IR 05000255/1998012

Insp Rept 50-255/98-12 on 980706-24.No Violations Noted. Major Areas Inspected:Engineering

ML18068A429
Person / Time
Site:	Palisades
Issue date:	09/02/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML18068A427	List: IR 05000255/1998012 ML18068A426 ML18068A428
References
50-255-98-12, NUDOCS 9809090210
Download: ML18068A429 (34)

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q U. S. NUCLEAR REGULATORY COMMISSION Docket No:

License No:

Report No:

Licensee:

Facility:

Location:

Dates:

Inspectors:

Approved by:

9809090210 980902 PDR ADOCK 05000255 G

PDR REGION Ill 50-255 DPR-20 50-255/98012(DRS)

Consumers Energy Company Palisades Nuclear Generating Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530 July 6 through 10, 1998 July 20 through 24, 1998 R. Westberg, Reactor Engineer, Team Leader I. Jackiw, Reactor Engineer R. Mendez, Reactor Engineer D. Jones, Reactor Engineer D. Prevatte, NRC Contractor John Jacobson, Chief, Lead Engineers Branch Division of Reactor Safety

• EXECUTIVE SUMMARY Palisades Nuclear Generating Plant NRC Inspection Report 50-255/98012 The purpose of the inspection was to evaluate the effectiveness of the engineering and technical support (E& TS) organization in their performance of routine and reactive site activities including identification and resolution of technical issues and problems. The inspection focused on system engineering functions, modifications, technical problem resolution, and engineering support to other plant organizations. The criteria use to assess E& TS performance was understanding of plant design and involvement in preventing and solving plant problems. In addition, the effectiveness of the corrective action program in identifying, resolving, and preventing problems was evaluate Engineering

Overall, for the 51 CRs reviewed, the corrective actions taken were good and root cause determinations were effective. The team also noted that a low threshold existed for identifying problems and issuance of condition reports. However, two minor examples were noted where corrective actions could have been improved (Section E 1.1 ).

Although the need for testing of molded case circuit breakers had been licensee identified in 1993, from review of industry operating experience information, a testing program was not developed until 1997, after 44 of 72 molded case circuit breakers failed to trip during testing. The failure to assure that this condition advers.e to quality was promptly identified and corrected was considered a violation (Section E2.1 ).

Surveillanc~ Test Procedure M0-7A-1 for the diesel generator went beyond the specific*

warning contained in IN 97-16 to assure that any adverse condition found concerning *

liquid in the cylinders would be formally documented and evaluated (Section E2.1 ).

Overall, the 10 CFR 50.59 screenings and safety evaluations reviewed for the past two years were of good quality and a good program had been established for ensuring that trained and qualified personnel prepared and reviewed 50.59 screenings and safety evaluations (Section E3.1 ).

The team reviewed 24 modifications and nine temporary modifications and concluded that they were of good quality, properly installed and tested (Section E3.2).

Based on interviews with station personnel and review of corrective action documents, the team concluded the licensee's corrective action, audit, and self-assessment programs were effective. The team concluded that quality assurance activities were of appropriate depth and scope (Section E7.1 ).

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The corrective action program at Palisades had shown improvements in identification, resolution, and prevention of problems in the past two years. Personnel interviewed indicated a willingness to identify problems, considered the process to be owned equally by all plant staff, and did not consider CRs written against themselves to be negativ Overall, the licensee has been effective in the identification and resolution of problems (Section E7.2).

The program for screening, analyzing and dispositioning industry experience issues appeared to be effective; ho~ever, the team noted two examples where Engineering concluded that concerns were not applicable to Palisades because the conditions were not precisely the same as those at Palisades, rather than taking the broader view.of how and where there were similarities (Section E2.1 and E7.3).

• The team concluded that the Self-Assessment Program was effective and capable of providing valuable performance insights. The team also fourid that the audit program covered the required areas and was identifying problems and concerns. Audit findings were documented on condition reports, which were used for tracking and to obtain corrective actions. Areas of concern identified by audit findings were promptly and effectively corrected (Section E7.4 ).

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Report Details Ill. Engineering E1 Conduct of Engineering E1.1 Problem Identification and Root Cause Determination Inspection Scope (37550)

The team reviewed 51 condition reports (CRs) and verified whether the CRs were properly evaluated for root cause determination and corrective actions. Documents reviewed are listed at the end of this repor Observations and Findings Overall, for the sample reviewed, the corrective actions taken for CRs were good and root cause determinations were effective in determining root causes. The team noted that a low threshold existed for identifying problems and issuance of CRs. However, the team had concerns with the disposition and corrective actions for the following CR During the NRC architect/engineer (A/E) design inspection (Inspection Report No. 50-255/97201) CR No. C-PAL-97-1568 was issued which noted that overcurrent relays for breakers 152-105 and 152-106 had not been calibrated since 1992. These relays were required to be calibrated in 1995, but the scheduled test was missed. The corrective actions for the CR committed tq calibrating the relays during the 1998*

refueling outage (April 24 - June 7) concluding that this was acceptable because past calibration records indicated no history of problems. Subsequently, a violation.was issued in the A/E inspection.follow up (Inspection Report No. 50-255/98003) for failure to test these relays. The June 24, 1998, response to this violation committed to test these relays by December 31, 1998; during the refueling outage, a decision was made to delay their calibration until then. While this was consistent with the violation commitment, no consideration was given to the prior commitment made in CR 97-1568 to test the relays during the refueling outag CR No. C-PAL-97-1112 documented a torque wrench lost in August 1997. The CR was closed on September 3, 1997 with the conclusion stating "no equipment operability was affected." The team noted that without as-found data for the torque wrench, the evaluation could not conclude that no equipment operability was affected. The procedure that controlled measuring and test equipment (M&TE}, required prompt notification should an M& TE item be nonconforming and defined lost M& TE to be considered potentially nonconforming. However, no nonconformance evaluation was initiated. In March 1998, the torque wrench was found and testing determined that it was in tolerance. The team determined that the significance of the lost torque wrench was low because it was found to be in tolerance; however, the initial disposition of the CR was without basi *

• Conclusions Overall, for the 51 CRs reviewed, the corrective actions taken were good and root cause determinations were effective. The team also noted that a low threshold existed for identifying problems and issuance of CRs. However, two minor examples were noted where corrective actions could have been improve E2 Engineering Support of Facilities and Equipment E Licensee Response to Industry Initiatives Inspection Scope (37550)

The inspection scope included review of responses to notifications of issues of generic*

. interest, particularly NRC originated notifications, such as Information Notices (INs) and Generic Letters (Gls). Documents reviewed are listed at the end of this report. Review of the licensee's Industry Experience Program is documented in Section E7.2 of this repor Observations and Findings Molded Case Circuit Breaker (MCCB) Failures On April 7, 1993, the NRC issued IN 93-26, "Grease Solidification Causes Molded Case Circuit Breaker Failure To Close," which discussed General Electric MCCBs that failed to close because of grease solidification.. A Palisades industry experience traveler was issued; however, its disposition stated that the*

MCCBs were not greased, were not disassembled, and the grease in them was per vendor requirements. Therefore, no further action was taken at this tim However, on April 28, 1993, action item record A-PAL-93-017 was issued to evaluate and develop a program for periodic testing of safety-related an non-safety-related MCCBs. On August 12, 1993, the NRC issued IN 93-64,

"Periodic Testing and Preventative Maintenance of Molded Case Circuit Breakers," which discussed numerous failures with Westinghouse MCCBs due mainly to age related degradation and infrequent exercising of MCCBs, some of which may not have been exercised since initial plant startu On November 15, 1996, in response to corrective actions for licensee event report (LER) 96005, "Appendix R Enhancement Analysis - DC Panels Breaker/Fuse Coordination Issue," the l~censee tested four MCCBs and all four failed to trip on overcurrent. The licensee determined that there was a potential that failure of a downstream breaker could result in the loss of an entire DC *

distribution panel. Therefore, failure of the four MCCBs resulted in a concern that all DC MCCBs installed in the distribution panels, 72 in total, could fail to trip when subjected to a short circuit. These 72 MCCBs were tested and all magnetic-only MCCBs, 44 in total, would not trip under high fault currents. The remaining 28 thermal-magnetic MCCBs passed the as-found testing within

• specifications. The root causes for the failure to trip were determined to be hardening of breaker lubrication which restricted movement of breaker internal components and an oversight by the plant's preventative maintenance program by not identifying the degraded breaker conditions. In addition, discussions with the licensee indicated that, in general, the MCCBs had not been exercised since initial plant startup. LER 96013, "DC Breaker Failure During Testing For As-Found Trip Setting," and a supplement were issued documenting the MCCB failures. Unresolved item (URI) 50-255/96018-01 was issued by the NRC to track this issu The licensee replaced all 72 breakers and established a preventative maintenance program on December 1, 1997, to test approximately a third of the breakers every refueling outage so that all the breakers would be tested in a six year period. The team considered these corrective actions were acceptable and comprehensive. Failure to promptly evaluate a previously identified generic problem with MCCBs, a condition adverse to quality, and to develop a program for their periodic testing in a timely manner was a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" (50-255/98012-01 ). However, Inspection Report No. 50-255/94002 had already adequately addressed on the docket corrective actions taken and actions to prevent recurrence for a significant breakdown in the control of the Palisades corrective action program which occurred in the same time frame as this violation. Therefore, no response to this violation is require.

