs, SDs, procedures, and calculations were reviewed for each interface syste E1.2,5.2 Observations and Findings E 1.2.5.2(a) Service Water (SW) System The team reviewed UFSAR Section 9.2.1, "Service Water;" DBD/R87038/SDO4, "Service Water System," Revision 0; Service & Cooling Water Flow Diagram G-190199, Sheets 1 and 2, Rev.55, Sheet 5, Rev. 41, Sheet 6, Rev.37, Sheet 8, Rev.29, Sheet 9, Rev. 43, and Sheet 10, Rev. 38; OP-903, "Service Water System," Revision 62; and AOP-022, "Loss of Service Water," Revision 1 The SW system design basis includes both active and passive failures. The SW system has multiple pumps and two headers, which are normally cross-connected; the alternate AFW system feedwater supply is from one of these headers. The SW supply to each of the two MDAFW pump oil coolers is from a different SW header, and the supply to the SDAFW oil cooler can be from either header (the SW supply valve to the SDAFW pump was normally locked closed and the pump in the self-cooling mode). The team determined that the SW system could supply the AFW system coincident with a single active failure. A passive failure would not affect the SW supply to the AFW system since such a failure was postulated for the long term, when AFW would not be require The team reviewed several SW system flow calculations of limiting flow conditions in the syste Calculation 789M-M-02, "Service Water System Model Evaluation of Traveling Screen Line Rupture for Double SWP, Single SWBP Operation with TB Isolated," Revision 0, appeared to be intended to analyze failure of all the non-seismic SW piping. The team questioned the exclusion of the SW system piping serving the instrument and station air compressors (this piping was not marked as "Q" on the flow diagram). The licensee performed a walkdown inspection and stated that the piping was not seismically supported. Therefore, the calculation did not determine the most limiting (minimum) flow to the cooled components. The licensee stated that calculation RNP-M/MECH-1 362, "SW Screen Wash Piping Flow Analysis," Revision 0, was the latest evaluation of this scenario. The team reviewed this later calculation and agreed that it demonstrated that the SW system could perform its safety-related functions for this scenario although the failure of the air compressor piping was not specifically evaluate Calculation 789M-M-05, "H.B. Robinson Service Water System Model with One SW Pump (D)

Running, One Booster Pump and with T. Bldg. Isolated," Revision 0, represents a loss of offsite power scenario with one SW pump failing to start automatically using onsite emergency powe This 789M-M-05 calculation determined a pump flow condition in excess of the maximum recommended by the pump manufacturer. The licensee stated that this calculation was superseded by calculation RNP-M/MECH-1 128, "Reduced SW Flow to EDG," Revision 2. This later calculation showed acceptable SW system operation for this scenari The licensee performed a self-assessment in December 1996 on the SW System - Service Water System Operational Performance Inspection (SWOPI) Closure and found that many SW computer models existed and some were outdated. CR 97-00363 was initiated to consider the self assessment recommendation that the service water calculations or models be incorporated into one model and obsolete models be deleted. The licensee stated that updating or voiding obsolete SW calculations would also be tracked by CR 97-0036 E1.2.5.2(b) MDAFW Pump Room Ventilation System (VS)

The team reviewed UFSAR Section 9.4, "Air Conditioning, Heating, Cooling, and Ventilation System"; DBD/R87038/SD36, "Post-Accident HVAC Systems," Revision 2; and System Description SD-036, Revision 0, "Heating Ventilation and Air Conditioning System." The team found the documents satisfactory and did not identify any concerns with the interface between the Ventilation System (VS) and the MDAFW pumps. The MDAFW pump room is equipped with two chillers supplied by the SW system. The chillers are safety-related, seismically designed, and supplied with power from the EDGs. A chiller is automatically started when an MDAFW pump is starte E1.2.5.3 Conclusions The team concluded that the design of the interfaces between the SW, the VS, and the AFW systems is satisfactory and that these interface systems adequately support AFW system operation. The SW system can cope with a single active failure but a passive failure is irrelevant since it is considered in the long term when AFW is not required. The MDAFW pumps are adequately supported in their operation by their room chillers since the later are qualified, redundant, and initiated simultaneously with the MDAFW pumps. The team identified concerns with the licensee not incorporating all SW calculations or models into one model and with obsolete models not being delete E1.2.6 UFSAR, DBD, SD, and Drawing Review The team identified the following discrepancies in the UFSAR:

Section 10.4.8.2 does not list AMSAC as a start signal for AF *

Table 10.4.8-1 has an incorrect value for the SDAFW pump horsepowe These discrepancies had not been corrected or the UFSAR updated to ensure that the information included in the UFSAR contained the latest material, as required by Section 10 CFR 50.71(e).

The licensee issued CR 97-01203 and CR 97-01216 to correct the above discrepancies. The team identified this item as part of Unresolved Item 50-261/97-201-0 The team noted that the "Deterministic Review of the AFW System" referred to in DBD/R87038/SD32, "Auxiliary Feedwater System," Revision 0, was not available. The licensee stated that the DBD would be correcte Section 3.4.2 of DBD/R87038/SDO4, "Service Water System," Revision 2, did not identify all of the portions of the SW system not designed to seismic Class 1 requirements and contained conflicting statements concerning the design basis passive failure in Section 3.0.2 when compared to Section 2.1 of GID/R87038/0013, "Single Failure," Revision 0. The DBD stated complete piping ruptures were the design basis, whereas the GID stated that the guidance of NUREG-0800, "Standard Review Plan," for moderate energy lines applied. The licensee stated that the DBD would be revised as appropriat Additionally, the team noted that stress isometric drawing FW-1, sheet 2, depicted incorrect locations of two valves. The licensee issued CR 97-01191 to correct this discrepanc E1.3 Safety Injection (SI) System E1.3.1 System Description and Safety Functions The primary function of the SI system is to automatically inject cooling water into the reactor vessel in the event of a LOCA. The SI system is designed to operate in three modes: passive accumulator injection, active safety injection, and RHR recirculation. The system is designed to limit the fuel PCT during analyzed accidents and thereby ensure that the core would remain intact, with its heat transfer geometry preserve The principal components of the SI system that provide emergency core cooling are the accumulators, the SI (high head) pumps, and the RHR (low head) pumps. During the passive accumulator injection mode, borated water is injected from the accumulators. During the active safety injection mode, borated water is injected from the RWST by the SI pumps. In addition, borated water is injected from the RWST by the RHR pumps if the RCS pressure is sufficiently low. During the RHR recirculation mode, spilled coolant, injected water, and Containment Spray (CS) System drainage are recirculated from the containment building back to the reactor by the RHR pump E1.3.2 Mechanical Design Review E1.3.2.1 Inspection Scope

The team evaluated the SI system's capability to perform its mechanical design safety function The evaluation included the UFSAR, TS, licensee event reports (LERs), DBDs, GIDs, SDs, drawings, calculations, design modifications, equipment specifications, and the operating and test procedures required to assess consistency with the system design and licensing basis. In addition, the team performed several walkdowns of the accessible portions of the SI syste E1.3.2.2 Observations and Findings E1.3.2.2(a) SI System Performance The team reviewed the available licensing, design, and operations documents related to the SI system's capability to provide adequate emergency core cooling under accident condition The capability of the SI and RHR pumps was reviewed by the team. The pumps were found to have adequate flow and head capacity to provide the emergency core cooling required under accident conditions. The pumps' motors had sufficient capacity to provide the power required for all operating conditions. Overall, the team found the SI system able to provide adequate emergency core cooling under accident conditions, subject to the results of the licensee's hydraulic analysis for the SI syste The Siemens Plant Parameter Document, "H.B. Robinson Unit 2 Cycle 18, EMF-96-049," dated May 1996, documents the principle SI system parameters used in the accident analyses. The results of these analyses were summarized in Section 15 of the UFSA With the exception of the transient associated with the SI system's transfer from injection to recirculation after a LBLOCA, the team found that the information presented in Section 15 of the UFSAR was consistent with the applicable licensing, design, and operations document End Path Procedure EPP-9, "Transfer to Cold Leg Recirculation," Revision 20, directed the operators to stop all RHR pumps and operate only one SI pump from the RWST under post accident conditions while the SI system was being realigned to the containment sump for recirculation operation. EPP-9 states that a RHR pump must be restarted within 30.5 minutes after it was stopped to prevent core damage. The 30.5 minute time limit was based on Siemens Power Corporation Calculation EMF-94-157(P), dated September 23, 1994. This calculation indicates that, during the transfer from injection to recirculation SI system operation after a LBLOCA, the maximum fuel PCT would reach approximately 2102 OF 30.5 minutes after stopping the RHR pumps. The 10 CFR 50.46 safety limit for PCT is 22000 The team noted that this calculated maximum PCT was higher than the LBLOCA PCT value listed in Section 15.6 of the UFSAR. The team asked the licensee why this transient was not described in the UFSAR and whether the transient associated with this transition had ever been reviewed by the NRC. In addition, the team asked the licensee whether the more limiting transient had resulted in an unreviewed safety question (USQ).

The licensee initiated CR 97-01023 to evaluate this condition. As a result of this evaluation, the licensee provided the following summary of the significant meetings and correspondence associated with this item:

On May 30, 1991, the.licensee briefed the NRC on Emergency Core Cooling System (ECCS)

.issues, including the SI system transfer from injection to recirculation SI system operatio This meeting was documented in NRC correspondence dated June 27, 199 *

In September 1991, Westinghouse sent the licensee a Topical Report, Westinghouse Commercial Atomic Power (WCAP) 13071," An Evaluation of the Revised Transfer to Cold Leg Recirculation Procedure." This WCAP discussed the fuel PCT transient that occurred during the SI system transfer from injection to recirculation after a LBLOCA. The analysis calculated a maximum PCT of 1991 OF during the transfer. The WCAP was submitted to the NRC for review by licensee's letter NLS-91-243, dated October 10, 1991. The licensee has correspondence that verifies the submittal of the Westinghouse analysis (WCAP-13071) to the NRC but has no documents that verify its revie In addition, no other analysis on this subject was submitted for NRC revie *

On January 31, 1992, in Rockville, MD, the licensee briefed the NRC on the SI system transfer from injection to recirculation after a LBLOC *

In January 1992, Westinghouse provided WCAP-13186, "An Evaluation of the Revised Transfer to Cold Leg Recirculation Procedure," to the licensee. This analysis calculated a maximum PCT of 1991 OF during the transfer. This WCAP was not submitted to the NRC for revie * When the Westinghouse analyses were performed for the facility, the limiting PCT reported to the NRC was 2178 oF. This value was higher than the calculated maximum PCT of 1991 OF associated with the SI system transfer from injection to recirculation after a LBLOC *

In September 1994, Siemens Power Corporation again analyzed the SI system transfer to cold leg recirculation after a LBLOCA and calculated a maximum PCT of 2102 OF. When the Siemens analysis was performed, the limiting PCT value due to a LBLOCA reported to the NRC was 2134 O *

On June 28, 1995, the licensee notified the NRC pursuant to 10 CFR 50.46 that effective May 31, 1995, the limiting PCT was 2006 OF in accordance with the updated Siemen's analysis for LBLOCA injection phase. The PCT of 2102 IF, that occurred during the SI system transfer to cold leg recirculation, henceforth, became the limiting PCT value but the licensee did not recognize that fact at the tim *

On October 14, 1996, the licensee notified the NRC pursuant to 10 CFR 50.46 that the limiting PCT was 21570 F, due to an LBLOCA. The licensee solely focused on the injection phase of a LBLOCA and recognized that the new PCT of 21570 F was greater than the previous value of 21340 F, but not that it was also greater than the existing worst case PCT value of 21020 F that occurred during recirculation phase of a LBLOC *

During the period from May 31, 1995, to October 11, 1996, the limiting PCT was associated with the SI system transfer to cold leg recirculation after a LBLOCA. A UFSAR change request approved on November 30, 1995, did not indicate that the limiting PCT occurred during that transfer to cold leg recirculatio Based on its review of the correspondence associated with this item, the licensee concluded that the 10 CFR 50.46 notification of June 28, 1995, due to a calculated PCT change greater than 50 OF, was inappropriate because the limiting PCT of the transfer to cold leg recirculation analysis (2102 oF) did not represent a 50 OF change from the previously reported LBLOCA value. The licensee also stated that the PCT values calculated between May 31, 1995, and October 14, 1996, should have been included in annual reports as required by Section 10 CFR 50.46(a)(3)(ii).

