IR 05000413/1993018
| ML20056D186 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/26/1993 |
| From: | Frudenberger R, Hopkins P, Lesser M, John Zeiler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20056D184 | List: |
| References | |
| 50-413-93-18, 50-414-93-18, NUDOCS 9308050080 | |
| Download: ML20056D186 (14) | |
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- p Mc UNITED STATES f
3" ;* e NUCLEAR REGULATORY COMMISSION REGloN 11 ,
- - 99( E 101 MARIETTA STREET, N.W., SUITE 2900 i
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ATLANTA, GEORGIA 30323-0199 7n-l
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- i Report Nos.: 50-413/93-18 and 50-414/93-18 l Licensee: Duke Power Company -
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422 South Church Street Charlotte, ;
Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-52 >
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Facility Name: Catawba Nuclear Station Units 1 and 2 Inspection Conducted: June 6, 1993 - July 3, 1993 ,
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Inspector: 7/2/3}93 R. J. freudenberger, Senior Resident Inspector Date Signed i Inspecto : d) '7lt3I93 P. C. Hopkins, Resident Inspector Date Signed l
i Inspector: /] ~7 f13 l U Date Signed p J. Zeiler, Resident Inspector .
Approved byY , 7 /2c [U ,
M.fS! Lesser, Chief Date Signed :
Projects Section 3A Division of Reactor Projects
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SUMMARY !
Scope: This routine, resident inspection was conducted in the areas of !
review of plant operations, Engineered Safety Features walkdown, :
surveillance observations and maintenance observation '
Results: No violations or deviations were identified. Two inspector followup items were identified involving (1) the licensee's evaluation of fibrous air filters inside lower containment (paragraph 3.e), and (2) the licensee's evaluation of a potential problem in the Nuclear Service Water System single pump flow ]
balance (paragraph 5.b). ;
930B050080 930726 l PDR ADOCK 05000413 )
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REPORT DETAILS l
, Persons Contacted ,
Licensee Employees !
S. Bradshaw, Shift Operations Manager >
l J. Forbes. Engineering Manager !
l R. Futrell, Regulatory Compliance Manager :
1 T. Harrall, Safety Assurance Manager ;
- W. McCollum, Station Manager
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i W. Miller, Operations Superintendent t
- K. Nicholson, Compliance Specialist s D. Rehn, Catawba Site Vice-President !
Other licensee employees contacted included technicians, operators, ;
mechanics, security force members, and office personne NRC Resident Inspectors .
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- R. Freudenberger, Senior Resident Inspector P. Hopkins, Resident Inspector ;
- J. Zeiler, Resident Inspector .
- Attended exit intervie ,
Acronyms and abbreviations used throughout this report are listed in the last paragrap . Plant Status ;
i Unit 1 Summary i l
Unit 1 began the report period operating at 30 percent power while !
completing repairs to a main condenser tube leak. The unit was-returned ;
to full power on the morning of June 7. On June 12 a power reduction was initiated to commence a forced maintenance outage to repair a ruptured expansion joint in the "E" feedwater heater steam extraction ;
line. That same day, with the unit at essentially hot zero power, an i automatic reactor trip occurred due to loss of control power (a fuse was .
blown) to one of the two. Intermediate Range Detectors. The unit ;
returned to full power on June 26 and operated at or near this level for l the remainder of the reporting perio !
I Unit 2 Summary Unit 2 operated at or near full power for the entire report period with !
no major problem ,
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3. Plant Operations Review (71707) )
! General Observations
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The inspectors reviewed plant operations throughout the report l period to verify conformance with regulatory requirements, TS and ;
administrative controls. Control Room logs, the Technical Specification Action Item Log, und the R&R Log were routinely reviewed. Shift turnovers were observed to verify that they were conducted in accordance with approved procedures. The number of licensed personnel on each shift inspected either met or surpassed 1
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the requirements of Technical Specifications. Further, daily plant status meetings were routinely attende Plant tours were performed on a routine basis. The areas toured included but were not limited to the following:
Turbine Buildings Auxiliary Building .
l Units 1 and 2 Diesel Generator Rooms i Units 1 and 2 Vital Switchgear Rooms .