Analyses of Safety-Related ECCS Pump NPSH The team reviewed the response to GL 97-04, "Assurance of Sufficient Net Positive Suction Head [NPSH] for Emergency Core Cooling and Containment Heat Removal Pumps," and its NPSH calculation, EA-A-PAL-96-003, "ECCS Evaluation in Post-RAS Recirculation Modes Using Pipe-Flo," Revision 1. The team also reviewed sixteen generic communications relative to emergency core cooling systems (ECCSs) and NPSH. Eleven of these sixteen communications, including GL 97-04, discussed the potential for debris inside containment to be transported to the pumps' suction screens post loss-of-coolant-accident (LOCA),

thereby reducing available NPSH below acceptable limits. GL 97-04 required licensees to provide the NRC with detailed information regarding their NPSH analyses, including the methodology for calculating strainer head losses and the required versus the available NPSH (NPSHR versus NPSHA respectively). The team identified two concerns. First, per the licensee's response letter, containment spray pump P-54A and high pressure safety injection pump P-66A had inadequate available NPSHs (3.5 feet and 3.3 feet less than the required NPSH respectively). The response letter stated that this condition was acceptable; however, there was insufficient information in the response to substantiate this conclusion. Second, the supporting analysis contained several discrepancie The GL 97-04 response stated that the inadequate NPSH condition was acceptable since it would exist for only a short period during pump suction switchover at the initiation of the accident recirculation phase, until the operators placed the subcooling lineup in service (up to 30 minutes). It also stated that operation with this degree of NPSH deficit for this period of tin:ie had been approved by the pump manufacturers. However, no documentation could be provided that the containment spray pump manufacturer had ever given this approval. As a result of the team's inquiries, the licensee contacted the containment spray pump vendor who provided approval to operate the pump in this mode for this amount of time, and the licensee performed additional evaluations that indicated that, even with degraded flow, the design basis flows would be provided. However, these evaluations were based on EA-A-PAL-96-003 and the team identified the following concerns during review of this calculation:

No allowance was ma.de for air entrainment. Both NUREG-0897,

"Containment Emergency Sump Performance," Revision 1, and Regulatory Guide (RG) 1.82, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," Revision 2, provided formulae to correct upward NPSH required under thes conditions. A licensee letter to the NRG.dated July 9, 1982, concerning Unresolved Safety Issue A-43, "ContainmentEmergency Sump Reliability," documented that the level of air ingestion that could be expected would be approximately 2%, a level which would have produced a small increase in the required NPS *

Although RG 1.82, "Sumps for Emergency Core Cooling and Containment Spray Systems," Revision 0, indicated an assumed blockage of 50% should be used in sump screen design, the licensee's analysis assumed only 10%. This assumption was based on a qualitative analysis which stated that the blockage would be very low because LOCA debris ge.neration would be low based on the following: 1) most of the insulation was jacketed or encapsulated; 2) low debris transport flow velocities to the sump; arid 3) a torturous path to the sump. However, GL 85-22, "Potential for Loss of Post-LOCA ReCirculation Capability Due to Insulation Debris Blockage," stated that using even 50% blockage usually will result in a non-conservative analysis for screen blockage effects, and that such assumptions should be replaced with a more comprehensive requirement to assess debris effects on a plant specific basis. It also singled out plants with debris screens less than 100 square feet area as being particularly vulnerable; Palisades' screens are 35 square fee Further, NUREG-0897, "Containment Emergency Sump Performance,"

Revision 1, showed that significant LOCA debris generation from fibrous insulation could be expected even with encapsulation, and it could be transported at near neutral buoyancy conditions and very low transport velocities, and would be deposited uniformly across screens. Such deposition would tend to progressively decrease the effective screen

• opening size, thereby filtering out increasingly smaller particles and thus significantly increasing the pressure drop over what would be calculated using a simple percentage blockage assumption such as that used by Palisades. The containment at Palisades had three types of fibrous of insulation installed; low-density, molded, encapsulated mineral wool; jacketed, molded calcium silicate (non-jacketed on main steam); and encapsulated fiberglas In order to gain an understanding of the magnitude of the blockage that might be expected on the Palisades screens, the team performed an order-:of ~*magnitude calculation using information available since 1985. The debris generation value assumed was taken from NUREG-0897 at the very low end of the values tabulated for a typical PWR, and an assumption was made that only 10% of this debris would be transported to the sump. The calculation was performed using a pressure drop formula also from this NUREG for mineral wool insulation, a fibrous insulation type that produced the least pressure drop. Even with these relatively simplistic assumptions, the pressure drop calculated was 8.6 feet. A *

comparison with the less than 0.1 feet from the licensee's original calculation and a subsequent licensee calculation performed during the inspection using 50%

blockage which yielded approximately 0.5 feet of pressure drop illustrated the apparent non-conservatism of a simple percentage blockage analyses. It was also noted that the licensee's qualitative analysis regarding the low debris transportation velocities considered only the post-LO CA flow rates provided by the various pumps; however, initial relatively high LOCA blowdown flow rates, which would significantly increase the probability of debris transport to the sump, were not considered.

. The team also performed an informal structural scoping calculation on the screen supporting structure. This indicated that, at the 8.6 feet differential pressure from the scoping calculation, the load on the screens' structure was quite high, i.e.,

approaching its material limits. This was an area that also had not been evaluated by the license At the conclusion of the inspection the licensee generated CR No. C-PAL-98-1408 to address this concern and committed to perform rigorous quantitative analyses of the debris generation, transportation, and containment sump screen loading, and to reassess the resultant pump NPSH conditions as required by GL 97-0 Pending completion of these analyses and subsequent NRC review, this was an Unresolved Item (50-255/98012-02(DRS)). Unanticipated Effect of Ventilation System on Tank Level Indications and Engineering Safety Features Actuation System Setpoint IN 97-33, "Unanticipated Effect of Ventilation System on Tank Level Indications and Engineering Safety Features Actuation System Setpoint," described a condition where the level indication for the safety-related refueling water storage pool cha.nged as a result of an auxiliary building controlled area ventilation

• system starting during a test. This apparent level change resulted from the level instrument reference leg being vented to an area whose pressure changed when the ventilation system starte The licensee reviewed this document and concluded that, at Palisades, variations of compartment space pressure affecting level instruments was not a concern for two reasons: First, the automatic switching of ECCS pumps to the containment sump suction was actuated by safety injection refueling water tank level conductivity prob_~s. whose signals were independent of pressure. Second, it was stated that the only ventilation system that created either a positive or negative pressure was the control room heating, ventilation, and air conditioning (HVAC) system, and there were no safety-related level instruments located in spaces pressurized by the control room HVA The team questioned if the control room HVAC system was the only ventilation system that created a differential pressure in ventilated spaces. The licensee responded that the control room HVAC was the only emergency ventilation system that affected pressures in ventilated spaces. Since the auxiliar)t building HVAC system was non-safety-related, the initial response to the IN was correc The licensee further responded that, since the non-safety-related auxiliary building HVAC system had less capacity than the system described inthe IN, the effect on tank levels would be insignificant, and therefore, the initial response was still vali *

The team concluded that the licensee's initial response to this IN and the subsequent reviews in response to team questions were too narrowly focused, and as such may not have detected all potential problems related to GL 97-33 concerns such as the following:

The evaluation considered only spaces affected by safety-related HVAC systems; it should have considered instrumentation that could be affected by any HVAC system, regardless of its safety classificatio *

The evaluations considered only tank level_ instrumentation; the condition_

described could potentially affect any pressure instrumentation that could be affected by HVAC system *

The evaluations did not consider the various failure scenarios of the auxiliary building HVAC system and how these could potentially create greater-than-normal negative pressure, or even positive pressure, such as, the collective and individual failures of the supply and exhaust fans and various system damper *

The evaluations considered that this was a concern only for higher capacity HVAC systems; however, even small capacity systems had the potential to create high differential pressures with low building leakag * At the conclusion of the inspection the licensee agreed that the initial reviews were not thorough and committed to readdress this question taking a broader view and a more rigorous evaluation of potential problem areas. Resolution of this concern was identified as an Inspector Follow up Item (IFI) (50-255/98012-03(DRS)).