The team identified this item as Unresolved Item 50-261/97-201-0 The licensee reviewed this inappropriate application of 10 CFR 50.46 reporting requirements and concludea that it was not a reportable event. Therefore, no additional NRC submittals or notifications on this event were planned. The licensee determined that the annual reporting requirements were not properly implemented due to a breakdown in communication Subsequent to the inspection, during a conference call with the NRR staff, the licensee agreed that it had been remiss in its compliance with 10 CFR 50.46 reporting requirements. The licensee did not report each significant PCT change for the most limiting transient of each evaluation mode or its applicatio Section 10 CFR 50.46(a)(3)(i) requires the licensee to determine if a change or error in an ECCS evaluation model significantly changes the PCT. The licensee must determine if there has been a greater than 50 OF change between the two most recently calculated PCT values for the limiting transient of an acceptable evaluation model or in the application of such a mode Section 10 CFR 50.46(a)(3)(ii) requires the licensee to report to the NRC within 30 days of discovery of a significant PCT change and to provide a schedule for submitting-a reanalysis of this ECCS model application or performing a similar required actio *

The licensee reported significant changes between the previous recorded maximum PCT value and a new calculated PCT value only for the large break model. The licensee incorrectly interpreted 10 CFR 50.46 to require reporting only one limiting transient for all ECCS evaluation models. The licensee believed the ECCS analyses were large break limited since the LBLOCA typically resulted in the most severe PCT values. The team, with the support of the NRR staff, believed the correct interpretation of 10 CFR 50.46 is that there is one limiting transient for each application of an ECCS evaluation model. The licensee did not report to the NRC when the PCT value of 2102 OF calculated in the Siemens 1994 analysis for the transfer to cold leg recirculation after a LBLOCA resulted in an increase of 111 OF from the previous PCT of 1991 OF calculated in the previous Westinghouse analysis. Section 10 CFR 50.46 requires the licensee to report to NRC within 30 days of discovery each significant PCT change between the two most recent calculated PCT values for the most limiting transient of each ECCS evaluation model or its applications. The team identified this item as Unresolved Item 50-261/97-201-0 The licensee concluded that Siemens' use of a different methodology from the Westinghouse analysis did not represent a USQ because the Siemens analysis was performed with an NRC approved methodology and the NRC was aware of the second peak in the PCT as calculated by Westinghouse in 1991. The licensee stated that there was no reduction in the margin of safety because the limiting PCT before the inappropriate 10 CFR 50.46 letter was 2134 OF. If the transfer to cold leg recirculation had been properly considered, the reported value would have been 2102 OF, a decrease in the PC *

Inspection Report (IR) 91-20 identified the issue of acceptability of the PCT transient during transfer to cold leg recirculation for both a LBLOCA and a small break LOCA (SBLOCA). The maximum PCTs stated for a LBLOCA and a SBLOCA were 1400 OF and 1936 OF, respectivel IR 94-19 stated that the concern had been resolved but did not give the specifics. Due to the wide margin between the PCT of 1400 OF referred to in IR 94-19 and the PCT of 2102 OF in the Siemens analysis, the team was concerned as to whether the licensee has been operating in accordance with 10 CFR 50.46(b)(5) regarding long-term cooling. The team identified this item as Inspector Followup Item 50-261/97-201-08 and referred it to the NRR staff for evaluation and resolutio As a result of the CR 97-01023 evaluation, the licensee identified additional corrective actions: (1)

to update UFSAR Section 15.6 to reflect a second peak in fuel temperature during the transfer to the cold leg recirculation phase of the accident, and to indicate the second peak's value; and, (2)

to develop a procedure to control internal notification of site Reactor Engineering and site Licensing of changes to the calculated PCT to meet the reporting requirements of 10 CFR 50.4 E1.3.2.2(b) Refueling Water Storage Tank The team reviewed the available licensing, design, and operations documents related to the capability of the RWST to provide an adequate water supply to the SI system under accident conditions. With the exception of the lack of an analysis supporting the RWST level setpoint of 9 percent used in End Path Procedure EPP-9, "Transfer to Cold Leg Recirculation," Revision 20, the

team found that the RWST design was consistent with the applicable licensing, design, and operations documents, and that the tank was capable of performing its function under accident condition End Path Procedure EPP-9, "Transfer to Cold Leg Recirculation," Revision 20, directed the operators to operate one SI pump and, if required, up to one CS pump from the RWST under accident conditions until the tank reached an indicated level of 9 percent. The team asked if the 9 percent level setpoint value included adequate margin to accommodate the calculated instrument channel uncertainty from calculation RNP-1/INST-1023, "Refueling Water Storage Tank Level Indication Accuracies," Revision 0, and potential vortexing in the tank above the nozzle. The licensee stated that it could not locate a setpoint calculation of an allowance for vortexing above the RWST nozzle. During the inspection, the licensee provided an informal calculation, (not reviewed or issued) which indicated that the 9 percent setpoint value included adequate margin to accommodate the calculated instrument channel uncertainty as well as potential vortexing in the tank above the nozzle. The licensee initiated CR 97-00988 to formally issue the calculatio The licensee originally did not evaluate the effect of the calculated instrument uncertainty dnd potential vortexing above the nozzle on the 9 percent level setpoint used in End Path Procedure EPP-9. The team identified this item as part of Unresolved Item 50-261/97-201-0 E1.3.2.2(c) Net Positive Suction Head for SI, RHR, and CS Pumps The team reviewed the available licensing, design, and operations documents related to the required and available NPSH of the SI, RHR, and CS pumps operating under accident conditions for both the injection phase and the recirculation phase of the SI system. In both cases the available design and licensing information appeared to be based on original Westinghouse design calculation UFSAR Section 6.3.2.2.3 (page 6.3.2-3) addresses the results of an analysis of the available and required NPSH for the SI, RHR, and CS pumps during the post-accident injection phase when the SI system is lined up to the RWST. This section lists the "worst case flows for determining the NPSH requirements," and stated that the injection phase would be terminated just before the RWST level decreased to the point at which the available SI pump NPSH was reduced to the required NPSH of 25 feet at the runout flow of 600 gpm. This NPSH limit appeared to be the basis of the 27 percent RWST level setpoint used as an entry condition for End Path Procedure EPP-9, "Transfer to Cold Leg Recirculation," Revision 20. The UFSAR description was based on three SI pumps operating at a flow of 600 gpm each. These values appeared to be based on original Westinghouse design calculation results summarized in WCAP-12070. The basis of the current SI system design was that only two of the three SI pumps were in service. The licensee stated that the system flow analysis indicated that a single SI pump could deliver up to approximately 640 gpm under post-accident conditions. The team asked the licensee to verify that the pump NPSH values and the maximum flow values presented in the UFSAR were the most limitin UFSAR Section 6.3.2.2.3 (page 6.3.2-4) addresses the results of an analysis of the available and required NPSH for the RHR pumps during the post-accident recirculation phase of the SI system from the containment building sump. This section states that, with a flow of 3750 gpm, the available RHR pump NPSH with 1.5 feet of water on the containment floor was 19 feet. These

values appeared to be based on original Westinghouse calculation results summarized in WCAP 12070. The team noted that the licensee's evaluation of NRC Information Notice 96-55,

"Inadequate Net Positive Suction Head of Emergency Core Cooling and Containment Spray Pumps Under Design Basis Accident Conditions," dated October 22, 1996, had not been completed at the time of the inspection. This Action Item Assignment, Project 96-03128, due date had been extended twice due to higher priority work tasks. The team asked the licensee to verify that the pump NPSH values and the maximum. flow values presented in the UFSAR are the most limitin In response to the team's questions, the licensee initiated an analysis during the inspection to verify that the SI, RHR, and CS pumps would have adequate NPSH under the most limiting conditions. The results of this analysis were not available during the inspection. The licensee submitted a report pursuant to Section 10 CFR 50.72 on June 27, 1997, that indicated potential problems with the NPSH available for the SI pumps and had already raised the RWST level to correct any potential problems before they were identifie UFSAR Section 6.3.2.2.3 presents the NPSH requirements for the SI, RHR, and CS pumps but the licensee was not able to demonstrate that they were the most limiting values or that the most limiting values were analyzed to verify that required NPSH was available. TS 3.3.1.1.c. and stipulate that two SI and two RHR pumps are to be operable without entering into limiting condition for operation (LCO) action statement The hydraulic analysis was completed in July 1997, and the licensee concluded (as verified in LER 97-08) that prior to raising the RWST level setpoint that Sl pumps B and C had insufficient NPSH while pump A had adequate NPSH during a LBLOCA. The team has not reviewed this analysis along with its potential impact on the past and present operability of the SI and RHR pumps. The team identified this item as Unresolved Item 50-261/97-201-0 E1.3.2.2(d) 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 Design Basis Differential Pressure Reports DP-002-SIS, "Motor Operated Valves in the Safety Injection System," Revision 2; and DP-003, "Motor Operated Valves in the Residual Heat Removal System," Revision 0. These reports provided maximum operating pressure and maximum differential pressure data for motor-operated valves in the SI and RHR systems. The team found the inputs, assumptions, methodology, and results of these reports to be appropriat End Path Procedure EPP-9, "Transfer to Cold Leg Recirculation," Revision 19, provided the required direction for transferring the SI system to the recirculation mode under post-accident conditions. In accordance with step 39 of EPP-9, steps 40 through 45 must be completed within 3 minutes to avoid damaging the fuel. Step 42 of EPP-9, Revision 19, directed the operators to verify that RHR heat exchanger discharge valves RHR-759A and B were closed. The procedure did not provide any specific direction to the operators in the event that one of these valves failed to close due to a single active failure. Under post-LOCA conditions, a failure of either RHR-759A or B to close could make one train of the SI system inoperable. The operator would be required to use the other train to establish "piggyback" operation. These valves might not be accessible for local operation under post-accident conditions because of radiological conditions. The licensee

'evaluated this condition and agreed that additional direction would be appropriate but was not required. Operations Procedure OMM-022, "Emergency Operating Procedures User's Guide,"

Revision 10, Step 5.1.1.1 states that if a contingency is not provided in left-hand column then

.

proceed to next action step in right-hand column. Subsequent steps in Revision 19 of EPP-9 clarified the basis for Step 42 and the operator would have had the insight to proceed in that procedure even for a single failure of one of the RHR heat exchanger discharge valves. Revision 20 of EPP-9 was issued on April 28, 1997 to provide the required directio The team found that the remainder of 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 E1.3.2.2(e) 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 Technical Specifications; the applicable surveillance test procedures (OSTs and Engineer Surveillance Tests (ESTs))

Technical Management Manual Procedure, TMM-004, "Inservice Inspection Testing," Revision 45; and Plant Program PLP-025, "Inservice Inspection Program," Revision 9. The team compared the acceptance criteria of the applicable surveillance test procedures with the design basis requirements of the SI system mechanical components and found the values to be consisten Overall, 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 SI system components are capable of performing their required

functions under accident condition However, the team identified three apparent inconsistencies between the SI system design and the Inservice Inspection Program. These inconsistencies were the lack of inservice inspection testing of accumulator fill valves SI-851A, B, and C; the lack of a formal engineering evaluation to justify not leak-testing containment sump isolation valves SI-860A and B and SI-861A and B; and inconsistent documentation regarding the requirement to leak test RWST isolation valves SI-864A and B and SI-856A and When an SI accumulator is filled by the SI system in accordance with Operating Procedure OP 202, "Safety Injection and Containment Vessel Spray System," Revision 42, a flow path is established from the operating SI pump to the accumulator through valves SI-869 and SI-851A, B, or C. Flow calculation RNP-M/MECH-1556, "Safety Injection/Residual Heat Removal Hydraulic Model for RNP," Revision 2, determined that the minimum required SI flow would not be available to the RCS if an SBLOCA occurred while the SI system was aligned to fill an accumulator. To ensure the operability of the SI system under these conditions, OP-202 directed a dedicated control room operator to close valves SI-869 and either SI-851A, B, or C if a SI signal occurred while the SI system was aligned to an accumulator. Valve SI-869 was included in the Inservice Testing program to verify its capability to close. However, if valve SI-869 failed to close due to a single active failure, closing valve SI-851A, B, or C was required to assure system operabilit UFSAR Section 3.1.1.2.7 states that engineered safety features (ESF) systems, including the SI system, were designed to withstand any single failure of an active component. The failure of valve SI-869 to close would be a credible single failure of an active component, requiring valve SI 851A, B, or C to be closed. TMM-004 excluded any surveillance testing for valves SI-851A, B,

and C because the licensee believed their only function to permit refilling of the accumulators, was non-safety related, which appeared not to be the case. The licensee committed to meet the 1986 Edition of American Society of Mechanical Engineers (ASME)Section XI as stated in TMM-00.

Section 10 CFR 50.55a(f) requires that safety-related valves be tested in accordance with ASME Section XI of Boiler and Pressure Vessel Code to verify their capability to perform the required safety functions. The licensee stated that these valves would be added to the IST inservice testing program. The team identified this item as Unresolved Item 50-261/97-201-1 Valves SI-860A and B and SI-861A and B are 14-inch motor-operated gate valves in the two lines from the containment building sump to the suction of the RHR pumps. During the SI system injection mode, these valves would remain closed to isolate the containment building sump from the SI system. During the SI system recirculation mode, these valves would be open to provide a flow path from the containment sump to the RHR pump suctions. TMM-004 did not include leakage testing for these valves and classified these valves as Category B. Category B is defined to contain valves for which seat leakage in the closed position is inconsequential for fulfillment of their function. The team asked if the licensee had verified that leakage through these valves would be inconsequential during the SI system injection mode so that these valve-did not require leakage testing. The licensee stated that no formal evaluation existed to address the potential radiological consequences of valve leakage under accident conditions and initiated CR 97-01218 to evaluate this condition. During the inspection, an evaluation based on normal leakage rates through 14-inch gate valves was performed. This evaluation determined that, due to dilution of the source term in the SI system piping and the shielding effect of the fluid in the piping and of the SI system pipe wall, the impact of valve leakage on the post-accident dose rates in the auxiliary building would be negligible and this valve leakage would not render areas inaccessible for the post-accident manual actions required by End Path Procedure EPP-9, "Transfer to Cold Leg Recirculation," Revision 20. The evaluation also concluded that significant leakage to the vented RWST, potentially resulting in radiological releases to the atmosphere, would be highly unlikel In addition, the licensee stated that these valves were tested as part of the containment integrated leak rate test (ILRT). Based on this evaluation, the licensee verified that these valves did not require leakage testin Valves SI-864A and B are 16-inch motor-operated isolation valves in the common header from the RWST to the SI, RHR, and CS pump suctions. Valves SI-856A and B are 2-inch air-operated isolation valves located in the minimum flow recirculation line from the SI pump discharge to the RWST. These valves would be isolated before initiating the SI system recirculation mode under post-accident conditions. NRC IN 91-56, "Potential Radioactive Leakage to Tank Vented to Atmosphere," addressed the issue of potential unmonitored radiological releases due to leakage through these valves at plants that had not classified these isolation valves as Category A in their inservice testing program. Category A was defined as valves for which seat leakage is limited to a specific maximum amount in the closed position as a condition of fulfilling their function. Both SI-864A and B and SI-856A and B were periodically leak-tested by Engineering Surveillance Test Procedure EST-140, "Leak Test SI-864A & B and SI-856A & B (Refueling)," Revision 1. The results of the EST-140 testing were used to demonstrate compliance with TS 4.4.3c. However, TMM-004 did not include references to the leakage testing performed under EST-140 for these valves. In addition, TMM-004 classified these valves as Category B, not Category A. Category B was defined as valves for which seat leakage in the closed position is inconsequential for fulfillment of their function. The team noted this inconsistent documentation and asked the licensee to verify that the inservice testing program. classification was correct. The licensee stated

that leakage testing of these valves will be evaluated to determine if it should be included in the inservice testing program. The team identified this item as part of Inspector Followup Item 50-261/97-201-1.