Units 1 and 2 Vital Battery Rooms Standby Shutdown Facility During the plant tours, the inspectors verified by observation and interviews that proper measures were taken, and procedures were followed, to ensure that physical protection of the facility met current requirements. Items inspected included the adequacy'of the security organization; the establishment and maintenance of ,
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gates, doors, and isolation zones in the prope* conditions; and the use of access control badgin In addition, the areas toured were observed for fire prevention and protection activities and radiological control practices. The inspectors also reviewed PIPS to determine if the licensee was appropriately documenting problems and implementing corrective action ' Licensee 10 CFR 50.72 Reports (1) Unit 1 Reactor Trip On June 12 the licensee made a report in accordance with 10 CFR 50.72 concerning an automatic reactor trip that occurred while the unit was being shutdown to repair the leaking expansion joint in the "E" feedwater heater steam extraction line. The trip occurred with the unit at essentially hot zero power due to an inadvertent NIS I/R High Flux trip signal. This trip signal was generated due to a loss of control power voltage, caused by a blown fuse in the I/R channel detector N-35. Since the unit was essentially shutdown when the trip occurred, only a minor transient
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resulted. All safety systems responded as expected with one exception; one train of the BDMS failed to come on scale following the tri The inspectors monitored the licensee's actions during the event recovery and reviewed the licensee's investigation of the cause of the event. The licensee was unable to determine the exact cause of the blown fuse, but, during troubleshooting, noted the detector was responding in an erratic manner. The entire N-35 drawer was therefore replaced and successfully tested prior to unit restart. The licensee determined that the cause of the BDMS problem was a malfunction of the high voltage power supply. The power supply was replaced and the channel was returned to service prior to unit restar IAE personnel indicated that further testing of the original N-35 drawer would be conducted to determine the exact cause of the blown fus (2) Unit 1 Reactor Trip Signal Generation While Shutdown On June 22, the licensee made a report in accordance with 10 CFR 50.72 concerning a Unit I reactor trip signal that was generated due to a spurious spike on a power range monitor while another power range monitor was in a trip condition for testing. The unit was in Mode 3, Hot Standby, and the licensee was conducting a monthly ACOT on power range monitor N-44. During this testing a spurious spike occurred on power range monitor N-41, resulting in the generation of a High Rate Flux trip signal. Since the reactor trip breakers were open and all of the control rods were inserted i in the reactor core, no control rod motion occurre l After the event the licensee conducted troubleshooting of the N-41 instrumentation. Continuous monitoring equipment l was installed to capture another spurious spike, but none occurred. IAE personnel determined that the cause of the spike was most likely an intermittent failure in the power supply unit to the instrumentation. This power supply was replaced and the system was successfully tested prior to the unit entering Mode 2 on June 24. The inspectors reviewed the troubleshooting and determined that it was comprehensive, and that each of the potential problems had been thoroughly examine ,
l On July 1 the licensee made a call to the NRC to retract the l 10 CFR 50.72 report that was made on June 22. The ;
retraction stated that after further evaluation of the event, the licensee determined that it did not meet the reporting criteria in 10 CFR 50.72 or 50.73. This determination was based on the plant's shutdown config- ,
uration with the control rods inserted and the reactor trip 1 breakers open at the time that the spurious trip signal was l
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i generated. The inspectors reviewed the guidance in NUREG-l 1022 for reporting criteria and determined that, since this spurious actuation occurred when the rod control system was not required to be operable and the system had been properly removed from service, the event was not reportable, c. Shutdown to Repair Unit 1 "E" Feedwater Heater Extraction Line On June 12 the licensee commenced a shutdown of Unit I to repair a ruptured expansion joint in the "E" feedwater heater steam l extraction line. The expansion joint was located inside the "C"
- main condenser, requiring that the secondary side condenser vacuum be broken to allow personnel to enter the condenser and perform the repairs. The licensee engineering staff evaluated the necessary repair work and determined that it could be accomplished in Mode 4, Hot Standby, with no decrease in the level of safety to the plant or to personnel. Once in Mode 4 the NC system temperature and pressure were reduced and maintained at 250 *F and 350 psig, respectively, during the outage wor Prior to shutdown the inspectors discussed the evolution with the engineering staff and reviewed their evaluation associated with