Potential for Ground-Level Radiation Release During the review of the licensee's evaluation of IN 97-033 described above, an additional concern was identified with the design and operation of the auxiliary building HVA The auxiliary building HVAC system is a non-safety-related system. During a LOCA, if offsite power and control air remained available, it would continue to operate normally, maintaining the auxiliary building at a slightly negative pressure exhausting to the environment through the plant vent stack. If high radiation was detected, the supply fan would be automatically tripped, potentially increasing the negative pressure, but the release would still be through the vent stack. The final safety analysis report (FSAR) values for accident offsite doses strongly indicated that a system operating in this manner had been assumed in the accident dose calculations. However, for a LOCA with a loss-of-offsite-power (LOOP), the system would cease to operate, allowing the building negative pressure to be lost, thus creating the potential for ground-level releases. This would substantially increase the offsite and control room dose Additionally, for certain failure scenarios, such as loss of control air pressure (the instrument air system was also non-safety-related), all air-operated system dampers would reposition closed, thereby pressurizing some building areas and ductwork sections relative to the outside environment. Individual failures of certain dampers could also cause even higher building pressurization. Pending licensee analysis of these scenarios and subsequent NRC review, this was identified as an Inspection Follow up Item (50-255/98012-04(DRS)). Diesel Generator Operability Evaluations IN 97-16 "Preconditioning of Plant Structures, Systems, and Components Before ASME Code lnservice Testing or Technical Specification Surveillance Testing,"

discussed various unacceptable test preconditioning practices obs*erved in the industry. Such preconditioning could render testing invali Included as an example in the IN was rolling over diesel generators with the cylinder blowdown petcocks open prior to testing to assure that there was no accumulated liquid in the cylinders that could damage the engine. In this case, the NRC concluded that the safety benefit of rolling the diesels outweighed the benefit of testing in the as-found condition. However, if liquid were detected in any of the cylinders, this would call into question the as-found operability of the diesel generator, i.e., would the quantity of liquid found have been sufficient to cause hydraulic lock that could have prevent the diesel from starting or could

• have caused damage? In order to properly assess this condition, any liquid detected would have to be quantified and an operability evaluation performe The team inquired if the diesel generator surveillance test procedure contained provisions to quantify any liquid detected and to perform an operability evaluation. An excerpt from Surveillance Test Procedu~e M0-7 A-1, Rev 48,

"Emergency Diesel Generator 1-1 (K-6A)," Section 5.2.3, was provided which described the rolling of the diesel with the petcocks open. It required that, for any moisture detected greater than a fine mist, the amount and the fluid type were to be recorded, and a CR was to be initiated. By initiating a CR, an operability evaluation would automatically be performed as a part of the CR proces Conclusions The team concluded that in general; response to industry initiatives was good; however, several cases* were identified where the responses could have been more thoroug Although the need for testing of molded case circuit breakers had been licensee identified in 1993, from review of industry operating experience information, a testing programwas not developed until 1997, after 44 of 72 molded case circuit breakers failed to trip during testing. The failure to assure that this condition adverse to quality was promptly identified and corrected was considered a violation.

Two examples were noted where Engineering concluded that the generic concerns were not applicable to Palisades because the conditions were not precisely the same as those at Palisades, rather than taking the broader view of how and where there were similaritie The team also noted one example of where a response to a generic concern went beyond the immediate issue. Surveillance Test Procedure M0-7A-1 for the EDG went beyond the specific warning contained in IN 97-16 to assure that any adverse condition found concerning liquid in the cylinders would be formally documented and evaluate E3 *

Engineering Procedures and Documentation E CFR 50.59 Program Review * Inspection Scope (37001)

The team reviewed implementation of the 10 CFR 50.59 program including procedures for screening changes, tests, and experiments (CTEs) and preparing safety evaluations; the processes for maintaining records of CTEs, reporting CTEs to the NRC and updating the updated FSAR; and training and qualifications of 1 O CFR 50.59 safety evaluation preparers and reviewers. In addition, the team reviewed 1 O CFR 50.59 screenings and/or safety evaluations associated with procedure changes and facility changes to the FSAR. Documents reviewed are listed at the end of this report. 1 O CFR

50.59 Safety Evaluations were also reviewed during the team's review of Design Change Packages, Modifications, and Temporary Modifications (See Section E3.2 of this report). Observations and Findings O CFR 50.59 Program The plant was performing 10 CFR 50.59 screening~ and safety evaluations in accordance with Administrative Procedure (AP) 3.07, Revision 9. The team reviewed the procedure and verified that the guidance in this procedure was in conformance with 10 CFR 50.59. The licensee used the updated FSAR to make 1 O CFR 50.59 applicability determinations and included the FSAR Change Request Log, Technical Specifications Change Requests, Design Basis Documents, and the proposed Improved Technical Specifications for the analysis. The program included an independent review of unreviewed safety question evaluations by the Nuclear Performance Assessment-Independent Safety Review Group following approval of 10 CFR 50.59 screenings and safety evaluations by the Safety/Design Review Sectio.

1 O CFR 50.59 Program Reporting Review 10 CFR 50.59(b)(2) requires that licensees submit a report containing a brief description of ariy CTEs including a summary of the safety evaluation of eac The team's review concluded that the reporting of completed safety evaluations to the NRC were as required by 10 CFR 50.59(b)(2).

. O CFR 50.59 Program Training Review The team reviewed the materials used in the training course for personnel that prepared and reviewed 10 CFR 50.59 safety evaluations and verified that.the information presented in the course was consistent with corporate procedures and NRC guidance. The training course, which included computer based training, appeared to be comprehensive. The inspector noted that in addition to successful completion of the training course, reviewers were required to be a General Engineerrrechnologist or above and have at least five years of nuclear industry experience, or to hold a senior reactor operator license with at least ten years nuclear industry experience, and to prepare a satisfactory 1 O CFR 50.59 screening or safety evaluation before becoming qualified to review 10 CFR 50.59 safety evaluations. The team concluded that the licensee had an good program for ensuring that trained and qualified personnel prepared and reviewed 10 CFR 50.59 screenings and safety evaluations.. O CFR 50.59 Safety Evaluation Review The team reviewed a sample of 10 CFR 50.59 screenings and *safety evaluations for the past two years and determined that, overall, the screenings and safety

• evaluations were appropriately prepared and were consistent with licensee procedures. In particular, the team determined that appropriate documents were reviewed during the preparation of 10 CFR 50.59 screenings and safety evaluations; the 10 CFR 50.59 screenings and safety evaluations adequately addressed the effects of the proposed changes on plant operations, interactions with other systems and components, any new failure modes, and the effects on accidents and transients; and the 10 CFR 50.59 safety evaluations adequately addressed unreviewed safety question criteri Conclu*sions E Overall, the team concluded that the 10 CFR 50.59 screenings and safety evaluations reviewed for the past two years were of good quality. In addition, the team concluded that the licensee had a good program to ensure that trained and qualified personnel prepared and reviewed 10 CFR 50.59 screenings and safety evaluation Design Change Packages. Modifications and Temporary Modifications Inspection Scope (37550)

• The team reviewed a sample of 24 modifications and nine temporary modifications to determine if they were designed, installed, and tested appropriately and whether canceled modifications were adequately justified. The team also reviewed whether temporary modifications were installed and removed in accordance with established procedures. In addition, the team reviewed 10 CFR 50.59 evaluations for the modifications, temporary alterations and procedure changes. Documents reviewed are listed at the end of this repor Observations and Findings Review of Modifications The team's review of modifications Nos. FES-97-094 and FES-970-095 produced questions regarding the adequacy of testing of two emergency diesel generator (EOG) breaker closure permissive auxiliary relays (one per diesel)

162-107X and 162-213X. The modifications had changed the original relays from Westinghouse type SG. to General Electric type HMA relays. These relays functioned to prevent the auto closure of the EOG breaker during an undervoltage condition; however, a test to verify this function could not be located. Subsequent to the on-site portion of the inspection, the licensee provided additional information on this issue as follows:

The purpose of this relay is to block closing of the EOG breaker for 1.5 *

seconds during a fast transfer from the Safeguards transformer to the startup transformer. A fast transfer also generates* an immediate trip

signal to the EOG breaker. During normal plant operations (diesel not running), if a fast transfer were to occur, the block signal would not be required as the EOG is not runnin *

During surveillance testing (diesel running), if a fast transfer were to occur, the EOG would trip and the relay would prevent the breaker from closing until three conditions are met: 1) the breaker from the safeguards*

transformer must be open; 2) the breaker from the startup transformer *

must be open; and 3) the EOG must have attained greater than 2000 volts. However, this is not considered a safety function since the EOG is considered inoperable during surveillance testin *

The original solenoid operated EOG breakers were replaced in 1995 with stored energy breakers. These breakers required approximately 300 cycles or five seconds for their springs to charge after opening or closin Since this is longer than the 1.5 second block signal, it does away with the any required function of the blocking relay. *

CR C-PAL-98-1507, "No Documentation of Functional Testing of Relay 162-107X (162-213X) was generated on August 12, 1998 to evaluate the current requirements for post maintenance testing and operability or full functional test The teani considered that this issue had been adequately addressed and had no further question.