Based on the two apparent inconsistencies between the SI system design and the inservice testing program discussed above, the team noted that the basis for including valves in the inservice testing program and for classifying valves as Category A or Category B was not well documented. The licensee stated that a Program Basis Document for Inservice Testing was being developed at the time of the inspection. The scope of the Program Basis Document included the basis for the initial component selection methodology, the inclusion/exclusion basis for all ASME Code Class 1, 2, and 3 components, and their testing/inspection frequencies. The licensee stated that the above discrepancies would be corrected E1.3.2.2(f) SI System Modifications The team reviewed four design modifications and one engineering service request for the mechanical portions of the SI system. The team identified no concerns with Modifications M-792,

"Modification to Valve SI-860A&B; SI-862A; Sl-863A&B; SI-865C; Sl-891C&D; and RHR-759A&B,"

Revision 0; M-981, "RHR-744B Valve Reorientation," Revision 0; or M-1134, "Install Permanent Strainers in SI Pump Recirculation Lines," Revision 0. The team found that the problem identification and justification for the modifications were clearly stated, the 50.59 evaluations were correct, post-modification testing was complete and appropriate, and the required plant documentation had been updated to reflect the modification Engineering Service Request (ESR) 9600012, "Replace Motor Pinion and Worm Gear Shaft

. Gear," Revision 0, increased the stroke times for motor-operated valves RHR-744A and B from approximately 10 seconds to 16 seconds. UFSAR Section 6.3.2.2.12 stated that valves that receive an SI signal were supplied with 10-second operators. These valves received an SI signal to automatically open and provide a flowpath from the RHR pumps to the RCS cold leg nozzle The team noted that the stroke times resulting from the modification were not consistent with the UFSAR criteria of less than 10 seconds and asked if these stroke times were acceptable. The licensee stated that this condition had been evaluated before the inspection but after the closeout of ESR 9600012. The increased stroke times were deemed acceptable by reexamining the applicable accident analysis and did not result in a USQ. However, when the valve stroke times were increased from approximately 10 seconds to 16 seconds by ESR 9600012, the associated 10 CFR 50.59 screening evaluation incorrectly stated that the change did not affect the UFSA The licensee initiated a change to correct the UFSAR and issued CR 97-01261 to address the 10 CFR 50.59 screening deficiency. Section 10 CFR 50.59(b)(1) requires licensees to provide a written safety evaluation that provides the basis that a change, test, or experiment does not involve a USQ. The team identified this item as Unresolved Item 50-261/97-201-1 Modification M-1017, Revision 0, eliminated the potential.for loss of both RHR pumps because of common mode flooding in the RHR pump pit. This modification rerouted the CCW pipes to the RHR pump coolers and the SW pipes to the pit coolers so that the manual isolation valves would be accessible during post-LOCA recirculation and added check valves in the CCW and SW return piping from these components. This modification provides the ability to isolate a CCW or SW pipe leaking in the pump pit. The team observed that the added check valves (CC-931, CC-926, SW 911, and SW-924) were periodically reverse-flow-tested in accordance with TMM-04, "Inservice

Insoection Testing," Revision 44; however the manual isolation valves on the inlet piping to the RHR pump pit (CC-927, CC-928, SW-75, and SW-77) were identified as passive valves in TMM 04 and not tested. Passive valves were defined in TMM-04 as those that do not perform a mechanical motion to accomplish a safety function. The team questioned the passive classification of these manual valves because they would be closed to perform the safety function to prevent common mode failure of the RHR pumps. The licensee is committed to the 1986 Edition of ASME Section XI. It appears that the requirements of 50.55a, to perform inservice testing in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, were not implemented for these isolation valves. Section IWV-1 100 of this code required valve testing in accordance with ASME/ANSI OM (Part 10); and OM-10 required testing of valves which perform a specific function in mitigating the consequences of an accident. Valves CC-927&928 and SW 75&77 performed such a function by ensuring the availability of the RHR system and performed this function as active valves. The licensee issued CR 97-01129 to determine if an ASME Section XI category change was needed for these valves. The team identified this item as part of Inspector Followup Item 50-261/97-201-1 The team found that the remaining aspects of the SI system modifications reviewed were consistent with the applicable licensing, design, and operations document E1.3.2.2(g) 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 RWST area, the SI pump room, the RHR heat exchanger room, the pipe alley, and the control room. The system configuration was found to be consistent with the SI flow diagrams in the areas walked down except for several CCW system vents and drains in the area of the CCW pumps; these vents and drains had caps and test connections (e.g., CC-853A, 854A, 855A) that were not shown on the CCW system flow diagram, 5379-376. The licensee had issued ESR 97-00183 in March 1997, to resolve the flow diagram discrepancie Both the SI and CCW pumps were located on the ground floor of the reactor auxiliary building (RAB), as were several SW pipes. The team asked how a SW system passive failure was controlled so that flooding would not disable safety-related equipment in the RAB. The licensee had no response and this is identified as Inspector Followup Item 50-261/97-201-1 E1.3.2.3 Conclusions Overall, the team found that the mechanical portion of the SI system was capable of providing adequate emergency core cooling flow under accident conditions, subject to confirmation by the NPSH calculations initiated during the inspection. The licensee confirmed inadequate available suction head for SI pumps B and C when supplied by the RWST during a LBLOCA prior to raising the RWST level setpoint subsequent to this inspection. The SI system performance information presented in the accident analysis sections of the UFSAR was consistent with the system design and operations document The only available design and licensing information regarding the SI system pump NPSH was based on original Westinghouse design calculations that were not under the licensee's contro The inability to revise those original design basis analyses hindered the licensee's capability to fully evaluate the impact of design and operational changes on systems required to perform safety-related function The licensee did not abide by the reporting requirements of 10 CFR 50.46 for significant PCT changes, and may be outside of the long term cooling requirements of 10 CFR 50.46(B)(5) during transfer to cold leg recirculation. The uncertainty calculation for RWST level instruments did not account for seismic considerations, NPSH analyses for SI, RHR, and CS pumps did not analyze the most limiting conditions, several SI system valves were inappropriately omitted from the inservice testing program, and the licensee did not perform a 10 CFR 50.59 evaluation for a modification that replaced the operators of MOVs RHR-744A & B even though their new stroke times differed from those in the UFSA E1.3.3 Electrical Design Review E1.3.3.1 Inspection Scope The team evaluated the electrical loads required for the SI system to perform its design function This evaluation addressed electrical bus loading, DC battery loading, protective coordination, cable sizing, and modification package E1.3.3.2 Observations and Findings

.

E1.3.3.2(a) SI System Electrical Distribution To verify SI system electrical loads on safety buses, the team reviewed calculations RNP-E 7.002, "Emergency Equipment Load Factor Study," Revision 5; Calculation RNP-E-8.016,

"Emergency Diesel Generator Static and Dynamic Analysis," Revision 5; and RNP-E-8.002, "AC Auxiliary Electrical Distribution System Voltage/Load Flow/Fault Current," Revision 3. All major SI system electrical loads were accounted for within these calculations for both normal and accident conditions. The motors were sized to accelerate the SI and RHR pumps within the required accident time lines and to drive them for long-term continuous operation. The team determined that the methodology and assumptions used were appropriat Calculations RNP-E-5.004, "Ampacity Evaluation of Safety Related Power Cables on 480V and 208V AC MCC's and Buses," Revision 4; RNP-E-2.001, "Overcurrent Protection for Au Feedwater Pumps A and B Motors," Revision 2; and RNP-E-8.045, "AFW, SIP, RHR, and CCW Pump Motor Characteristics Impacts due to High Ambient Temperatures," Revision 0, were also reviewed by the team, which determined that the sizes of installed SI cables, the overcurrent protection for SI electrical loads, including cables, and the operating characteristics of major SI motors conformed to industry standard The SI system consists of three pumps (A, B, and C), with pumps A and C supplied by trains A and B, respectively. The B pump could be powered from either train and is maintained in a standby state (with supply breakers open and control fuses removed). The B pump, when

activated, would be powered by the same electrical source as the pump (A or C) being taken out of service. The team verified that the one-line configuration for the B pump permitted the train A and B power sources to be fully independent and redundant for all operating condition Administrative controls exist to ensure that only one supply breaker was closed when the B pump was in us The team requested the licensee to verify that the redundant system cables for the A, B, and C SI pumps were adequately separated as defined in UFSAR Section 8.3.1.3 and DBD/R87038/SD62,

"Cable and Raceway System," Revision 0. These documents specified that these cables are to be separated into channels using separate trays, trays with metal barriers, or separate conduit During the review of the cable routings, the licensee determined that the SI pump C autostart (on a SI signal) cable, C2239C, and manual start (from the RTGB) cable, C2239D, were routed in the same tray as the SI pump A SI autostart cable C2237D and the SI pump B (train A line up) SI autostart cable C2891C. The licensee also determined that the SI pump C SI autostart cable C2239D did not contain a raceway routing on the Cable and Conduit List (drawing B190634, sheet C2239A) and the licensee could not determine if the required separation requirements of DBD/R87038/SD62 had been met. The licensee declared SI pump C inoperable tier TS 3.3.1.2, placed SI pump B in service, and reported this event to the NRC on May 21, 1997 as required by 50.72. The licensee issued CR 97-01177 to track this concern upon initial discovery and implemented ESR 97-00274 to replace and reroute cables C2239C and C2239D. The licensee stated that the validity of the present routing of these cables had probably been confirmed in the past and that it was looking for the appropriate documentation. However, the licensee could not retrieve that documentation and after the inspection stated that, in all likelihood, it would not be found. It appeared that conditions adverse to quality had not been identified by the licensee and therefore the appropriate corrective actions had not been undertake S In the letter dated March 17,1989, to E.E. Utley, CP&L, from Hugh L. Thompson Jr, NRC, the NRC requested the licensee to provide assurance that HBR complied with single worst failure requirements of Section 10 CFR 50.46 and Appendix K of 10 CFR Part 50. In a letter dated May 19, 1989, to the NRC from R.A. Watson, CP&L, the licensee committed that the ECCS at HBR met the single failure criterion of Appendix K of 10 CFR Part 50 and the core cooling-performance requirements of Section 10 CFR 50.4 The licensee has not adhered to that commitment in regard to the single failure requirements of Criterion D.1. "Single Failure Criteria" of 10 CFR Part 50, Appendix K. The potential existed in the past when SI pumps A and C were inservice that for a single mechanical or electrical failure that affected their autostart cables that initially the accident mitigation function of the SI system for a LOCA could be hampered until SI pump B was started manually from the RTGB. During the initial minutes of an accident, the emergency safeguards systems are designed to automatically respond and not depend on the prompt action of an operator to manually actuate the one operating SI pump minimally required in accordance with the accident analysis to meet the ECCS requirements of 10 CFR 50.46. The routing of the control cables for SI pumps A and C in the same tray should have been identified during the period when the licensee reconfirmed their commitment to single failure criterion of Appendix K. In addition this also represents a weakness with Criterion XVI,

"Corrective Actions" of 10 CFR Part 50, Appendix B, because of the licensee's apparent ineffective correction actions in accordance with their recommitment to 10 CFR 50, Appendix K and also during their efforts to reconstitute their electrical design basis. However, the presiding weakness is with respect to Criterion III "Design Control" of 10 CFR Part 50, Appendix B in that the design basis for electrical separation was not properly implemented into work instructions

dictating the correct routing of these control cables. The team identified this item as Unresolved Item 50-261/97-201-1 O The team determined that a 1.0 service factor was used for SI pump motors A, B, and C in calculation RNP-E-5.004, even though the motor load used in calculation RNP-E-8.016,

"Emergency Diesel Generator Static and Dynamic Analysis," Revision 5, appeared to exceed the motor nameplate rating. Investigation revealed that the three SI pump motors could operate in excess of their nameplate rating of 350 horsepower because the actual service factor was 1.1 On May 21, 1997, the licensee initiated an operability determination that determined the affected components were operable. CR 97-01194 was.issued to track this condition, and ESR 97-00276 was issued to evaluate the effects on the motors, cables, breaker protection, and supporting systems. The team identified this item as Inspector Followup Item 50-261/97-201-1 E1.3.3.2(b) Electrical SI Modifications Modifications M-947, "SI Pump Availability Upgrade," and M-951, "SI Pump B - Deletion of Autostart," were reviewed for their technical adequacy. The team identified no concerns with either modification and observed that the safety evaluations were properly performe E1.3.3.2(c) SI System Walkdown The team performed a walkdown of the SI system. The team verified that nameplate loadings were used in the plant design analyses. The reactor auxiliary building area containing the system was clean of debris and the raceway systems were properly identified as stated in DBD/R87038/SD6 During the walkdown review of the emergency switchgear area, the team verified that both 480V AC breakers feeding the B SI pump motor were in the disconnected position as required while the B SI pump was in the standby position. The team questioned the seismic qualification of these breakers in the disconnected position. The licensee stated that the seismic qualification of these breakers in both the disconnected (racked out) and removed (physically out of the cabinet on rails) was identified as an outlier to the seismic adequacy verification of electrical equipment and was the subject of CR 96-3086. This verification, which implemented the Seismic Qualification Utility Group (SQUG) guidance, was conducted in response to NRC Generic Letter 87-02,

"Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46."

