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**UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555-0001**

October 26, 1998

**Addressees**

All holders of operating licenses for nuclear power reactors, except those licensees who have permanently ceasedoperations and have certified that fuel has been permanently removed from the vessel.

**Purpose**

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to the fact thatincorrect level instrument setpoints or other control deficiencies can render emergency core cooling system (ECCS) pumpsinoperable during certain design-basis accidents. It is expected that recipients will review the information for applicabilityto their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in thisinformation notice are not NRC requirements; therefore, no specific action or written response is required.

**Description of Circumstances**

Oconee Nuclear Station

In late 1997, during a self-assessment audit of the high pressure injection (HPI) and low pressure injection (LPI) systems,the licensee noted that the design drawing for the borated water storage tank (BWST) did not have a zero reference point.Subsequently, the licensee's engineering staff determined that an elevation difference between the level transmitters and theinstrument taps for the BWSTs of all three Oconee units had resulted in up to an 18-inch nonconservative error betweenindicated and actual BWST level. The difference was caused by a failure to compensate for instrument tap height whencalibrating the BWST level instruments. At plant construction, the magnitude of the error was approximately 4 inches, butthe error increased to approximately 18 inches following modifications in 1989 that replaced the BWST level transmitters.In addition, on February 19, 1998, the licensee's engineering staff determined that the emergency operating procedures(EOPS) did not adequately account for uncertainty in the reactor building emergency sump (RBES) wide-range levelinstruments, which could have resulted in the instruments reading up to 18 inches lower than the actual level. The RBESlevel instrument uncertainties were caused by inadequate design analysis. This was discussed in Inspection Report 50 50-269, 270, 287/98-12 and Licensee Event Report (LER) 50-269/98-04, Revision 1.

St. Lucie Nuclear Plant

During the Unit 1 steam generator replacement outage in 1997, the licensee replaced the engineered safety featuresactuation system (ESFAS) bistables. A system engineer later determined that the ESFAS recirculation actuation setpoint(RAS) bistable setpoint for the Refueling Water Tank (RWT) level was incorrect. An investigation of the discrepancyfound that, during a setpoint calculation enhancement effort in 1993, a new calculation was created, which changed thespan of the RWT level measurement and indication instrumentation loop. This calculation produced a new setpoint for theRWT level by revising the measurement span to indicate the actual tank level bottom as "0 feet". Previously, themeasurement span indicated 0 at the 1-foot level where the RWT level instrument tap is located. The new setpointinformation was not incorporated into the procedure used to calibrate the ESFAS bistables, resulting in a RAS setpoint of 3feet from the tank bottom instead of 4 feet required by Technical Specifications. This was discussed in Inspection Report50-335, 389/97-16 and LER 50-335/97-11.

H.B. Robinson Nuclear Plant

Between April 7 and May 23, 1997, NRC conducted a design inspection at Robinson and raised several Refueling WaterStorage Tank (RWST) instrumentation related issues that impacted ECCS components. Plant Emergency OperatingProcedures (EOPs) directed all ECCS pumps, except one safety injection (SI) pump and one containment spray (CS) pump,to be stopped when the RWST level reached 27 percent. The remaining SI and CS pumps were directed to be stopped whenthe RWST level reached 9 percent. The plant evaluated and modified the number of SI pumps starting following a LOCA.The modification resulted in two (as opposed to three) SI pumps starting following a LOCA. With two SI pumps getting astart signal, and assuming a single active failure of one SI pump, the NPSH requirement for the running SI pump washigher, and this higher NPSH requirement was not considered in the modification. A calculation to determine the level atwhich vortexing became a concern had not been performed prior to the modification. The licensee regained the margin byreducing instrument uncertainties and by raising the water level in the RWST.

The NRC design team also found that the containment sump level setpoints utilized channel uncertainty for normalenvironmental conditions rather than the adverse conditions that could exist in the containment after an accident. This hadthe potential for adversely affecting residual heat removal (RHR) pumps in the recirculation mode of operation. Theseitems were discussed in Inspection Report 50-261/97-201 and LER 50-261/97-08.

**Discussion**

In the Oconee Nuclear Station's Final Safety Analysis Report (Sections 6.2 and 6.3), the licensee states that during certainloss-of-coolant accidents, reactor operators must be capable of manually providing a flowpath from either the BWST or theRBES to the HPI, LPI and reactor building spray (RBS) pumps. The errors described above created a conflict between theBWST and RBES levels specified in the EOPs and the BWST and RBES levels indicated in the control room. As a result,during certain design-basis accident scenarios, including small-break

LOCAs, the level errors could have delayed swapover initiation. This could have caused vortexing in the BWST or reducedNPSH to the ECCS pumps, or both.

The design basis of the St. Lucie facility requires that during certain LOCAs, ECCS subsystems must be capable ofautomatically transferring suction to the containment sump on receipt of a RAS. Because of the incorrect trip setpoint ofthe RWT level instrument bistables, automatic transfer of the ECCS pumps' suction source from the RWT to thecontainment sump, under certain conditions, would cause an open-channel flow condition. Without operator intervention toinitiate manual transfer to the containment sump before the open-channel condition, damage to the ECCS pumps couldoccur as a result of air entrainment.

At H.B. Robinson, the cause of the reduction in SI pump NPSH was a failure to adequately assess the impact of single SIpump operation on system flow and NPSH during a 1988 modification. Inadequate NPSH to ECCS pumps could have ledto the inoperability of critical safety-related systems and loss of core cooling under some design-basis LOCA conditions.The preceding examples demonstrate the importance of thorough assessment and analysis for any modification involvingsafety-related level instrumentation or ECCS pump operating conditions. Information Notice 98-22, "DeficienciesIdentified During NRC Design Inspections," dated June 17, 1998, also described ECCS swapover analysis errors at D.C.Cook, H.B. Robinson, Three Mile Island, Wolf Creek, and Ginna.

This information notice requires no specific action or written response. If you have any questions about the information inthis notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear ReactorRegulation (NRR) project manager.

Attachment:  List of Recently Issued NRC Information Notices

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