Review of Temporary Modifications Temporary Modification No.98-009 installed a temporary strainer in the spent fuel pool (SFP) tilt pit drain in order to minimize the spread of contamination while working in the tilt pit. No technical concerns were identified with this modification; however, a procedural deficiency was identified as a result of review of this modificatio In reviewing the modification's 10 CFR 50.59 safety evaluation, the team noted that it stated that Off-Normal 'Procedure (ONP) 23.3, "Loss of Refueling Water Accident," Revision 3, directed the operators that, if level was being lost from the fuel pool, and the south tilt pit gate was not installed, to realign the spent fuel pool cooling system suction to the south pit. The procedure als9 allowed a system suction lineup to the reactor cavity tilt pit. *

The team was concerned that this procedure had no direction to the operators to identify the cause of the level loss before realigning the system suction. If the cause of the level loss was a breach in the SFP cooling system pressure boundary, this realignment would allow loss of inventory down to the sill of the SFP gate rather than the maximum inventory loss of only approximately four feet if the system suction were left at its normal point near the pool surface. With the water level at the gate sill, in addition to normal cooling being lost, there would be only approximately one foot of water above the top of the fuel assemblies

(calculated fuel pool area radiation level - 893 Rem/hour), for the design basis pool heat load there would be only slightly more than one hour until the pool would boil and approximately one more hour before fuel assemblies would begin to be uncovered due to evaporative losse The licensee acknowledged the weaknesses in this procedure and offered the fact that when it was written, the only failures envisioned that could cause such water losses were a reactor cavity seal failure or a steam generator nozzle dam failure, in which case switching the suction point would not matter. However, since the procedure, as written, would cover level loss due to any cause, its structure as found was inadequate to cover any cause. Pending NRC review of licensee action on this procedural weakness, this was considered an Inspector Follow up Item (50-255/98012-05(DRS)). Conclusions The team reviewed 24 modifications and nine temporary modifications and concluded that they were of good quality, properly installed and teste E7 Quality Assurance in Engineering Activities E7.1 General Comments Inspection Scope (40500)

The inspectors reviewed the licensee's assessment activities to evaluate the effectiveness of licensee controls in identifying, resolving, and preventing issues that degrade the quality of plant operations or safety. These controls included the corrective action and self-assessment programs, implementation of timely and effective resolution of technical issues, active involvement in ensuring the reliability of plant systems, and awareness of industry events and how they impact the plan The inspectors selected a sample of issues/problems for detailed analysis to assess the licensee's ability to identify and correct problems. Additionally, the inspectors evaluated the process for initial identification and characterization of the specific problems, elevation of the problems to proper levels of management for resolution, disposition of any operability/reportability issues and implementation of corrective actions, including evaluation of repetitive conditions. Items reviewed included:

(1)

Deficiencies requiring safety evaluations, root cause assessments or operability determination (2)

QA audits and self-assessment (3)

Deficiencies tracked in the corrective action programs, including the evaluation of deferred items, or interim resolution *

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(4)

Results of licensee audits that evaluated the effectiveness of the associated corrective action program (5)

Interviews with selected individuals involved with the problem identification process to determine the extent of the individual's understanding of the process and willingness to report problem The documents reviewed are listed at the back of this repor Observations and Findings The inspectors reviewed a number of licensee documents which focused on compliance of the plant departments with applicable requirements of the corrective action process, audits, and self assessments. The inspectors noted that these documents identified both strengths and areas requiring improvements. The team noted that the level of detail in the reports indicated an in-depth review of the issue Conclusions Based on interviews with station personnel and review of the above documents which indicated that problems were being identified and corrective actions for those problems were being implemented, the inspectors concluded the licensee's corrective action, audit, industry experience and self-assessment programs were effective. The inspectors considered that quality assurance activities were of appropriate depth and scop E Corrective Action Process Inspection Scope (40500)

The team assessed the Corrective Action Process (CAP) through review of implementing procedures, CRs, corrective action management reports, corrective action effectiveness reviews, Corrective Action Review Board (GARB) and Condition Review Group (CRG) activities, and action taken for previously identified trends. Documents reviewed are listed at the end of this report. The team also attended one GARB meeting and one Plant Review Committee meeting during the on-site inspection period and interviewed cognizant personnel concerning the corrective action and CR processes. In addition, the team assessed corrective actions taken for problems previously identified in resident report Observations and Findings The t~am reviewed Procedure 3,03, "Corrective Action Process," Revision 20, which described the methods used for documenting problems and the corrective action process. This procedure described the use of the CR for problem identification and tracking and indicated that a CR would be categorized as level 1, 2, 3 or 4 based on the importance and priority of the problem. Level.1 CRs were used to document the most

significant problems, and Level 2, 3, and 4 CRs, those problems of decreasing importance and priority. Problems documented on Level 4 CRs did not require cause investigations and actions to prevent recurrenc The team noted that during two Joint Utility Management Audits (JUMA) conducted in 1996 and 1997, the licensee identified that improvements in managing the CR process were necessary. Other issues identified during these audits were that internal audit reports lacked definitive support for some findings; the backlog of low significance CRs was high; and that feedback to the CR initiator after a CR had been closed out was weak. The team concluded that the licensee's response to these issues was prompt and effectiv The team noted that in response to a JUMA audit finding relative to the high ba.cklog of low significant CRs, the licensee established new expectations and revised the applicable plant procedures to reflect these expectations. The expectations for closure of Level 3 and 4 CRs were 30 to 40 days. The team noted that since 1996, close out times for Level 4 CRs was reduced from 105 to 61 days and for Level 3 CRs from 151 to 122 days. The team also noted that the threshold for identifying problems via CRs was considered low and the number of CRs generated was high. Of the sample of CRs reviewed by the team, the root cause analyses appeared to be thorough and effectiv When corrective actions were extended, due dates were appropriately extended..

The team also reviewed a Corrective Action Log, dated February 11, 1998, which listed condition reports issued from October 1997 to March, 1998. The list contained 775 CRs which had been *written during this period. The number of CRs w~itten arid a cursory review of the type of problems documented indicated that the threshold for writing CRs was appropriately low. A listing of open CRs, which included scheduled completion dates, was also reviewed. None of the listed CRs indicated a major problem with the completion date assignment The team reviewed the licensee's Correction Action Review Board (CARS) charter and noted that the CARS had been established to review CRs which had been categorized as Level 1 or 2: The duties of the CARS included evaluating the appropriateness of operability and reportability determinations and assuring that appropriate immediate corrective actions were taken to resolve these important matter On July 7, 1998, the team attended a meeting of the Condition Review Group (CRG) to observe their review of CRs issued the previous day. The discussions during this meeting appeared to be appropriate and individuals were assigned to follow and expedite required actions. The CRG appeared to be a valuable tool to ensure prompt and thorough management review of significant problems. A total of 9 condition reports were reviewed/evaluated during the CRG meeting. While none of these CRs reached the level that required CARS reviews, the CRG appeared to be effective in its oversight of CR prioritizations and corrective action recommendation The team also observed several noteworthy practices that contributed to the effectiveness of the corrective action process. Palisades employed trending at all

~ *.

levels, including monthly corrective action management reports and periodic NPAD audits of corrective action effectiveness. Other examples included daily reviews of new CRs, the review of existing significant CRs, generation of quarterly reports of performance indicators, and senior management involvement in all Level 1 and 2 CR The team observed that the corrective action process at Palisades* had improved since enhancements to the corrective action process were implemented in 1996 and 199 Problems were identified via the CR process, the more significant issues were investigated for root causes, trends were identified and tracked, significant corrective actions received interdiscfplinary review through the CARS, observations were made in the field to improve problem prevention, and the overall collective significance of issues and trends was assessed quarterly. A review of the past two-year period indicated that corrective action process improvements at Palisades have been effectiv Conclusions The corrective action program at Palisades had shown improvements in identification,

. resolution, and prevention of problems in the past two years. Personnel interviewed indicated a willingness to identify problems, considered the process to be owned equally by all plant staff, and did not consider CRs written against themselves to be negativ Overall, the licensee has been effective in the identification and resolution of problem E Industry Experience Program Inspection Scope (40500}

The team evaluated the adequacy of the licensee's programs that implement industry experience information. Documents reviewed are listed at the end of this repor Observations and Findings The team reviewed AP 3.16, "Industry Experience Review Program," Revision 5, and noted that the procedure provided adequate guidance for ensuring that industry operating experience was integrated into the plant operating, engineering and maintenance activities. The team observed that CRs doc1,Jmented that system engineers had analyzed events and information at other plants for applicability to Palisades. The team also observed that the quarterly system health assessments incorpor~ted the industry operating experience The team noted that the Industry Experience (IE) group under the leadership of the Industry Experience Coordinator provided the initial screening of all industry experience documents. As possible relevant subjects applicable to Palisades were identified, the IE group coordinated a review/evaluation effort and assessed the adequacy of these reviews/evaluation *

The team observed an IE group daily meeting to discuss industry current news and the status of on-going industry experience evaluations and the backlog of these. evaluations.