Based upon the licensee's review, the breakers were seismically qualified in the disconnected position as described in WCAP 7397-L, "Seismic Qualification of Electrical and Control Equipment." The licensee stated that the breakers had not been seismically qualified in the removed position and that either a modification would be installed to ensure the seismic adequacy of the breakers in this position or an analysis would be performed to demonstrate seismic adequacy in the removed position. The team identified this item as Inspector Followup Item 50-261/97-201-1 E1.3.3.3 Conclusions To address a deficiency identified by the team, the licensee initiated action to reroute the SI

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pumps' control cables to meet electrical separation requirements. Therefore, the electrical system design for components that perform the normal and accident functions of the SI system supports the design basis functions of the system and provides independent, redundant, safety-related power to the electrical SI loads. The team verified that the SI and RHR pumps' motors were sized in accordance with system demands and industry guidelines. The basis for determining the ampacity and overcurrent protection of installed SI cables seemed appropriate. The team verified that the one-line configuration for the B pump permitted the train A and B power sources to be fully independent and redundant for all operating conditions. The team concluded that the licensee did not properly address the single failure considerations of 10 CFR 50, Appendix K by the routing of the SI pump A and C autostart cables in the same tray and also did not take effective corrective actions to eliminate any such single failure issues based on previous commitments to the NR The team identified concerns with whether the current carrying components of the SI pump motors were sized using the appropriate service factor of 1.15 for the SI pump's motors and whether 480 volt breakers had been seismically qualified for the disconnected positio E1.3.4 Instrumentation and Controls (I&C) Design Review E1.3.4.1 Inspection Scope The team evaluated the ability of the instrumentation and controls for the SI system to perform their design safety functions. The team reviewed sections of the UFSAR, TS sections, calculations, modification packages, the SI system DBD, flow diagrams, control wiring diagrams, and plant procedures. System walkdowns were also conducted. An in-depth evaluation was performed of the RWST level instrumentation, SI pump actuation circuitry, instrumentation configuration, containment sump level instrumentation, and instrument loop uncertainty calculation E1.3.4.2 Observations and Findings E1.3.4.2(a) Accumulator Pressure Relief Valve Setpoints The team evaluated the accumulator relief valves and pressure alarm setpoints. Each accumulator has redundant pressure instrumentation loops with high and low pressure alarm The alarms alert the operator when the accumulator pressure either approaches the minimum pressure TS limit or the relief valve setpoint pressur The pressure relief valve setpoint and its calibration tolerance were compared with the current high pressure alarm setpoint and its calibration tolerances. The team was concerned that the relief valve might actuate before the control room was alerted to a accumulator high pressure conditio The setpoint for the high pressure alarm was 680 psig, with a tolerance of -10 psig; and the setpoint for relief valve SI-858A was 700 psig +/- 21 psig. Comparing the lower limit of the relief valve setpoint (679 psig) to the high limit of the high pressure alarm setpoint (680 psig), the team observed that the relief valve could actuate before the alar The original plant design established the high pressure alarm at 665 (-10 psig) in accordance with WCAP-12070, "SIS Calc Summaries." Modification and Setpoint Change # 14, dated January 28, 1972, revised the high pressure alarm setpoint from 665 psig to 680 psig for accumulator tanks A, B, and C. The licensee initiated this corrective action to change the setpoint value and created the existing overlap between the calibration tolerances of the pressure relief valve setpoint and the high pressure alarm setpoint. The licensee was unable to provide a setpoint calculation or an uncertainty calculation to support the current setpoint Although the team identified no safety concern, a rising pressure in an accumulator might indicate a leaking nitrogen header control valve. The licensee issued CR 97-00964 to evaluate the alarm setpoin Corporate Quality Assurance Manual, Revision 18, Section 3.4.3, requires that design verification be performed to substantiate final design documents meet the appropriate design inputs. This issue represents a weakness with respect to Criterion III, "Design Control" of 10 CFR Part 50, Appendix B. The licensee also changed the high pressure alarm setpoint without considering the impact on its primary function of indicating a high pressure condition. This indicates ineffective corrective action in that the overlapping of the relief valve and high pressure alarm setpoints'

tolerance ranges was not detected and corrected. This demonstrates a weakness with Criterion XVI, "Corrective Actions" of 10 CFR Part 50, Appendix B. The team identified this item as Unresolved Item 50-261/97-201-1 E1.3.4.2(b) RWST Level Instrument Uncertainty Calculation RNP-l/INST-1023, "Refueling Water Storage Tank Level Indication Accuracies,"

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Revision 0, addressed the RWST level instrument channel uncertainty for design basis earthquake (DBE) conditions. However, the applicable EOP setpoint analysis, calculation RNP-I/INST-1 111,

"RWST Level EOP Setpoint Parameters," Revision 0, used an uncertainty value that did not include DBE effects. The failure of the RWST level instruments to accurately indicate RWST tank level during the SI system transfer to cold leg recirculation could potentially result in insufficient NPSH for the SI pumps. The team asked why the DBE channel uncertainty value was not used for the RWST level setpoints. The licensee stated that the DBE channel uncertainty value was not included in the Emergency Operating Procedure (EOP) setpoint analysis because the RWST level instruments perform their only safety functions during and after a LOCA, and that the plant licensing basis does not require a seismic event to be postulated in conjunction with a LOC UFSAR Section 3.1.1.2.1 states that all Class I systems and components were designed so that there would be no loss of function in the event that the maximum hypothetical ground acceleration acted in the horizontal and vertical directions simultaneously. The RWST level instruments are Class I and provided operators key information on when to initiate SI system transfer from injection to cold leg recirculatio Appendix A of Part 100 requires that if the safe shutdown earthquake occurs, certain structures, systems, and components will remain functional. These structures, systems, and components are those necessary to assure "the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the

guideline exposures of this part." This regulation also states that, in addition to seismic loads, applicable concurrent functional and accident-induced loads shall be taken into account in designing these safety-related structures, systems, and component RWST level was classified as a RG 1.97, Type A, Category 1 variable by the licensee in their RG 1.97 submittal to the NRC, which required it to be continuously indicated and seismically qualifie Engineering procedure, EGR-NGGC-0153, "Engineering Instrument Setpoints," Revision 2, Section 9.4.10 requires the uncertainty term, specified by the vendor, to be included in a uncertainty calculation when the instrument is used in an application requiring seismic qualification. Corporate Quality Assurance Manual, Revision 18, Section 3.4.3 requires design verification be performed to substantiate final design document meets the appropriate design input Criterion Ill, 10CFR Part 50, Appendix B states that the design control measures shall provide for verifying or checking the adequacy of the design. The team considered the calculation's design inputs as nonconservative because of the exclusion of the seismic uncertainty term. This issue represe-ts a potential weakness with respect to Criterion III, "Design Control" of Part 50, Appendix The team has referred this item to the NRC staff to determine if the seismic uncertainty factor should have been included in the uncertainty calculation for the RWST level instrument and similar uncertainty calculations for other accident-mitigation instruments. The team identified this item as Unresolved Item 50-261/97-201-1 E1.3.4.2(c) Containment Water Level EOP Setpoint Calculation RNP-1/INST-1 109, "Containment EOP Setpoint Parameters," Revision 0, determined the containment building water level required for post-accident RHR pump recirculation operation (EOP setpoint No. M.20). This setpoint value was used in EPP-9, "Transfer to Cold Leg Recirculation," Revision 20, to verify that sufficient water was available in the containment sump before SI recirculation was initiated. The containment building water level was indicated by level instruments LI-801 and LI-802, located in the containment building sump. The containment building water level setpoint was calculated to be 354 inches, based on a required containment water level of 9 inches above the containment floor and the calculated channel uncertainty for these level instruments during normal environmental conditions. UFSAR Section 6.3.2.2.3 (page 6.3.2-4) states that the RHR pump NPSH analysis was based on a.containment building water level of 1.5 feet, not 9 inches, above the containment floor. In addition, these level instruments could be exposed to adverse environmental conditions inside containment after an acciden Therefore, the use of normal environmental channel uncertainty values for the level instruments did not appear appropriate to the team for this applicatio The licensee initiated an evaluation to resolve this issue and determined that both calculation RNP-I/INST-1109 and the channel uncertainty analysis, RNP-1/INST-1058,"Containment Water Level Instrument Uncertainty Calculation," Revision 0, for level instruments LI-801 and LI-802 were incorrect. After the inspection, the licensee stated that these calculations were revised, based on the use of adverse environmental conditions and a required containment water level of 1.5 feet, and that the revised calculations did not affect the EPP-9 setpoint of 354 inche In this case, the licensee used erroneous values for the instrument's tolerance range and the required water level above the containment floor. The team identified this item as part of Unresolved Item 50-261/97-201-03. The adequacy of the calculation in regard to the actual setpoint value and its tolerance range needs to be reviewed by the team. The team identified this item as part of Inspector Followup Item 50-261/97-201-1 E1.3.4.2(d) Instruments used in Emergency and Abnormal Operating Procedures (EOPs and AOPs)

Reviewing various instrument uncertainty and EOP setpoint analyses, the team asked if the licensee had verified that all instruments used in AOPs or EOPs to verify specified parameters were operable with an appropriate accuracy. The licensee stated that all RG 1.97, Category 1 instruments had been verified to be operable with an appropriate accuracy. However, the licensee stated that an additional review was required to identify and verify the accuracy of other instruments used in AOPs or EOPs. During the inspection, the licensee developed a list of instruments, not classified as RG 1.97 Category 1, that were used in these procedures. An initial review of EOPs, AOPs, and Westinghouse Owners Group (WOG) guidance, performed by the licensee during the inspection indicated an adequate and logical application of function and accuracy. The licensee initiated CR 97-01221 for followup review. The team identified this item as part of Inspector Followup Item 50-261/97-201-1 E1.3.4.2(e) Instrument Loop Uncertainty Calculations The team identified an inconsistency with the licensing bases in two calculations: the calculations did not include uncertainty caused by an earthquak The first calculation was RNP-1/INST-1040, "Main Steam Flow Accuracy and Scaling Calculation,"

Revision 0. Section 2.2 of this calculation stated that failure of this portion of the instrument loop could compromise the ability to mitigate the consequences of an accident, and Section 2.3 stated that these loops must function during normal and accident conditions, including during and after an earthquake. Furthermore, Section 2.6 stated that the differential pressure signal was used directly to detect a high steam flow condition and to initiate the safeguards system. However, Section 5.1.2 stated that this equipment was not designed to operate during or after an earthquak The second calculation was RNP-l/INST-1043, "Main Steam Pressure Channel Accuracy and Scaling Calculation," Revision 1. Section 2.2 of this calculation, "Safety Significance," states that failure of the steam pressure loop could compromise the ability to mitigate the effects of postulated accidents and could compromise the ability of the operator to achieve and maintain a safe-shutdown condition. Section 2.6, "General Design Basis," states that the pressure transmitters supplied signals to the safeguards system for low steam pressure and high steam line differential pressure bistables. All nine of the pressure transmitters were used for these safeguards signals. Section 5.1.8 states that the transmitters are mounted in a non-seismic building, and therefore no requirements for operation during or after a seismic event were imposed toa mitigate the effects of a postulated accident The above statements in Section 5.1.2 of RNP-1/INST-1040 and Section 5.1.8 of RNP-1/INST-1043 were inconsistent with UFSAR Section 3.10, which states that control equipment which initiates

safeguards systems must be capable of performing its functions during and after an earthquak The licensee issued CRs 97-01209 and 97-01042 to evaluate the seismic uncertainty input in these calculations. As stated in Section E1.2.4.2(c) of this report, the seismic uncertainty factor should have been included in these respective calculations. The team identified this item as part of Unresolved Item 50-261/97-201-0 Otherwise, the calculations were adequate and the methodology was consistent with the setpoint progra E1.3.4.2(f) Modifications Three modifications were reviewed: 86-0094, "Evaluate Raising the Alarm Setpoint of PC-934 From 200 psig to 800 psig"; 92-0114, "Temporary Strainer Installation on "B" SI Pump"; and 958,

"Supplement-Add Auto Start to "B" SI pump." Each modification was reviewed for adequacy and consistency with the design basis and design configuration. The 10 CFR 50.59 evaluations were also reviewe The team had no concerns with the modifications. The modifications were consistent with the design basis. No concerns were identified with the 10 CFR 50.59 evaluation E1.3.4.2(g) Walkdown During the team's walkdown of the SI system, various instrument configurations were inspecte The team observed that various instrument sensing-line runs were not consistent with industry standards or licensee requirements for sloping. The process lines of several SI flow instruments (SI-FT 940, SI-Fl 941, SI-Fl 944) appeared to have no slope. In response to the team's concern, the licensee confirmed the absence of slope by taking appropriate measurements. The inspection revealed that the absence of slope in several cases was due to original design deficiencies or damage from previous outages. Inadequate slope in the lines can impair flow readings by trapping air between the process and the transmitter and inducing errors. Drawing A-190299107, Revision 3, showed the typical installation for each instrument. The horizontal portions of the tubing were identified on this drawing as sloped from the root valves. The actual installation is inconsistent with this desig Corporate Quality Assurance Manual, Section 6.2.3 states that accomplishments of activities affecting quality shall be in accordance with approved procedures and drawings appropriate to the circumstance One of these instruments (SI-FT 940) is a RG 1.97 Category 2 post-accident flow instrument providing flow monitoring to the operators. The configuration of the tubing was found to be inconsistent with CP&L Standard Procedure EGR-NGGC-0151, "Evaluating Instrument Sensing Line Installations," Revision 0. The procedure requires a minimum slope of /4 inch per foot up to the fitting assemblies, unless a different slope had been approved. The licensee issued Engineering Service Request (ESR) 97-00257 to correct various sensing lines to restore the tubing to the plant design requirements, and issued CR 97-01005 to correct the other deficiencie It appears that the design basis for instrument tubing slope was not translated into the documents used to install the tubin The licensee has experienced similar sensing line configuration problems and gas entrapment problems with other instrumentation. NRC violation 50-261/94-17-02 identified improperly sloped sensing lines for RHR flow transmitter FI-605 which resulted in conditions conducive to the

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trapping of gases. Corrective actions for this incident addressed the proper sloping of sensing lines associated with flow transmitters with full compliance to be achieved by November 16, 199 Indicator FI-605 during injection and recirculation phases of accident mitigation for a LBLOCA was considered part of the SI system. The licensee's corrective actions investigated any flow transmitters with potential gas entrapment problems which should have included those in the SI system. Those corrective actions appear to have been ineffective in that the team identified two additional SI flow transmitters with similar problem Similarly NRC violation 50-261/95-30-01 identified improper sensing line configurations for accumulator level transmitters LT-924, 926, 928, and 930 that resulted in increased probability of entrapped gases affecting the level readings. Licensee corrective action evaluated transmitter configurations that could exhibit erroneous indication due to air or gas entrapment with compiance by February 1996. Hardware corrections were to be made as required. The licensee emphasized level transmitter configurations, because flow transmitter configurations had been previously reviewed and deemed adequat The team has identified related problems with flow transmitters in the SI system for similar reasons and contends that licensee has taken ineffective corrective actions for this issu Criterion XVI, "Corrective Actions" of 10 CFR Part 50, Appendix B, requires that measures be established to ensure that conditions adverse to quality are promptly identified and corrected. It also requires that corrective actions be taken to preclude repetition of significant conditions adverse to quality. The inadequate slope examples identified by the team indicate that corrective actions in response to the above violations have been ineffective. The team identified this item as part of Unresolved Item 50-261/97-201-0 E1.3.4.3 Conclusions The SI system initiation logic is capable of performing the design safety function. RG 1.97 requirements for various SI system instrument ranges were found to be consistent with the plant design. The slopes of several SI flow instrument sensing lines were incorrect, the seismic uncertainty factor was not included in several SI instrument uncertainty calculations, an accumulator relief valve could potentially actuate prior to the initiation of the corresponding accumulator high pressure alarm, and the instrument tolerance range for containment water level was in error by at least one order of magnitud E1.3.5 System Interface Design Review E1.3.5.1 Inspection Scope The team evaluated the ability of the CCW System to supply cooling water to the RHR heat exchangers and the RHR and SI pumps, together with other design basis aspects of the CCW system, and the ability of the ventilation system (VS) to provide acceptable ambient conditions for operation of the SI and RHR pumps. Applicable UFSAR sections, system flow diagrams, DBDs, SDs, procedures, and calculations were reviewed for each interface syste * E1.3.5.2 Observations and Findings E1.3.5.2(a) Component Cooling Water (CCW) System The team reviewed UFSAR Section 9.2.2, "Component Cooling System"; DBD/R87038/SD13,

"Component Cooling Water System," Revision 1; system description SD-013, "Component Cooling Water System", Revision 0; CCW system flow diagram 5379-376, Sheet 1, Rev. 31; Sheet 2, Re ; Sheet 3, Rev. 21; Sheet 4, Rev. 29; CR 96-01165, "Engineering Evaluation (EE) 87"; ESR 97 0014, "CV Isolation Reclassification of the Unit 2 CCW System," Revision 0; Operating Procedure OP-306, "Component Cooling Water System," Revision 23; applicable portions of procedure TMM 004, "In-Service Inspection Testing," Revision 44; and several calculation The team reviewed calculations RNP-MN/MECH-1 112, "Analysis for CCW Hx Heat Transfer with 90/10 Tubes for RNP-2," Revision 1, and RNP-M/MECH-1074, "Effect of Tube Plugging on CCW Hx and System Performance," Revision 1. These calculations used flow data from the original CCW heat exchanger data sheet as input. The team requested system flow calculations showing that the design flows were achieved, but no calculations were made available by ie license The licensee stated that the CCW requirements for the RHR heat exchanger after a LOCA were no greater than during normal cooldown since normal cooldown had been successfully performed numerous times and adequate CCW flow to the RHR heat exchangers was assured. The licensee also stated that flow to the SI and RHR pump coolers was assured in that low flow alarms consistent with the original Westinghouse design were installed and there had been no difficulty in maintaining adequate flow to these components. The team also reviewed test procedure OST 908, "Component Cooling System Component Test," Revision 34, which includes flow testing of the CCW pumps. The minimum acceptance criterion for the test is approximately 10 percent below the manufacturer's pump curve, which reflected the original system design. The licensee could not provide any calculations demonstrating that the test acceptance criterion was consistent with the design basis analysis. The licensee stated that the acceptance criterion was satisfactory because with waste evaporator A and the boric acid evaporators isolated, fewer demands are made on the system, and during cooldown less flow is required than in the original design. The team concurred that the system flow and pump test acceptance criteria were adequate for post LOCA service based on the system design margin and demonstrated performance. The team also verified that the CCW pump motors provided sufficient power to drive the pumps for the original design condition The team reviewed the NPSH for the CCW pumps. The licensee did not locate a calculation that compared required to available NPSH. The normal water level in the CCW surge tank is approximately 43 feet (at gauge pressure) above the pump suction and the team judged that greater than 70 feet (at absolute pressure) of NPSH was available. The CCW pump required approximately 12 feet of NPSH at the design flow rate of 6000 gpm, and thus significant NPSH margin existe The team noted that the CCW pump room had no dedicated room coolers and that normal auxiliary building ventilation did not function after a LOCA. The licensee had performed ESR 96 0692, "Evaluate Equipment Important to Safety in CCW Pump Room," Revision 0, to evaluate the subject equipment with no auxiliary building ventilation. This ESR concluded that all safety-related components would perform their intended safety function after a LOCA provided that all three CCW pump motors were refurbished at the end of the year 2007. The team had no concerns with the ESR and verified an open action item remained on the ESR for the CCW system engineer to

• set up a new preventative maintenance model for refurbishing the CCW pump motors. The ESR could not be closed until this action item was complete.

Westinghouse letter CPL-84-14, "Component Cooling Water System Potential Overpressurization Notification," dated July 26, 1984, contained two long-term recommendations: (1) convert the air operated CCW surge tank vent valve to a locked open valve; and (2) remove the surge tank relief valve and replace it with a section of piping, or remove the internals of the relief valve. These recommendations would achieve a low-pressure-drop overflow path from the surge tank and keep the CCW system from being overpressured. The licensee had gagged open the vent valve and left the relief valve as originally designed. The team questioned, whether, if the relief valve were functioning, the CCW system would operate satisfactory with the design basis inleakage of 260 gpm from a ruptured reactor coolant pump thermal barrier cooling coil. The team also questioned the possibility of overpressurizing the system using the demineralized or primary water systems used to supply makeup to the surge tank. These questions remained open and were identified as Inspector Followup Item 50-261/97-201-2 The backpressure for the CCW thermal relief valves (CC-747A&B, CC-774, and CC 791A,B,C,D,E,F,G,H,J,K, and L) was specified in the Westinghouse specification for auxiliary relief valves, E-Spec. No. 676257, as 10 psig. These valves relieve into the CCW return line to the pump suction. The CCW DBD stated that this pressure is slightly higher that the normal CCW pump suction pressure. However, the normal suction pressure at the pump is approximately 17 psig and could be higher at higher surge tank levels. The team questioned whether the relief setpoints of any relief valves could be affected by a backpressure greater than design and whether the applicable code requirements for the system and components were met. The licensee reviewed the elevations of the relief valves and stated that only four (CC-747A&B, CC

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774, and CC-791G) were at a low enough elevation to have a backpressure greater than 10 psig (approximately 15 psig) and the increased backpressure had no safety significance. The licensee also stated that thermal relief valves are not an explicit requirement of the applicable code and that code inquiries NI.97.008 and NI.97.011 were being addressed to clarify the code requirements. The team noted that the design basis for the thermal relief valves was not correctly translated into the drawings that controlled the as-built plant as required by Criterion III, "Design Control," of Part 50, Appendix B. This item was identified as part of Unresolved Item 50 261/97-201-2 The team reviewed calculation RNP-M/MECH-1620, "Evaluation of Effects of High Energy Pipe Ruptures on the CCWS," Revision 0. This calculation concluded that the CCW system piping inside the containment missile barrier was protected from high energy line breaks (HELBs) and was part of the basis for the conclusion of ESR 97-00014, "CV Isolation Reclassification of the Unit 2 CCW System," Revision 0, that the CCW system inside containment was a "closed" syste According to Generic Issues Document GID/90-181/00/RCI, "Reactor Containment Isolation,"

Revision 0, this classification was the basis for the acceptability of the containment isolation valve arrangement for penetrations 18, 19, and 20. The team identified the following concerns with this calculation:

The calculation only considered HELBs in piping other than Reactor Coolant System (RCS)

piping. The licensee stated that RCS piping was excluded from the calculation based on the leak-before-break piping design bases and the HELB exclusion documented in WCAP 9558, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated

. Circumferential Throughwall Crack," Revision 2; WCAP 9787, "Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation," May 1981; and NRC Generic Letter 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."

However, this HELB exclusion only applied to the reactor coolant loop piping and not to other RCS piping. The licensee reviewed the reactor coolant piping and flow diagrams and concluded that the effects of HELBs in this piping would not prevent the CCW system from performing its intended function and issued ESR 97-00269 to revise the calculation to include an evaluation of RCS branch line *

Calculation RNP-M/MECH-1620 did not document the break locations, the spatial relationship between the HELBs and the CCW piping, the possible jet impingement areas due to cracks/breaks in CCW piping, source piping motion due to the HELB forces, or existing structures and components between the HELB and CCW piping that prevent adverse effects on the CCW piping. Additionally, the calculation did not specifically address the CCW piping between the containment penetrations and the missile barrie The licensee stated that the calculation would be revised to include this documentatio *

Calculation RNP-M/MECH-1620 evaluated the jet effect on the CCW pipe by comparing the jet pressure to the pipe internal pressure and concluded that jet pressures less than internal pipe pressures were acceptable. The team observed that this criterion was incorrect because it did not consider the force of the jet on the target CCW piping and its support system. The licensee stated that the calculation would be revised to evaluate the effects of the jet forces on the CCW pipin It appears that this calculation did not meet the quality standards required by UFSAR Section 3.1.2.1 (GDC 1) and was not documented in the detail required by the "Carolina Power & Light Company Corporate Quality Assurance Program," Revision 18, Section 3.4.1 and by procedure MOD-002, "Modification and Design Control Procedure," Revision 10. This item was identified as part of Unresolved Item 50-261/97-201-0 E1.3.5.2(b) SI and RHR Pump Ventilation System The team reviewed UFSAR Section 9.4, "Air Conditioning, Heating, Cooling, and Ventilation System"; DBD/R87038/SD36, "Post-Accident HVAC Systems," Revision 2; and System Description SD-036, Revision 0, "Heating Ventilation and Air Conditioning System." The team found the documents satisfactory and did not identify any concerns with the interface between the VS and the SI or RHR pumps. The SI pump room and the RHR pump pit are equipped with two chillers supplied by the SW system. The chillers are safety-related, seismically designed, and supplied with power from the EDG. A chiller is automatically started when any SI or RHR pump is starte E1.3.5.3 Conclusions Calculation RNP-M/MECH-1620 was deficient in identifying break locations, jet impingement area due to CCW piping breaks, and potential movement of CCW piping due to cracks or breaks. The team concluded that CCW pumps had adequate available NPSH and provided sufficient system flow for post LOCA requirements, that the present functional state of the CCW surge tank relief valve increased the possibility of overpressurizing the CCW system, and that the safety related

, rooM chillers assist the CCW system in carrying out its safety functio E1.3.6 UFSAR and System Description Review The team identified the following discrepancies in the UFSAR:

The licensee had not evaluated, and incorporated into the UFSAR, the effect on the SI, RHR, and CS pumps' NPSH due to system design and operational changes, such as requiring only two of the three Sl pumps be in service, modifying the RHR pump minimum flow recirculation line, and realigning system valve *

UFSAR Table 6.4.2-1 did not indicate that the discs of containment isolation valves SI-860A and B or SI-861A and B were drilled. As shown on drawing 5379-1082 sheet 5, Safety Injection System Flow Diagram, Revision 34 the upstream discs of Sl-860A and B and the downstream discs of Sl-861A and B were drilled for pressure relie *

UFSAR Table 6.3.2-5, "Safety Injection Pump Design Parameters," indicated that the maximum SI pump flowrate was 550 gpm. WCAP-12070; calculation RNP M/MECH-1556, Revision 2; and test results for SP-986, Revision 0, dated January 26, 1991, all reflected maximum SI pump flows greater than 550 gpm. Additionally, manufacturer's information stated that each SI pump could operate up to a maximum flow of 650 gp *

UFSAR Section 6.3.2.2.8 (page 6.3.2-8) implied that a minimum of 300,000 gal was

"available for delivery" from the RWST. The correct value was 277,999 ga *

UFSAR Section 6.3.2.2.17 (page 6.3.2-15) stated that the SI system high pressure branch lines were designed for a pressure of 1500 psig. The correct value was 1750 psi The above discrepancies had not been corrected and the UFSAR updated to ensure that the information included in the UFSAR contained the latest material as required by 10 CFR 50.71(e).