The team was informed that on a weekly basis, the IE group discussed on-going industry issues at the Managers' Morning Meetin Additionally, the licensee informed the team that they were addressing a number of issues identified with the IE program. These issues were: 1) Palisades was not internally using IE information effectively to prevent occurrences or improve the process, 2) Palisades was not always consistent or timely in reporting events to the industry data bank and 3) Self-Assessments of the IE program were ncit sufficiently critical to identify/quantify the Program's effectiveness. The team noted that the licensee's on-going responses to these issues appeared to be adequat The team reviewed several CRs and noted that industry experiences were being properly evaluated and addressed. For example, the team reviewed condition report C-PAL-98-0700 titled, "Parker Fittings Used with Swagelok Fittings," dated April 26, 1998 that correctly identified and evaluated the root cause for this issues. Specifically, the licensee accurately noted that the issue of interchanging hardware from different manufacturers may potentially damage the fittings. However', the team also identified.

examples of less than adequate response to industry initiatives (See Section E2 of this report). Conclusions The team concluded that the licensee's program for screening, analyzing and dispositioning industry experience issues appeared to be effective; however, two examples of where responses could have been better were identified during the engineering revie E Self-Assessment and Audit Activities a,

Inspection Scope (40500)

The team evaluated the effectiveness of the licensee's self-assessment capability by reviewing department self-assessment reports, quality and self-assessment (Q&SA)

quarterly self-assessment reports, and Q&SA audits. In addition, the team interviewed cognizant personnel. Documents reviewed are listed at the end of this repor Observations and Findings The team reviewed Procedure No. 1.09, "Self-Assessment," Revision 4, which was the implementing procedure for the self-assessment program. The procedure was revised to provide better guidance regarding training of self-assessment evaluators, management's expectations for the respective plant departments, assessment criteria and assessment objectives. The team concluded that the new procedure appeared to be more rigid on how self-assessments should be conducted which should provide further improvements in the Palisades Self-Assessment Progra The team sampled a number of self-assessment activities conducted by the various plant departments over the previous year and concluded that improvements were being made in this area. For example, the Maintenance, Planning and ~cheduling, Operations and Engineering Departments were all generating good quality self-assessment report These reports contained the purpose and scope for the assessment, personnel assignments, standards/expeCtations assessed against, facts supporting where deviations from standards. or expectations existed, conclusions based on identified facts, and recommendations for performance iniprovement The team's initial observation of several Operations self-assessments suggested that the assessments did not provide enough details of the identified problem or what corrective actions were needed. Further review of this matter revealed that the licensee had self-identified these concerns and had initiated the appropriate corrective action The team verified that these issues were addressed in the department's self-assessment report dated March 23, 199 The team also noted the following strengths and improvement initiatives relate.d to the Self-Assessment Program: A number of plant management personnel had received valuable experience during rotational tours of duty at INPO; Engineering Aid'

Administrative Procedure' EGAD'.'ADM-08, "Guidelines for Performing Self-Assessments," was in the process of being issued; "Hot Button" reports were being utilized to identify adverse trends; and Operations had implemented detailed guidance for Management Monitoring of Supervisory Skill The team reviewed the licensee's audit logs and schedules and found that they adequately covered the appropriate plant activities. Records of selected audits that were reviewed indicated that the audits were adequately performe Conclusion The team concluded that the Self-Assessment Program was effective and capable of providing valuable performance insights. The team also found that the audit program covered the required areas and was identifying problems and concerns. Audit findings were documented on conditio_n reports, which were used for tracking and to obtain corrective actions. Areas of concern identified by audit findings were promptly and effectively correcte E Miscellaneous Engineering Issues E (Closed) Violation 50-255/94002-01: Five examples of SWSOPI Inadequate corrective action which represented a significant breakdown iri control of the corrective action program. Failure to recogniz~ the significance of and to take corrective actions to resolve the single failure vulnerability of ESS pump seal cooling and lubrication heat removal, wtiich'C;ould result in eventual ESS pump failure. Failure to take prompt corrective action to incorporate Non-critical Service Water System header isolation valve, CV-1359 onto a leakage test progra*m. Failure to appropriately question and take action to evaluate the seismicity of bent instrument and unistrut supports routed in front

~

• of the CCW HXs. Failure to recognize the significance and take corrective actions to couple the Service Water lnservice Test pump reference values to the Service Water flow balancing test. Failure to take prompt corrective actions to resolve a concern that Service Water System flow verification test T-216 balance flow to the CCW HXs at or very near their required flow rates and did not allow for pump degradatio The licensee responded to this violation on June 6, 1994, committing to corrective actions in management dire~tion and oversight, upgrades to th.e corrective action process, and evaluation of th~ failure to take adequate or prompt corrective actions to the significant conditions identified during the service water operational performance inspection. In addition, corrective actions for the five specific examples cited in the Notice of Violation were addresse *

The team reviewed the corrective actions to the specific examples and Event Report E-PAL-94-012, "Service Water System Operational Inspection (SWSOPI) - Inadequate Corrective Action," which evaluated the broad issue of failure to take adequate or prompt corrective action with a more global perspective. The team considered the corrective actions to be acceptable and comprehensive. This item is close E (Closed) Violation 50-255/95010-01: Failure to use an updated and controlled wiring list. An uncontrolled wire list was used to implement Facility Design Change FC-88 This modification was initiated because of the high maintenance and obsolescence of.

. the reactor protection system (RPS) trip logic. The licensee sent an old Wire list that had not been updated to the manufacturer. This wire list contained 12 incorrect module circuit connections that bypassed the six matrix logic channels for the RPS containment high pressure (CHP) trip function. The errors in the CHP trip logic occurred, in part, because the original plant design did not include a CHP trip. Although Combustion Engineering was aware of the CHP trip, they installed printed circuit jumpers that bypassed the RPS matrix trip logic for containment high pressure on all four independent safety channels (A, B, C and D).

• The licensee's corrective actions included revising the administrative procedures to clearly establish the roles of design engineers when implementing design changes or modifications. Changes were also made to the design change program that required that information in vendor manuals, drawings and wiring lists shall be verified and validated prior to its use in design changes. The team verified that the corrective *

actions were acceptable. This item is close E (Closed) Violation 50-255/95010-02: The post modification test for Facility Design Change FC-888 was not suitable to verify or check the adequacy of design. During implementation.of FC-888, the licensee introduced an inadvertent change to the existing RPS matrix logic that bypassed the CHP trip; however, the post modification test was inadequate because it did not include testing to determine whether any portion of the two out of four (2/4) RPS CHP trip logic was functional. The licensee determined that the inadequate post modification test occurred because of the reliance on the Technical

Specification (TS) surveillance that did not test all combinations of the RPS trip logic. In addition, the licensee concluded that the role of the system engineer was not clearly defined when determining the adequacy of the post modification tes As part of the corrective actions, the licensee modified the RPS matrix channels to restore the CHP trip function. The TS surveillance test for the RPS matrix logic was revised to provide adequate testing. This assured that the requirements of the TS Table 4.17.1 were being adequately verified during the monthly test. Additionally, the licensee assigned responsibility for review of all modification testing to system engineers who will act as the testing authority. This item is close E (Closed) Violation 50-255/95010-03: The licensee did not demonstrate operability of the six matrix logic trip unit channels for high containment pressure from April 11, 1992, to May 22, 1995. The surveillance test required by TS 3.17 and as implemented by Procedure No. M0-3, "Reactor Protection Matrix Logic Tests," had not demonstrated operability of the RPS trip logic during this period. This occurred because the 2/4 logic combinations for only 1 of the 11 trips on the RPS logic system was testing at random each month. Consequently, if a channel such as the high power trip, high rate trip or high containment pressure were to become inoperable, surveillance testing would not promptly identify this condition because the testing n:iethod was random and only one channel was selected each mont The team verified that t.he licensee changed surveillance procedure No. Ml-3, "Reactor Protection Logic Tests," Revision 0, so that all 11 channels and each of the 4 RPS logic combinations (A, B, C and D) were checked. The licensee now tests monthly each combination (AB, AC, AD, BC, BD and CD) in the ladder*logic to provide adequate overlap in testing for all 11 RPS trips. The team reviewed a completed copy of the last RPS surveillance test and found it to be acceptable. This item is closed:

E (Closed) IFI 50-255/94014-53: Primary coolant system (PCS) cooled below 70°F temperature limit. As discussed in inspection report 50-355/94014, the PCS was cooled to below the temperature limit of 70°F (21°C) on two occasions with the reactor vessel (RV) head bolts fully tensioned. During this inspection, the team reviewed Engineering Analysis EA-D-PAL-94-170-01, "Impact of Less Than 70°F Shutdown Cooling Water on Reactor Vessel," which concluded that no ASME code limits for the RV had been exceeded. Consequently, there had been no cumulative effects on the ductility or integrity of the RV. This item is close *

E (Closed) IFI 50-255/94014-60: Lack of Design Basis Information/No Clear Holes &

Responsibilities~ The causes for weaknesses included a historical lack of design basis information, lack of clearly defined roles and responsibilities between NECO and System Engineering, ineffective technical reviews, and an ineffective process to assure documents, processes, and activities affected by the modification were appropriately revise *

The team reviewed Palisades Performance Enhancement Plan (PPEP) action items 1.2, 2.4, and 4.2. The Design, Systems, Programs and Plant Support Engineering groups

• now report to the Plant Engineering and Modifications Manager. A multi-discipline review and required training and a qualification card process was instituted to address the ineffective technical reviews. Administrative Procedures (AP) 9.00, "Design Engineering and Configuration Management Program Description and AP 9.03, "Facility Change," were rewritten and strengthened in the area of updated of documents and proce.sses affected by modifications. This item is close E8. 7 (Closed) I Fl 50-255/94014-61: Plant Configuration Control Weaknes The team reviewed PPEP action items 2.4, 4.2, and 5.1. AP 9.0 was revised fo add specific management expectations to specific engineering departments and an overview.

of the Design Engineering and Configuration Management Process. In addition, the plant now has a dedicated Configuration Control Manager and department. This item is close E (Closed) LER 50-255/91017: Potential inter-system loss of coolant accident (ISLOCA)

with primary coolant pump. As discussed in LER 50-255/91017-00, on August 5, 1991, the licensee identified that a postulated break in the PCP integral heat exchanger could result in a primary coolant system leak outside the containment building. Four identical

. Byron-Jackson primary coolant pumps.are installed at Palisades. The primary coolant at the integral heat exchanger is pressurized to about 2060 psia; the component cooling water (CCW) system has a design pressure of 150 psig.. In the event of a postulated ISLOCA, primary coolant would enter the CCW and pressurize the CCW system beyond its design pressure resulting in a potential for leakage path outside the containment buildin The licensee performed a probabilistic risk assessment PRA of this scenario to determine the likelihood of core damage occurring from the postulated event. The initiating event frequency was determined to be 100 times smaller than that of a small break LOCA. The licensee determined that the risk of core damage could be limited to acceptable levels by updating operator training on how to respond to the ISLOCA. The team verified that the scenario was included in the training module for continued licensed operator training in Instructor Lesson Plan LOCT01.93D, "Mitigating Core Damage Review." This item is close E (Closed) LER 50-255/94003: Lack of environmental qualification for the position switches for the service water inlet and outlet valves to the containment air coolers (CACs). As discussed in LER 50-255/94003-00, on February 3, 1994, the licensee identified that the position switches for the CAC service water inlet and outlet valves were not environmentally qualified for submergence. The valve position switches provide the operator with CAC service water inlet and outlet valve status indication as one of the inputs to verify operability of the containment heat removal system. The licensee performed a~ operability determination and determined that the service water inlet and outlet valves would be operable in the event of the failure of the position switches. In addition, justification was provided to show that submergence would not have an effect on the switches until after the valve positions were verified and documented. The switches were qualified for submergence for a period of three hour * E8.10 EB.11 E8.12 This provided ample margin since the switches that change position will be verified approximately 30 minutes after the start of an accident. This item is close (Closed) IFI 50-255/93020-01: Evaluation of fuel assembly 1-024 failure. As discussed in inspection report 50-255/93020(DRS) the licensee had not completed the root cause analysis in time for the teams assessment and inclusion into the report. On September 30, 1993, the licensee provided a detailed description and the results of the root cause analysis in a letter to the NRC. In addition, on October 14, 1993 the licensee issued a supplemental LER to include a summary of the results of the root cause analysis which was performed fo determine the reasons for the failure of fuel assembly 1-024 and the subsequent discharge of nuclear fuel into the primary coolant of the Palisades reactor. This supplemental LER was closed in report 50-255/98006(DRS).

This item is close (Closed) IFI 50-255/94014-63: Fuse control program weaknesses. As discussed in *

inspection report 50-255/94014(DRP) the Palisades Performance Enhancement Program (PPEP) Action Plans 2.2 and 4.2 required further NRC review and evaluatio Action plans 2.2, "Establish an Improved Planning and Prioritization Process," and 4.2,

"Establish Strong Sensitivity to Design Basis," were reviewed and evaluated by the team. In addition, action item A-PAL-94-152, "Calculations For Fuse Size and Type" *

and guideline EGAD-ELEC-10, "Sizing of Control and Power Fuses," were also reviewed. Palisades Maintenance Standards Handbook uses references from AP 4.02,

"Control of Equipment," and EGAD-ELEC-10 to address fuse replacement. This item is close (Closed) LER 50-255/93007: Degradation of boraflex neutron absorber in surveillance coupons. As discussed in LER 50-255/93007-02, on August 17, 1993, the licensee identified that a Boraflex surveillance coupon, upon removal form the spent fuel pool (SFP), had disintegrated approximately 90%. Boraflex is the trade name of a boron impregnated, polymer-based sheet material that is utilized as a neutron absorber in the construction of SFP storage racks. The use of Boraflex allows minimal center to center cell spacing in the SFP storage rack The cause of the event was determined to be flow induced deterioration of the full length surveillance coupons due* to inadequate holding canister desigr:i. The corrective actions included completion of neutron attenuation testing (blackness testing) on the SFP racks and a change to the surveillance method for verification of rack Boraflex conditio Blackness testing is the most effective method of boraflex surveillance since it "looks" at the actual rack Boraflex panels and does* not rely upon secondary methods such as surveillance coupons. Blackness testing uses a neutron source to verify the presence of Boraflex in the walls of the spent fuel rack. This item is close EB.13 (Closed) LER 50-255/96002-01: Initiation of TS required shutdown due to safeguards cable fault As discussed in LER 50-255/96002-01, on January 16, 1996, the licensee identified a loss of the safeguards power source caused by a phase-to-phase fault in the 2400V AC safeguards bus.. The loss of the safeguards power source placed the plant into a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> TS action statement which led to a plant shutdow *

E8.14 E8.15 X1 The cable failure was determined to be caused by localized water and contaminant treeing (treeing is a condition where microscopic voids in cable insulation look like tree branches when a wafer-thin cable insulation sample is viewed under a microscope)

which initiated in the cable insulation. The initiating points for the treeing degradation were localized foreign matter (contaminants) and voids found in the cable insulatio Corrective actions included cable replacement, testing of the other cable, and a review of plant equipment response. This item is close (Closed) LER 50-255/96013-01: Testing of four molded case circuit breakers (MCCBs)

revealed that the breakers would not trip on overcurrent. These failures resulted in concerns that all 72 DC MCCBs could fail to trip when subjected to a short circui Eventually all 72 breaker were tested and the licensee found that all 44 magnetic only MCCBs would not trip when subjected short circuit currents. The remaining 28 breakers were of the combinatiOn thermal-magnetic type and these breakers tested within the manufacturer requirements. The licensee replaced all 72 breakers and established a preventative maintenance program on December 1, 1997, to test approximately a third of the breakers every refueling outage so that all the breakers would be tested in a six year period. This item is closed (See Section E2.1 ).

(Closed) Unresolved Item 50-255/96018-01: The failure of the MCCBs and related problems was considered unresolved pending the licensee investigation and determination of the extent and significance of the problem. This URI was determined to be a violation in Section E2.1. This item is close Exit Meeting Summary The inspector.presented the inspection results to members of licensee management at the

• conclusion of the inspection on July 24, 1998. The licensee acknowledged the findings presented. *

The inspectors asked the licensee whether any material examined during the inspection should be considered proprietary. No proprietary information was identified.