The licensee stated that the above discrepancies would be corrected. The team identified this item as part of Unresolved Item 50-261/97-201-0 The team identified the following discrepancies in the DBDs and system descriptions:

Section 4.2.1.1 of DBD/R87038/SD13, "Component Cooling Water System,"

Revision 1, stated the design pressure of the CCW surge tank as 150 psig. The correct value was 100 psi *

Section 3.7 (page 20) of SI System Description SD-002, Revision 0, stated that the stroke time for valves other than those that must function on the SI signal was 20 seconds. The correct value was 120 second *

CCW System Description SD-013, Revision 0, listed the design heat transfer rate for the CCW heat exchangers as 28.9x10[6]. The correct value was 29.35x10[6].

• Thelicensee stated that the above discrepancies in the DBDs and system descriptions would be correcte E1.4 Other Related Electrical Systems E1.4.1 Inspection Scope The team evaluated the EDG loading calculation, cable sizing calculations, AC and DC system calculations, electrical modifications, and testing. The team focused on assessing the consistency between the electrical systems and their design and licensing base E1.4.2 Observations and Findings E1.4.2.1 AC System The team evaluated the /,C system voltage calculation, the EDG calculation, cable ampacity calculations, electrical penetration protection, relaying, and cathodic protectio The team reviewed calculation RNP-E-8.002, "AC Auxiliary Electrical Distribution System Voltage/Load Flow/Fault Current Study," Revision 3. The methodology, assumptions, and inputs were consistent with the design criteria, design bases information, and as-built data. Adequate AC voltage was determined to be available during normal, maintenance/refueling, and accident conditions for those electrical loads reviewe The team reviewed calculation RNP-E-8.016, "Emergency Diesel Generator Static and Dynamic Analysis," Revision 5, to verify that the analysis was consistent with the design basis information of the UFSAR and DBD/R87038/SDO5, "Emergency Diesel Generator System," Revision 1. The team verified that the analysis accounted for those accident loads associated with the LOCA/LOOP scenario in the design basis and in pending UFSAR changes identified by the license The team reviewed calculation RNP-E-5.004, "Ampacity Evaluation of Safety-Related Power Cables on 480V and 208V AC Motor Control Centers [MCCs] and Buses," Revision 4. No ampacity derating was included for cables routed in fire stops and seals or for cables with fire wrapping. The licensee did not derate cables, considering the impact to be negligible, but could provide no basis for this position. The licensee issued CR 97-01155 to address cable derating in fire stops and seals. The team identified this item as part of Inspector Followup Item 50 261/97-201-2 The licensee stated that the A and C CCW pump cables were wrapped in accordance with Modification M-81 1, "Dedicate/Alternative Shutdown System Fire Suppression for the CCW Pump Room," and that derating was considered for the modification, but calculation RNP-E-5.004 was not updated to reflect the fire wrapped cable. The licensee issued CR 97-01085 to revise the calculation. The team identified this item as part of Inspector Followup Item 50-261/97-201 2 The team questioned the replacement schedule for Agastat E7000 series relays. The team was aware that the manufacturer had recommended a 10-year replacement schedule for these relays

in correspondence to other utilities. The licensee stated that there was no replacement schedule because testing performed on these relays had concluded that the relays have a qualified life of greater than 50 years. The licensee referred to Acton Laboratories Test Report 15761, dated

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November 30, 1983, but could not produce this report for evaluation by the team during the inspection. The team identified this item as Inspector Followup Item 50-261/97-201-2 The team reviewed the licensee's evaluation of IN 92-04, "Potter & Brumfield Model MDR Rotary Relay Failures." Only two such relays were identified at the plant and the relays had been replace The team also reviewed the licensee's evaluation of IN 91-45, Supplement 1, "Possible Malfunction of Westinghouse ARD, BFD, and NBFD Relays, and A200 DC and DPC250 Magnetic Contactors," July 1994. The IN alerted licensees of cracking of the relay housing and potential relay malfunction. The licensee evaluated replacement of the BFD relays with NBFD relays in 1994. CR 96-00977 was issued on April 16, 1996, to document cracked relays in the reactor protection racks, and 20 relays were replaced. The licensee evaluation of the cracked reiays concluded that the cause for cracking was thermal degradation of the coil housing materia Subsequently, the licensee inspected all other BFD and NBFD relays for similar cracking and 13 additional relays were replaced. During RFO-17 an additional 13 relays were found cracked. On April 4, 1997, during a system engineering inspection, three additional relays were found to be cracked, CR 97-00908 was initiated, and the three relays were replaced. An enhanced inspection, using fiber optics, was conducted on May 15, 1997, of the B train of the RPS. Seven additional relays were identified as requiring replacement. An enhanced visual inspection was also scheduled for train A of the RPS. The licensee planned to continue visual inspections of these relays as part of the system engineers' walkdown of the RPS, and issued ESR 97-00283 to evaluate these relays. This item is being tracked by Region 11 because it was originally identified in one of the monthly resident inspector reports for HB E1.4.2.2 DC Battery Systems The safety-related DC system consists of two redundant battery trains with each train consisting of a 125V DC battery, battery chargers, and an inverter to supply power to the loads. For each station battery there are two battery chargers; one battery charger supplies the normal DC loads and maintains the battery fully charged, while the other provides 100 percent backup capabilit Only one battery charger per station battery is on-line at a time. The team reviewed the following analyses associated with the 125V DC system

. RNP-E-6.002, "B Battery 500F Ambient Temperature," Revision 0;

RNP-E-6.004, "DC Short Circuit Study," Revision 2;

RNP-E-6.005, "Overcurrent Protection and Coordination of the 125V DC Distribution System (Trains A & B)," Revision 0;

RNP-E-6.018, "DC Control Circuit Loop Analysis," Revision 0;

RNP-E-6.019, "IVSW Solenoid Valves EV-1 922A and EV-1 922B," Revision 1;

RNP-E-6.020, "Load Profile and Battery Sizing Calculation for Battery B,"

Revision 2;

• RNP-E-6.021, "Load Profile and Battery Sizing Calculation for Battery A,"

Revision 1;

RNP-E-6.022, "DC Voltage Profile," Revision 2

RNP-E-6.023, "Minimum Inverter Voltage Verification," Revision 2; and

RNP-E-6.028, "DS Battery Sizing Calculation," Revision Station batteries A and B were replaced in 1995 with batteries of the same physical size and with similar electrical characteristics from the original manufacturer. The replacement batteries were purchased to the requirements of the original plant specification Calculations RNP-E-6.020 and RNP-E-6.021 were prepared based upon IEEE Standard 485-1983,

"IEEE Recommended Practice for Sizing Large Lead Storage Batteries for Generating Stations and Substations." IEEE 485, Section 6.2.2, recommended that a design margin of 10-15 percent be included in the sizing calculation for future load additions. Additionally, Section 6.2.3 of IEEE 485 recommended that "the battery be replaced when its actual capacity drops to 80% of its rated capacity; therefore the battery's rated capacity should be at least 125% of the load expected at the eind of its service ofe." That implies an aging margin of 25 percent. Station battery A's rated capacity included a 25 percent aging margin and a 25 percent design margin. Station battery B's rated capacity included aging and design margins of only 10 percent each. The licensee has taken this smaller aging margin into account within surveillance test procedure MST-920, "Station Battery Performance Capacity Test," Revision 6, by committing to replace battery B when it drops to 90 percent of its rated capacity. The team noted that the load growth of station battery B was limited due to its existing spare capacit A review of calculation RNP-E-6.021 for sizing station battery A showed that 1-minute discharge rates (taken from manufacturer test results) differed from the published manufacturer data sheet Attachment J to RNP-E-6.021 stated that "Testing at the 1 minute rates showed that the published rates were inaccurate. It was consistently observed that at the beginning of each test: cell voltages decreased below failure levels within 10 seconds; then recovered to near, or above, failure levels; followed by decreasing voltage values consistent with theoretical discharge curves for lead acid batteries. This effect, known as the coup de fouet, is caused by the transition time required for a fully charged cell to completely initiate the chemical reaction at the plate/electrolyte boundary during battery discharge." Review of calculation RNP-E-6.020 for station battery B revealed that the battery sizing used the published 1 minute discharge rates and that the coup de fouet effect was not included in the calculation. The team questioned this inconsistency since both batteries, although different in size, came from the same manufacturer. Testing by the manufacturer was not performed on the type of battery installed as station battery B. The licensee contacted the manufacturer during the inspection and verified that the published ratings were accurate for station battery B. The licensee issued CR 97-00996 to confirm the battery ratings with the manufacturer. The team identified this item as Inspector Followup Item 50 261/97-201-2 The team noted that sizing calculations did not exist for the field flash batteries for the dedicated shutdown (DS) diesel and the EDGs. The primary means of flashing the EDGs is by using DC power from station batteries A and UFSAR Section 8.3.1.1.5.1 states that the EDG field flash battery could be used if the station batteries were unavailable. The licensee stated that calculations would be initiated for both of

. these field flash batteries. Section E1.5 of this report contains additional calculation discrepancies related to DC battery system E1.4.2.3 System Testing The team noted that battery capacity and service testing was not being performed for nickel cadmium batteries used for the EDG field flash, the DS diesel generator field flash, or the DS DC control storage battery. IEEE Standard 1106-1995, "Recommended Practice for Installation, Maintenance, Testing, and Replacement of Vented Nickel-Cadmium Batteries for Stationary Applications," recommended that capacity and duty cycle testing be performed on this type of battery. The licensee performed a self-assessment, Robinson Engineering Support Section (RESS) Report No. 96-04 (1/3/96-1/12/96), on the Station Blackout (SBO) Program, and noted that battery capacity and duty cycle testing was not being performed on either DS battery, and recommended that testing be developed. A self-assessment followup, RESS Report No. 96-04 (f)

dated March 27, 1997, noted that battery capacity and duty cycle testing had not yet been developed. Section 16 CFR 50.2 (3) and (4) state that an alternate AC source is required to be available in a timely manner after the onset of station blackout and that it must have sufficient capacity and reliability for operation of all systems required for coping with station blackout and for the time required to bring or maintain the plant in safe shutdown. Section 10 CFR 50.63(c)(2)

states that the capacity of an alternate AC source is to be demonstrated by analysis and its response time by test. Emphasis is placed on the availability and capacity of the alternate AC source because of the safety significance of plant shutdown. The alternate AC source at HBR is the dedicated shutdown diesel. If the field of this diesel is not flashed at the start of an SBO then it will be unavailable to perform its intended function. For its field to be flashed, the dedicated battery servicing this diesel must be demonstrated to be fully operational. The licensee can not verify the availability or operability of either the diesel or that battery because it has not incorporated testing of that battery into a testing program. The team identified this as Unresolved Item 50-261/97-201-2 Station batteries A and B, which are tested, are the primary means for field flashing the EDGs; however, UFSAR Section 8.3.1.1.5.1 states that EDG field flash batteries could be used if power from the station batteries was unavailable. The licensee stated that testing would be developed and implemented to ensure the adequacy of these nickel-cadmium batterie The team reviewed the latest test results for battery capacity and duty cycle testing for station batteries A and B. These tests were performed in accordance with procedures MST-920, "Station Battery Performance Capacity Test," Revision 6, and MST-921, "Station Battery Service Test,"

Revision 7. The team noted the following discrepancies:

Step 7.5.10 of procedure MST-921 accepted individual cell voltages as low as 0.5V D This voltage limit was below the value of 1.OV DC of IEEE Standard 450-1980,

"Recommended Practice for Maintenance, Testing, and Replacement of Large Storage Batteries for Generating Stations and Substations." Polarity reversal could occur at individual cell voltages of 1.OV DC or less. The licensee initiated CR 97-01017 to evaluate this ite The team's review of the manufacturer's capacity testing of station battery B and capacity testing performed by MST-920 revealed that the tests were performed for different

durations. The manufacturer's test was based upon an 8-hour discharge and MST-920 testing performed during refueling outage 17 (RFO-17) was based upon a 2-hour discharge. Comparison of the licensee test results with the original manufacturer's test results as described in IEEE 450-1980 was not possible because the tests had not been done using the same discharge ampacity rate and duration. Section 5.5.1.12 of DBD/R87038/SD16, "Electric Power Distribution System," Revision 1, stated that station battery surveillance testing would be performed in accordance with IEEE 450-1980. The licensee demonstrated to the satisfaction of the team that the two capacity tests showed that the performance of station battery B was typical of new lead acid batteries and that sufficient capacity existed to comply with MST-920 and IEEE 485-1983. The system engineer stated that the test documentation would be updated so that only 2-hour testing will be done in the futur The team reviewed the procedures used for intercell resistance checks. Both PM-410,