PARTIAL LIST OF PERSONS CONTACTED Licensee D. DePuydt, Design Engineering R. DesJardins, Design Engineering R. Gerling, Manager Design Engineering K. Haas, Engineering Director N. Haskell, Licensing Directer M. Nordin, Design Engineering K. Osborne, Engineering Programs T. Palmisano, Site Vice President D. Rogers, General Manager - Plant Operations G. Szczotka, Manager NPAD K. Toner, Licensing Supervisor S. Wawro, Director Maintenance and Planning R. Westerhof, Configuration Control IP 37001:

IP 37550:

IP 40500:

. IP 92700:

IP 92703:

INSPECTION PROCEDURES USED 10 CFR 50.59 Safety Evaluation Program Engineering Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Onsite Follow up of Written Reports of Non-Routine Events Follow up - Engineering

• ITEMS OPENED, CLOSED, AND DISCUSSED Opene /98012-01 VIO Failure to implement testing of MCCBs in a timely manne /98012-02 URI Licensee to perform rigorous quantitative analyses of the debris generation, transportation, and containment sump screen loading and to. reassess resultant pump NPSH conditions 50-255/98012-03 IFI Licensee to re-evaluate IN 97-33 50-255/98012-04 IFI Licensee to evaluate the potential for ground-level radiation release 50-255/98012-05 IFI Licensee to evaluate weakness identified in ONP 23.3, "Loss of Refueling Water Accident" Closed 50-255/94002-01 VIO Five examples of SWSOPI Inadequate corrective action which represented a significant breakdown in control of the corrective action progra /95010-01 VIO Failure to use an updated and controlled wiring list 50-255/95010-02 VIO Failure to perform an adequate post modification test 50-255/95010-03 VIO Failure to implement and perform an adequate surveillance test 50-255/96013-01 LER DC MCCB Failures 50-255/96018-01 URI DC MCCB Failures 50-255/94014-53 IFI PCS cooled below 70°F temperature limit 50-255/94014-60.

IFI Lack of Design Basis Information/No Clear Roles &

Responsibilities 50-255/94014-61 IFI Plant Configuration ControfWeakne*ss 50-255/94014-63 IFI Fuse control program weaknesse *

50-255/94003 50-255/93020-01 50-255/93007 50-255/96002-01

LER Lack of environmental qualification for position switches IFI Evaluation of fuel assembly 1-024 failure LER Boraflex degradation LER Initiation of Technical Specifications {TS) required shutdown due to safeguards cable fault

• LIST OF ACRONYMS USED A/E Architect Engineer AC Alternating Current AP Administrative Procedure CAC Containment Air Cooler CAP Corrective Action Program CARB Corrective Action Review Board CCW Closed Cooling Water CFR Code of Federal Regulations CHP. Containment High Pressure CRG. Condition Review Group CTE Changes, Tests, Experiments CR Condition Report DC Direct Current DRS Division of Reactor Safety E& TS Engineering and Technical Support ECCS Emergency Core Cooling System EOG Emergency Diesel Generator FSAR Final Safety Analysis Report GL *

Generic Letter HVAC Heating, Ventilation, and Air Conditioning IFI Inspection Follow Up Item IN Information Notice ISLOC lntersystem Loss of Coolant Accident JUMA Joint Utility Management Audit LER Licensee Event Report LOCA Loss of Coolant Accident LOOP Loss of Offsite Power M&TE Measuring and Test E<:iuipment MCCB Molded Case Circuit Breaker NPSH Net Positive Suction Head PDR Public Document Room PPEP Palisades Performance Enhancement Program PCS. Primary Coolant System

. RPS Reactor Protection System RV Reactor Vessel SFP Spent Fuel Pool TS Technical Specifications URI Unresolved Item V

Volt VIO Violation

PARTIAL LIST OF DOCUMENTS REVIEWED Procedures 1.09, "Self-Assessment Program," Revision 4 *

3.03, "Corrective Action Process," Revision 20 3.07, "Safety Evaluations," Revision 9 3.15, "Design Basis Document Maintenance," Revision 6 3.16, "Industry Experience Reyiew Program," Revision 5 4.00, "Operations Organization, Responsibilities and Conduct," Revision 20 5.07, "Control of Measuring and Test Equipment," Revision 8 10.41, "Procedure Initiation and Revision," Revision 27 Ml-3, "Reactor Protection Logic Tests," Revision 0 M0-7A-1&2, "Emergency Diesel Generators 1-1 & 1&2," Revision 6 M0-7 A-2, "Emergency Diesel Generator 1-2 (K-6B)," Revision 46 SOP-22, '.'Emergency Diesel Generators," Revision 24 T-SC-95-090-01, "SIS Actuation Logic Modification Test," Revision 1 COP 3,"Determination of T-103 Volume to Add to Sump Following a LOCA," Rev 8, Attach 5 SAP 4, "Containment Spray and Iodine Removal System," Revision 8 GOP-2, "Plant Heatup (Cold Shutdown to Hot Shutdown)," Revision 20 SOP-1, "Primary Coolant System," Revision 41 EOP-9.0, "Functional Recovery Procedure," Revision 9 EOP-4.0, "Loss of Coolant Accident," Revision 9 *

ONP-23.3, "Loss of Refueling Water Accident," Revision 3 EGAD-ADM-08, "Guidance for Performing Self-Assessments," Revision O Condition Reports C-PAL-93-0017, "Develop Program for Periodic Testing of Molded Case Circuit Breakers" C-PAL-94-0564, "Diesel Generator Failed to Reach Full Load" C-PAL-96-1453, "72-228 Breaker Failed Lab Services Test" C-PAL-96-0885, "Pressure Control Valves Out-of-Tolerance" C-PAL-97-0087, "Main Generator Wire Insulation Damage at Current Transformer" C-PAL-97-0~10. "Breaker 52-2239 Failed During Testing" C-PAL-97-1112, "Loss of M&TE Control" C-PAL-97-1261, "Found Breaker 52-277 Out of Tolerance" C-PAL-97.,.1430, "Inadequate Implementation of Battery Calculation" C-PAL-97-1566, "Diesel Generator for Design Base Load Range Incorrect" C-PAL-97_1568, "Failure to Verify Calibration of 152-105 and 152-106 Breaker Relays" C-PAL-97-1596, "Station Batteries Unanalyzed for OBA Conditions with Battery Chargers Crosstied" C-PAL-97-1619, "Electrical Engineering Calculations Were Not Updated" G~PAL-97~_t620,_"Lack of Analysis for 125 VDC Loads*at Degraded Voltages" C-PAL-97-1652, "125VDC Distribution Panel Breakers'Short Circuit Capability" C-PAL-97-1656, "DC Load Flow with Incorrect Temperature Correction Factor

C-PAL-97-1685, "*Documentation for Station Battery Short Circuit Current not Available" C-PAL-98-0724, "Peak Inrush Motor Current Exceeds Motor Locked Rotor Rating"

• C-PAL-98-1153, "Test Overcurrent Relays for Breakers 152-105 and 152-106" C-PAL-98-1358, "Essential Safeguards Room Cooler Breaker Found Tripped" C-PAL-96-1648, "HPSI Operabilityin Question Immediately Following RAS Until Subcooling Established" C-PAL-97-0091, "Main Steam Safety Valve Inlet Line Losses" C-PAL-97-0582, "Potential Preconditioning of EDGs During M0-7A-112" C-PAL-97-1209, "Less Than adequate Guidance for DIG Room Temperature at Which to Repower the DIG Rooms Ventilation Fans" C-PAL-97-1210, "Diesel Injection Tube Part 21 Notification" C-PAL-97-1396, "CCW Pump Discharge Pressure Below SOP16 Requirement" C-PAL-98-0394, "Lack of Overpressure Protection Vulnerability in HP Air System to ESS Valves" C-PAL-98-0733, "VHX-2 and VHX-4 Outlet Valves Failed ASME Section XI Testing" C-PAL-96-0664, "Design and Installation Questions Associated with C-3B, "Diesel Generator 1-2 Air Compressor" Replacement Under FES-95-239" C-PAL-97-0196, "No Calculations to Support FSAR Functional Statements" C-PAL-97-1450, "SOP-2A Lacks Adequate Guidance to Adjust Purification Demineralizer DIP" C-PAL-98-1323, "Potential Lifting of Letdown Relief Valve RV-2013" C-PAL-97-1438, "PCS Leakage From Letdown Relief Valve" C-PAL-96-0853, "Purification Ion Exchanger High DIP" C-PAL-98-0121, "Failed Fuel Monitor Low Flow Condition" C-PAL-97-0919, "Charging Pump P-55B Seal Lube Return Line Plugged" C-PAL-97-1027, "Nitrogen Station 3B Relief Valv~ Failures" C-PAL-97-1116, "Documentation of Design Function and Operability of Guard Pipe on Containment Sump Suction" C-PAL-97-1172, "Debris Found in Containment" C-PAL-97-1370, "CCW Flow to ESS Pumps Not Analyzed for IST Pump Degradation in Cal Of Record" C-PAl-97-1499, "Improper Air Pressure Used for-Actuator Capability Calculation in T-372" C-PAL-96-0883, "Containment Spray Pump & Sump Check Valve Flow Rate Discrepancies" C-PAL-97-1571, "Potential Flow Paths That Bypass Containment Sump Screens Following a OBA"