"Installation of Battery Bank & Cell Connections," Revision 6, and PM-41 1, "Disassembly, Cleaning, Assembly, and Testing of A and B Station Battery Cell Connections," Revision 6, gave acceptance criteria of 50 microohms for station batteries A and B. The team questioned these criteria and the licensee stated that the vendor calculation required that the B station battery not exceed 50 microohms and that the A station battery not exceed 34 microohms. The licensee performed a review of resistance data since 1994 and verified that no resistance exceeded the values in the vendor calculation. The licensee issued CR 97-01027 to revise the procedure Calculation RNP-E-6.018 evaluated loads from auxiliary panel DC on the basis of a projected station battery A voltage of 109.8V, whereas the acceptance criterion used in the station battery load profile testing procedure MST-921 was a minimum battery voltage of 107V, which was less than the voltage required by RNP-E-6.018. The licensee reviewed the recent MST-921 test results and verified that battery testing demonstrated that voltages were in excess of those required by the calculation. The licensee issued CR 97-01152 to coordinate the voltage requirements of RNP-E-6.018 and MST-92 Corporate Quality Assurance Manual, Section 6.3.2 states that activities affecting quality should be in accordance with approved procedures and/or drawings which are appropriate to the circumstances. Criterion XI of 10 CFR Part 50, Appendix 8 states that structure, system, or component (SSC) service requirements demonstrated by test must be performed per written test procedures which incorporate design requirements and acceptance limits contained in design documents. It appears that the requirements of Criterion XI, "Test Control" of 10 CFR Part 50, Appendix B, were not implemented in that the design basis was not correctly translated into procedures MST-920, MST-921, PM-410, and PM-41 1. The team identified this item as part of Unresolved Item 50-261/97-201 2 Documentation of the current applied during performance of MST-921 for station batteries A and B was inadequate in that duty cycle changes at 1 minute and at 59 minutes could not be substantiated with the test data provided. The licensee stated that the test device used was properly programmed for the load profile changes at times of 0, 1, and 59 minutes; however, the recording device only recorded at intervals of 5 minutes after the test initiation. The licensee resolved this discrepancy through CR 97-0104 * MST-920 testing is required to terminate when the battery terminal voltage dropped to 105V DC. The team reviewed data from a station battery B test during RFO-17: the data showed that the battery test stopped when the battery voltage reached 104.8V DC, contrary to IEEE 485-1983 and MST-920. The licensee could not determine how long the voltage took to drop to 104.8V from 105V DC. The licensee used the last reading before 104.8V DC and concluded that the battery had 108 percent capacity. The licensee issued CR 97-1046 to evaluate the test equipment used to perform this testin The team identified these two test discrepancies as part of Inspector Followup Item 50-261/97-201-2 MST-921, Section 3.4-3.6, requires the test engineer to refer to various design analyses for the required battery discharge currents and battery voltage acceptance criteria. This approach allows the testing procedure to change if the licensee changed the referenced calculation without performing a 10 CFR 50.59 evaluation as required by Section 5.1 of procedure PLP-032, " 10 CFR 50.59 Reviews of Changes, Tests, and Experiments,"

Revision 9. The licensee stated that 10 CFR 50.59 reviews are performed on revsions to procedures and the electrical requirements would be put into the testing procedur Additionally, Section 3.6 of MST-921 referred the test engineer to calculation RNP-E-6.023 for the battery "minimum terminal voltage." However, calculation RNP-E-6.023, "Minimum Inverter Voltage Verification," Revision 2, provided only a "minimum allowable voltage" and a "minimum battery voltage." The team reviewed the latest test results and verified that the correct voltage was used as an acceptance criterion. The licensee issued CR 97 01138 to include the actual required currents and numerical acceptance criteria in procedure MST-92 During the review of testing performed during RFO-17 in accordance with MST-920, the team also noted that the test dates and times that were manually documented did not agree with the start/stop clock data from the test equipment. Even though the start/stop times differed, the overall time duration was the same when derived from either the manual or the test equipment data. The licensee issued a request for procedure change, MI-506, on May 2, 1997, to revise the procedure appropriatel The team identified these two test procedure discrepancies as part of Inspector Followup Item 50-261/97-201-2 E1.4.2.4 Modifications The team reviewed Modification M-1065, "DGV Relay Setpoint Revision," Revision 0. This modification raised the existing degraded grid setpoints from 415V to 430V. The modification package was adequately prepared and provided the necessary technical basis for the change, installation and calibration instructions,;and acceptance testing requirements. The 10 CFR 50.59 evaluation for this modification was reviewed and no concerns were identified. The team verified that appropriate UFSAR and TS changes were incorporate El.4-2.5 System Walkdown The team conducted walkdowns to review electrical components in the control room, EDG rooms,

480V emergency switchgear areas, and the station battery rooms. The team observed the following:

There were inconsistencies between the vendor-specified material and torquing requirements and the installed condition of the station battery A and B intercell connections. Specifically, PM-41 1, "Disassembly, Cleaning, Assembly, and Testing of A and B Station Battery Cell Connections," Revision 6; vendor Technical Manual 728-147-75,

"Station Battery Installation and Operating Instructions," Revision 6; and vendor drawings 063969D, "Layout for 60 Cell MCX-340 Battery," Revision 1, and 063970D, "Layout for 60 Cells NCX-1050 Battery," Revision 1, were not in agreement. For example, the installed intercell connector for station battery A was 1/8-inch thick, whereas the vendor drawing specified it as 1/4 inch; and PM-411 required the intercell connection to be torqued to 140 in-lbs nominal, whereas the vendor technical manual required 100 in-lbs of torque. The licensee contacted the vendor and determined that the installed condition was acceptabl The licensee stated that the above documents would be reviewed for station batteries A and B and brought into conformance with each othe The team observed that a battery hydrometer well was attached to each of the station battery A and B racks. Station battery rack drawings 063969D, Revision 1, and 063970D, Revision 1, did not reflect this installation. The licensee issued Action Request 97-02045 to remove the unapproved installatio Additionally, a cable was observed hanging outside the cable tray above Safeguards Rack 64, contrary to DBD/R87038/SD62, "Cable and Raceway System," Revision 0. The licensee took immediate action to have the cable placed back into its racewa Screw-type mechanical connection lugs were utilized to connect the field installed cables to the positive and negative connections of station batteries A and B. The existing maintenance procedures did ndt require these connections to be torqued. Proper torque was required for these connections to ensure their integrity and to achieve the design electrical characteristics. The licensee contacted the vendor who recommended that these connections be torqued to 375 in-lbs. In the past, the licensee used standard maintenance practices for tightening these connections. These connections were checked in accordance with preventative maintenance procedures to ensure that minimal resistance existed so as to limit voltage drop across them. Additionally, thermography performed on each battery bank at 6-month intervals did not detect any high resistance connections on the battery terminals. The licensee stated that the torque recommendation of the vendor would be included in the appropriate preventive maintenance procedur UFSAR Section 17.3.2.1 requires work to be accomplished and verified using instructions, procedures, or appropriate means. This appears not to have been done for the sizing of the battery intercell connections, the inclusion of the battery hydrometers, or the protruding cabl UFSAR Section 17.3.2.2.1 stipulates that design documents reflect applicable regulatory, performance, quality, and verification requirements along with design bases. This was not done for the torquing of the intercell connectors. It appears that the requirements of Criterion IlI,

"Design Control" of Part 50, Appendix B, were not implemented in that the design basis was not

correctly translated into the installed components for these inconsistencies because of incorrect drawings, procedures, and work instructions. The team identified these three discrepancies as part of Unresolved Item 50-261/97-201-2 *

Cable pulling criteria were discussed with the licensee. Although procedures existed for cable installations, the criteria did not consider cable sidewall pressure during installatio The licensee stated that previous cable installations were installed by hand pulling. The method described by the licensee would result in an acceptable cable installation. The licensee issued CR 97-1090 to develop a program to monitor cable pulling tension and sidewall pressure during safety-related cable pull *

The team noted that conduits were installed near the EDG exhaust piping and questioned whether the cables within these conduits could be adversely affected by the hot local temperature during EDG operation and, if so, whether they were derated accordingly. The licensee stated that these circuits were used intermittently and their exposure to an elevated temperature would be relatively short; therefore, cable ampacity ierating was not necessary. The licensee issued CR 97-01196 to consider the need to provide additional design guidance for cable ampacity derating due to localized heat sources. The team identified this item as part of Inspector Followup Item 50-261/97-201-2 The team also noted that all areas were clean of debris, appropriate fire doors were closed, room temperatures were consistent with design analyses, station batteries and racks were free of corrosion, and the various rooms visited were sufficiently illuminated. The C station battery room floor had been painted so as to prevent any acid spillage from leaking through the floor and into the emergency switchgear area E1.4.2.6 UFSAR, DBD, and Drawing Review The team identified the following discrepancies in Section 8 of the UFSAR:

Section 8.3.2 stated the station battery A cell type as NCX, instead of the type NCN installe *

Section 8.3.1.1.2 referenced MCCs 5A1, 5A2, and 6A. These MCCs did not exis Additionally, the number of phases for the transformer supply to MCCs 9 and 10 was stated as 30 instead of *

Table 8.3.1 did not reflect the diesel generator loading of calculation RNP-E-8.016, Revision *

Figures 8.3.1-3 and 8.3.1-4 identified the CCW pump motor as 400 horsepower. The correct horsepower was 35 Theabove discrepancies had not been corrected and the UFSAR updated to ensure that the information included in the UFSAR contained the latest material as required by 10 CFR 50.71(e).

The licensee issued CR 97-00714 to update the UFSAR. The team identified this item as part

of Unresolved Item 50-261/97-201-0 The team identified the following discrepancies in the design basis documents, drawings, and system descriptions:

Section 4.2.2 of DBD/R87038/SD16, "Electrical Power Distribution System," Revision 1, refers to MCCs 5A1, 5A2, and 6A, which do not exis *

Drawing G-1 90626, sheet 3,"125V DC & 120V AC One Line Diagram." Revision 6, shows the output cable from inverter A as #4 AWG when the actual cable size is #2 AWG. The licensee issued CR 97-00992 to correct the drawin *

Although UFSAR Table 8.3.1-3 states that the EDGs are rated at 3125 kilovolt amperes (KVA), drawing G-190626, sheet 2, "480V & 120/208 Volt One Line Diagrai," Revision 5, identified the EDG ratings as 3437 KVA. The licensee issued DCR 97-159 to correct the drawin *

Page 4.6-4 of the TS basis references UFSAR Section 8.2 and UFSAR Table 8.2-4 for full load EDG acceptance criteria and load sequencing times, instead of UFSAR Section and Table 8.3.1-5. The licensee stated that corrections would be made when the Improved Technical Specifications (ITS) are issue *

System Description SD-016, "480/120 VAC Electrical System," Revision 1, identified the operating range of the inverters as 105-140V DC instead of the actual operating range of 100-140V DC. System Description SD-038, "DC Electrical System," Revision 0, stated that the battery room temperature is maintained between 72-820F, instead of the 690F to 850F range between the Lo and Hi room temperature alarms. SD-038 also states that the A battery is type NCX, instead of type NCN, and the information for battery chargers A-1 and B-1 do not reflect their voltage operating tolerance of +/- 10 percent. The licensee stated that these inconsistencies would be correcte The DBD. drawing, and SD discrepancies listed above have been or were to be evaluated by the license E1.4.3 Conclusions The team concluded that adequate AC supply is available for both normal and accident conditions. The electrical loading of the individual equipment was considered in the EDG calculation. The station battery DC system is adequate to supply the required power.during normal and accident conditions. The AC and DC design is consistent with the design bases. The team. noted apparent problems with derating of cables for fire stops and seals; qualifying Agastat E7000 relays; testing of Class 1 E and dedicated shutdown diesel batteries; and terminating battery lead E1.5, Calculations E1.5.1 Inspection Scope The team evaluated numerous engineering calculations related to the AFW and SI and interfacing systems, as discussed in other sections of this repor E1.5.2 Observations and Findings E1.5.2.1 Control of Calculations In several instances, the team observed that several calculations that performed the same or similar analyses with different input data and conclusions were currently active with none of them being identified as voided or superseded. The team was concerned that existing design basis calculations could be superseded and used in subsequent analyses and design decision Additionally, one calculation was performed without the required design verification. The following examples were identified:

(a)

Two calculations determined the CST level at which to switch the AFW pump suction supply to the SW system: RNP-1/INST-1015, "Condensate Storage Tank Level Alarm Setpoints," Revision 0, determined a 10 percent level, whereas calculation 84065-M-06-F, New Basis for CST Level Indication for CST Repair and Restoration - HBR Unit 2,"

Revision 3, determined this level to be 15 percent. Although these calculations provided a different value for the same level, neither of these calculations were marked void or supersede (b)

Calculation 789M-M-02, "Service Water System Model Evaluation of Traveling Screen Line Rupture for Double SWP, Single SWBP Operation with TB Isolated," Revision 0, was superseded by calculation RNP-M/MECH-1362, "SW Screen Wash Piping Flow Analysis,"

Revision 0, as discussed in Section E1.3.5.2 of this report. Calculation 789M-M-02 was not identified as void or superseded. The licensee stated that this item would be tracked by CR 97-0036 (c)

Calculation 789M-M-05, "H.B. Robinson Service Water Model with One SW Pump (D)

Running, One Booster Pump and with T. BIdq. Isolated," Revision 0 was superseded by calculation RNP-M/MECH-1 128, "Reduced SW Flow to EDG," Revision 2, as discussed in Section E1.3.5.2 of this report. Calculation 789M-M-05 was not identified as void or superseded. The licensee stated that this item would be tracked by CR 97-0036 (d)