C-PAL-98-1408, "Adequacy ofECCS Pump NPSH Under Increased Screen Blockage" C-PAL-98-0007, "ONP 6.2 (Loss of Component Cooling) Immediate Action not Performed" C-PAL-98-0032, "ONP 6.2 Inadequate for Loss of Inventory" C-PAL-98-0573, "Inadequate Job Preparation (FW Bypass Valve Position)"

C-PAL-98-1160, "Steam Generator Level Below SOP-7 Requirement" C-PAL-98-1316, "Work Order Readiness Review Inadequate" C-PAL-98-1358, "RCS West safeguards Room Cooling V-27C Breaker Found Tripped" C-PAL-98-1409. "Manual Reactor Trip due to Loss of "A" Main FW Pump" Modifications EAR 96-0123, "Configuration Problem with Breaker 52-8226" EAR 96-0196, "Temporary Covers for Containment Floor Drains" EAR 96-0214, "Breaker Replacements"

EAR 96-0264, "15 VDC Power Supply for ComparatorlAverager"

• EAR 97.,.0592, "Pressure Switch Model 680 No Longer Available" EAR 98-0303, "LS-1440 Wiring Problem" FES96-059, "Calculation for Second Level Undervoltage Time Relay" FES97-092, "Replacement of Load Shedding Relay 194-108 due to SQUG Relay Program" FES97-095, "Replacement of Diesel Breaker Auto Close Permissive Relay 162-213X" FES97-107, "Replacement of Circuit Breakers on Panel D11A" FES98-035, "Replacement Power Supply for Comparator/Averager" SC 95-090, "Provide Left Channel Safety Injection Signal Actuation" EAR 98-0008, "Containment Sump Vent Screen" FES98-048, "Allow Use of HeliCoil Inserts in Steam Generator Secondary Side Handhole Stud Mounting Holes" EAR 98-0259, "SW Pump Impeller Metal Buildup" FES98-031, "Traveling Screen Basket Material Change" FES98-001, "MSIV Disk Assembly Belleville Washers" EAR 96-0576, "Sealing Pipe Penetrations in Lube Oil Storage Blockwall" EAR 95-0268, "Circulating Water Discharge Radiation Detector" EAR 96-0809, "Pressure Regulator Has a Blown Diaphragm" EAR 96-0645, "Justification of Heat Exchanger Cover Plant Thickness" EAR 95-0512, "LPSI Pump Seal Enhancement" EAR 96-0727, "Safety Injection Refueling Water Tank Low Level Switches" EAR 97-0634, "Document Available NPSH for 200°F CCW Temperature" Temporary Modifications TM 97-027, "Install Mechanical Blocks to Hold Damper Open" TM 97-040, "Install a1ank Flange on Discharge Piping" TM 97;.052, "Disable Low Flow Alarm for Failed Fuel Monitor" TM 97-017, "Install Jumper to Bypass 25F7 Interlocks to Disconnect Switch" TM 98-018, "Install Hydrostatic Plug for In-Core Instrumentation" TM 97-031, "Block Closed CV-0825 &-CV-0878" TM 98-009, "Install Temporary Filter in Spent Fuel Pool Tilt Pit Drain" TM 96-032,"Modification of Plant Interface Connections to Supply Hook-Up Locations Needed to Install the Temporary VRS Unit" TM 97-026, "Install Bypass Line Around TC-0852" Licensee Event Reports LER 96013, DC Breaker Failure During Testing For As-Found Trip Setting LER 96005, Appendix R Enhancement Analysis - DC Panels Breaker/Fuse Coordination Issue Licensing Documents Palisades D_esjgn_lnspec_tion Report No. 50-255/97-201,.12/30/9 Palisades L..etter to USNRC dated 10/30/97, 30-Day Response to Generic Letter 97-0 Palisades Letter to USNRC dated 1/5/98, 90-Day Response to Generic Letter 97-0 Palisades Letter to USN.RC dated 6/3/93, Response to NRC Bulletin 93-0 _Palisades Letter to USNRC dated 9/11/97, Supplemental Response to NRC Bulletin 93-0 *

FSAR Table 14.22-2, Parameters Used in the Offsite Radiological Consequences Analysis of the Palisade Plant Technical Specification 6.5.2, Primary Coolant Sources Outside Containmen Drawings M74, Sheet 1, "Underground Piping, Reactor Building," Revision 12 C-155, "Reactor Refueling Cavity & sump Liner," Revision 12 M202, Sheet 1, "Chemical & Volume Control System," Revision 60 M225, Sheet 1, "High Pressure Air Operated Valves," Revision 36 M221, Sheet 2, "Spent Fuel Pool Cooling System," Revision 30 Calculations EA-A-PAL-96-003, "ECCS Evaluation in Post-RAS Recirculation Modes Using Pipe-Flo,"

Revision 1 EA-C-PAL-96-088301, "Containment Spray Pump Runout and Impact.of Low Flow Rates on Pump,", Revision 0 EA-DBD-1.05-03, "Engineering Analysis for DBD-1.05 Open Item 2," Revision 0 NRC Generic Letters 85-22, Potential for Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage 97-04, Assurance of Sufficient Net Positive Suction Head [NPSH] for Emergency Core Cooling and Containment Heat Removal Pumps 93-64, Periodic Testing and Preventative Maintenance of Molded Case Circuit Breakers NRC Information Notices 87-63, Inadequate Net Positive Suction Head in Low Pressure Safety Systems 88-74, Potentially Inadequate Performance of ECCS in PWRs During Recirculation Operation Following a LOCA

.

.

.

_ 90-07, New Information Regarding Insulation Material Performance and Debris Blockage of PWR Containment Sumps 92-71, Partial Plugging of Suppression Pool Strainers at a Foreign BWR 93-02, Debris Plugging of Emergency Core Cooling Suction Strainers 93-26, Grease Solidification Causes Molded Case Circuit Breaker Failure To Close 93-34, Potential Loss of Cooling Function Due to a Combination of Operational and Post-LOCA Debris in Containment

93-64, Periodic Testing and Preventative Maintenance of Molded Case Circuit Breakers 96-55, Inadequate Net Positive Suction Head of.Emergency Core Cooling and Containment Heat Removal P1,.1rnps Under Design Basis Accident Conditions

• 97~27, EffecCof Incorrect Strainer Pressure Drop on Available Net Positive Suction Head 33-

• NRC Bulletins 95-02, Unexpected Clogging of a Residual Heat Removal Pump Strainer While Operating in Suppression Pool Cooling Mode 96-03, Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling Water Reactors Audits and Assessments PA-96-29, "Palisades JUMA Audit," dated 11/9/96 PA-97-09, "Palisades Mechanical Maintenance Program Audit," dated 3/20/98 PA-97-010, "Palisades JUMA Audit," dated 7/18/97 PA-98-03, "Palisades Operations," dated 8/21/97 Assessment Report, Operations/Shift #3 - Compliance with Alarm Response Standards, dated 2/28/98 Assessment Report, Operations/Shift #4 - Quarterly Caution Tag Verification, dated 3/30/98 -

Assessment Report, Operations/Shift #4 - Caution Tag Administration Review, dated 3/31/98 Miscellaneous Close-out Memo for IN 90-007, 8/18/9-Specification M-136, Rev 9, 4/12/95, Furnishing and Installing Conventional Type Insulatio Industry Experience Traveler PS 32948, Potential for Fibrous Insulation Material to Cause E_CCS Sump Blockage at the Cook Nuclear Plan PRC Meeting 98*03 Minutes Dated 1/22/9 Operation Department, Self-Assessment Plan

"Hot Button" Trend Data Maintenance Organization Critical Self-Assessment Plan-Fuel Cycle 14 Maintenance "Hot Button" Report Palisades Performance Monitoring - Management Summary 1997 JUMA Audit PA-97-10 Recommendation Status 1996 JUMA Audit PA-96-29 Concerns and Recommendation Status Engineering Department Self-Assessment Completed 1997 - 1998 Condition Reports Requiring Root Cause Analysis- (June 1997 - June 1998)

Open/Closed Operability Evaluations (June 1997 - 1998)

Open/Closed Corrective action Document(June 1997 - 1998)

34