ESR 96-00474 contained a calculation to evaluate whether a rupture of a non-seismic pipe could result in insufficient water available for the SI pumps to perform their safety functio This calculation was not design verified as required by the Corporate Quality Assurance Program, nor was it performed in accordance with procedure MOD-002, "Design Calculations," Revision 10. The licensee issued CR 97-01226 to evaluate this ite (e)

Calculation RNP-E-6.002, "B Battery 500 Ambient Verification," Revision 0, was voided and replaced by calculation RNP-E-6.020, "Load Profile and Battery Sizing Calculation for Battery B," Revision 2. Calculation RNP-E-6.002 was not marked void as required by procedure MOD-002; and calculation RNP-E-6.020 stated on the cover page that it superseded calculation 7988-El, Revision 4, but did not refer to calculation RNP-E-6.002 on the cove It appears that the control of calculations did not meet the requirements of UFSAR Section 17.3.2.2.4 which requires design documents and procedures to be controlled to reflect design modifications and "as-built" conditions and also USFAR Section 17.3.2.2.5 which requires design interfaces be defined and controlled, including the review, approval, release, and distribution of design documents and revisions. This represents a weakness with the design control measures stated in Criterion Ill "Design Control" of 10 CFR Part 50, Appendix B. The team identified this as Unresolved Item 50-261/97-201-2 E1.5.2.2 Calculation Errors The team identified numerous errors in calculations. These errors were incorrect inputs, incomplete analyses, undocumented assumptions, incorrect analyses, and inattention to detai The following are examples of these errors:

(a)

Siemens calculation EMF-94-203(P), "H.B. Robinson Unit 2 Small Break LOCA Analysis,"

dated October 1994, erroneously considered that the flow from the MDAFW pump was uniformly distributed over the three SGs. Actually, a single active failure would result in flow to only two SGs. A preliminary evaluation by Siemens indicated that the limiting break with flow to only two SGs had no significant impact on the PCT. The licensee initiated CR 97-01163 to document the preliminary evaluation and to check the effects of flow to only two SGs on other event (b)

RNP-M/MECH-1362, "SW Screen Wash Piping Flow Analysis," Revision 1, the analysis provided to the team to demonstrate adequate performance of the SW system with ruptures of non-seismic piping, did not include rupture of the non-seismic piping that supply the instrument and station air compressors. The licensee stated that this additional rupture was bounded by an earlier evaluation and initiated CR 97-00993 to determine the long term action neede (c)

Calculation RNP-MN/MECH-1060, "CCW Hx Performance with Reduced Service Water Flow," Revision 1, used input data identified as from calculation 789M-M-02, but these data were different than the approved Revision 0 of calculation 789M-M-02. The licensee stated that this item would be tracked by CR 97-0036 (d)

Calculation RNP-E-6.020, "Load Profile and Battery Sizing Calculation for Battery B,"

Revision 2, incorrectly referenced a time period of "2 minutes to 59 minutes," instead of "1 minute to 59 minutes," and referenced an incorrect battery cell type. (MCT instead of MCX)

in Attachment U. The licensee issued CR 97-00996 to.correct these discrepancie.

This calculation also did not consider some of the connected non-safety-related loads in the load profile. The licensee issued CR 97-01082 to analyze one load and stated that another would be addressed by CR 97-0099 Additionally, this calculation neglected the coup de fouet effect without justification. The licensee stated that this discrepancy would be tracked by CR 97-0099 (e)

Calculation RNP-E-5.004, "Ampacity Evaluation of Safety Related Power Cables on 480V and 208V MCC's and Buses," Revision 4, did not address the cable derating due to fire wrap on cables. The licensee issued CR 97-01085 to revise the calculatio (f)

Calculation RNP-E-6.023, "Minimum Inverter Voltage Verifications," Revision 2, did not consider the increased inverter current at reduced battery voltage. The licensee issued CR 97-01070 to revise the calculatio Additionally, ths calculation diu not justify neglecting the effect of circuit breaker and bus resistance (g)

Calculation RNP-E-6.004, "DC Short Circuit Study," Revision 2, did not consider a small DC motor that was connected to the system. The licensee issued Design Change Backup Form (DCBF) RNP-E-6.004-0001 to resolve this ite The calculation also used a battery terminal voltage during a short circuit of 120V DC. The manufacturer's test data for station batteries A and B duty cycle testing performed in accordance with procedure MST-921 and for the battery capacity testing performed in accordance with MST-920, indicated that the open circuit voltage was larger than 120 volt Additionally, this calculation, along with calculation RNP-E-5.018, "Ampacity Evaluation of Safety Related 125VDC and 120VAC Power Cables," Revision 4, analyzed for cables rated at 750C, whereas 60 0C rated cables were installed. The licensee issued CR 97-01126 to correct the calculation (h)

Calculation RNP-E-6.018, "DC Control Circuit Loop Analysis," Revision 0, used incorrect solenoid valve power values as input. The licensee stated that this discrepancy would be tracked by CR 97-0099 (i)

Calculation RNP-E-8.016, "Emergency Diesel Generator Static and Dynamic Analysis,"

Revision 5, used an incorrect reference and only modeled the B SI pump motor. The licensee issued CR 97-01074 to correct the calculatio (j)

Calculation RNP-M/MECH-1460, 'NPSH VS. CST level for SDAFW pump," Revision 0, assumed the CST water temperature to be 1000 F, instead of 1150 F as in the Plant Parameter Document for Cycle 1 (k)

Calculation RNP-M/MECH-1394, "AFW Pump Recirculation Flowrates for RNP-2," Revision 2, used an incorrect specific gravity for the CST wate Calculations with errors did not meet the appropriate quality standards as required by UFSAR Section 3.1.2.1, GDC 1, or the requirements of procedure MOD-002. Design procedure EGR NGGC-003, "Design Review Requirements", stated that verification shall confirm that design inputs are correct, that the final design meets design inputs, and that the design is technically adequate and accurate. The licensee, as illustrated by example, had not adhered to Criterion III

"Design Control" of 10 CFR Part 50, Appendix B, in regard to verifying the adequacy of calculations. The team identified this item as part of Unresolved Item 50-261/97-201-0 E1.5.3 Conclusions Several calculations that performed the same or similar analyses with different input data and conclusions were currently active with none of them being identified as voided or supersede The team identified numerous errors in calculations. These errors were incorrect inputs, incomplete analyses, undocumented assumptions, and incorrect analyses. However, none of the deficiencies affected the final conclusion of any calculatio APPENDIX A Open Items This report categorizes the inspection findings as unresolved items and inspection follow-up items in accordance with the NRC Inspection Manual, Manual Chapter 0610. An unresolved item (URI)

is a matter about which more information is required to determine whether the issue in question is an acceptable item, a deviation, a nonconformance, or a violation. The NRC Region II office will issue any enforcement action resulting from their review of the identified unresolved items. An inspection follow-up 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 information is needed that was not available at the time of the inspectio Item Number Findinq Title Type 50-261/97-201-01 IF[

Operating Event Review of IN (Section E1.2.2.2(b))

50-261/97-201-02 URI Single Failure of CST Level Instrumentation (Section E 1. 2.4.2 (a))

50-261/97-201-03 URI Lack of Design Verification for Calculations (Sections El.2.4.2(c); El 3.2.2(b), El 3.4.2(c), El 3.4.2(e);1 E1. 3.5.2(a); El1 5.2.2)

50-261/97-201-04 URI Instrument Sensing Line Slope (Sections E

E1.2.4.2 (g); E1. 3.4.2 (g))

50-261/97-201-05 URI AFW UFSAR Discrepancies (Sections E1.2.6; E1.3.6; El 1.4.2.6)

50-261/97-201-06 URI Notification of Changes in PCT (Section El 1.3.2.2 (a))

50-261/97-201-07 URI Reporting of Significant PCT Changes (Section El.3.2.2(a))

50-261/97-201-08 IFI Evaluation of Transfer to Cold Leg Recirculation (Section El2.3.2.2(a))

50-261/97-201-09 URI SI, RHR, and CS Pump NPSH (Section El1 3.2.2(c))

50-261/97-201-10 URI SL Valve Testing (Sections E3.2.2(e))

50-261/97-201-11 IFI Classification of Valves in SST (Sections El1 3.2.2(e); E1.3.2.2(f))

• AS

5.0-261/97-201-12 URI 10 CFR 50.59 Screening Deficiency (Section E 1. 3.2.2(f)

50-261/97-201-13 IFI RAB Flooding due to SW System Passive V

Failure (Section El.3.2.2(g))

50-261/97-201-14 URI SI Cable Separation Discrepancy (Section E 1. 3.3.2 (a))

50-261/97-201-15 IFI SI Pump Motor Load Evaluation (Section El 1.3.3.2 (a))

50-261/97-201-16 IFI Seismic Qualification of 480V AC Circuit Breakers (Section E1.3.3.2(c))

50-261/97-201-17 URI SI Accumulator Pressure Alarm Setpoint (Section E1.3.4.2(a))

50-261/97-201-18 URI RWST Level Instrument Uncertainty (Section El 1.3.4.2 (b))

50-261/97-201-19 IFI Containment Water Level Setpoint and Instruments used in EOPs and AOPs (Sections E1.3.4.2(c);

E 1. 3.4.2 (d))

50-261/97-201-20 IFI CCW System Overpressurization (Section E 1. 3.5.2 (a))

50-261/97-201-21 URI Translation of Design Bases into Drawings, Procedures, and Installed Components (Sections E1.3.5.2(a); E1.4.2.3; El.4.2.5)

50-261/97-201-22 IFI Ampacity Derating of Cables (Section El.4.2.1, El.4.2.5)

50-261/97-201-23 IFI Agastat Relay Lifetime (Section El.4.2.1)

50-261/97-201-24 IFI Station Battery B Rating (Section El.4.2.2)

50-261/97-201-25 URI Field Flash Battery Testing DSDG (Section E 1.4.2.3)

50-261/97-201-26 IFI Station Battery Test Control Deficiencies and Test Procedure Revisions (Section El.4.2.3)

50-261/97-201-27 URI Control of Calculations (Section E i.5.2.1)

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APPENDIX B LIST OF ATTENDEES AT PUBLIC EXIT MEETING ON JUNE 12, 1997 NRC CP&L Jaudon, Campbell, Keenan, Norkin, Young, Warden, Mozafarri, Chernoff, Boska, Kleeh, Clements, Habermeyer, Desai, Moyer, Orser, Hanna, Krich, Miller, Gaffney, Crook, Cafarella, Taylor, McCauley, Carley, A. *

B-1

APPENDIX C LIST OF ACRONYMS AC Alternating Current AFW Auxiliary Feedwater AMSAC (ATWS (Anticipated Transient Without Scram) Mitigation System Actuation Circuitry)

ANSI American National Standards Institute AOP Abnormal Operating Procedure ASME American Society of Mechanical Engineers CCW Component Cooling Water CFR Code of Federal Regulations CP&L Carolina Power & Light CR Condition Report CS Containment Spray CST Condensate Storage Tank DBA Design Basis Accident DBD Design Basis Document DBE Design Basis Earthquake DC Direct Current DCBF Design Change Backup Form DCF Document Change Form DCN Design Change Notice DCR Design Change Request DP Differential Pressure DSS Dedicated Shutdown System ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EE Engineering Evaluation EOP Emergency Operating Procedure EPP End Path Procedure ESF Engineered Safety Features ESR Engineering Service Request EST Engineering Surveillance Test FCV Flow Control Valve GDC General Design Criteria GID Generic Issues Document gpm Gallon Per Minute HELB High Energy Line Break HBR H. B. Robinson Steam Electric Plant, Unit No. 2 I&C Instrumentation &-Controls IEEE Institute of Electrical and Electronic Engineers IFI Inspector Follow-up Item ILRT Integrated Leak Rate Test IN Information Notice IP Inspection Procedure IVSW Isolation Valve Seal Water System kV Kilovolt C-1

KVA KiloVolt-Amp kW Kilowatt LBLOCA Large Break LOCA LCO Limited Condition for Operation LER Licensee Event Report LOCA Loss of Coolant Accident LOOP Loss of Off-site Power MCC Motor Control Center MDAFW Motor Driven Auxiliary Feedwater MOV Motor-Operated Valve MST Maintenance Surveillance Test NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation, Office of OE Operating Events OMM Operations Management Manual OP Operating Procedure OST Operations Surveillance Test PCT Peak Clad Temperature PM Preventive Maintenance PRA Probabilistic Risk Assessment psig Pounds per Square Inch Gauge PWR Pressurized Water Reactor RAB Reactor Auxiliary Building RCS Reactor Coolant System

RESS Robinson Engineering Support Section RET Release Engineering Task RFO Refueling Outage RG Regulatory Guide RHR Residual Heat Removal RNP Robinson Nuclear Plant RPS Reactor Protection System RTGB Reactor Turbine Generator Board RWST Refueling Water Storage Tank SBO Station Blackout SBLOCA Small Break LOCA SD System Description SDAFW Steam Driven Auxiliary Feedwater Pumps SG Steam Generators SI Safety Injection SQUG Seismic Qualification Utility Group SSC Structure, System, or Component SSFI Safety System Functional Inspection SW Service Water SWEC Stone and Webster Engineering Company SWOPI Service Water System Operational Performance Inspection TMM Technical Support Management Manual TS Technical Specification C-2

URI Unresolved item UFSAR Updated Final Safety Analysis Report USQ Unreviewed Safety Question V

Volt VS Ventilation System WCAP Westinghouse Commercial Atomic Power WOG Westinghouse Owners Group C-3