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existed. The inspectors determined that the licensee had adequately evaluated the potential safety concerns associated with remaining in Mode 4 for the duration of the outag The inspectors followed the outage activities and attended many of the outage scheduling meetings. There was good pre-planning of activities, resulting in the completion of almost all outage
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related work items as planned. Likewise, outage activities were l well coordinated and properly executed with no major discre-pancies. The inspectors also noted that the operations department maintained effective control of plant conditions during this abnormal configuratio d. Review of Salem Rod Control System Failure On May 27, 1993, Salem Nuclear Generating Station, Unit 2, experienced a rod control system failure which resulted in the withdrawal of a RCCA 15 steps from the core while the operator was applying a rod insertion signal. The cause of the failure was l subsequently determined by Westinghouse possibly to have been due l to a single failure in the rod control system, which potentially l affects all Westinghouse-designed plants. The current licensing basis for plants with this system, including Catawba, assumes that only multiple failures could cause the withdrawal of a single RCC On June 10, 1993, the HRC issued Information Notice 93-46, l informing all Westinghouse-designed plants of the potential i problem with the rod control system. The following day,
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Westinghouse distributed a Nuclear Safety Advisory Letter (N NSAL-93-007) to all applicable plants and provided details of i their investigation to determine (1) the impact on rod control system functionality, (2) current regulatory requirements, (3)
failure detectability, and (4) failure scenarios. Based on this preliminary investigation, Westinghouse determined that a i substantial safety hazard may not exist, but recommended several actions that plants should take until their investigation was completed. These actions included: (1) continuing with the normal .
process of verifying that rod motion is proper for requested rod '
movement, (2) confirming the functionality of rod deviation alarms, and (3) requiring operators to review the Advisory Letter to ensure they understand the Salem even .
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The inspectors verified that the licensee had received both the NRC Information Notice and Westinghouse Advisory Letter and that actions were being taken to evaluate the potential impact of a t similar problem occurring at Catawba. When this issue arose Unit I was in Mode 4 for repairs of a leaking expansion joint in the ;
"E" feedwater heater steam extraction line. Prior to unit startup l the inspectors verified that the Westinghouse recommendations ,
described above had been completed. The inspectors witnessed
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several of the special briefings that were given to all of the :
operators alerting them to the problem. At these briefings, the operators were provided with details of the Salem event and were ,
required to review the NRC Information Notice and Westinghouse Advisory Letter. Engineering personnel conducting these briefings were knowledgeable of the event and were able to adequately answer operator questions on the issue. The inspectors noted good licensee management involvement in this process as evidenced by their attendance at these briefings. On June 24, following condenser repairs, unit startup commenced. No rod control system problems were encountere On June 21 the NRC issued GL 93-04, Rod Control System failure and '
Withdrawal of Rod Control Cluster Assemblies. This GL required the licensees of all Westinghouse-designed plants to provide the NRC with information describing their findings related to this issue and actions taken. At the end of the reporting period the licensee was continuing to assess this issue to respond to the G The inspectors determined that, to date, the corrective actions >
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initiated by the licensee were adequat e. Review of Licensee's Response to NRC Bulletin 93-02 The inspectors reviewed the licensee's evaluation of NRC Bulletin 93-02, Debris Plugging of Emergency Core Cooling Suction Strainers. This Bulletin notified the licensee of a concern involving the potential for degradation of the ECCS post-LOCA recirculation capability due to blockage of the containment sump screens from fibrous materials. The licensee's response to this
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Bulletin was transmitted to the NRC via a letter dated June 10, 199 The inspectors determined that the licensee had adequately identified the sources of fibrous material inside containmen The licensee determined that three ventilation systems inside containment contained fibrous air filter material. Of these three, the inspector considered that the fibrous air filters from I the CACFUs would have the greatest potential to migrate to the j ECCS sump. Each unit has two CACFUs located inside the polar i crane wall of lower containment. Each CACFU has 12 air filters, but they are not contained within the filter housing structur Two retaining clips hold each filter to the outside panel of the filter housing. Based on the torturous flow path to the sump and l low calculated flow velocity during a LOCA, the licensee concluded J that even if these filters were dislodged the likelihood that they ;
would migrate to the sump was remot In addition, the licensee !
determined that, assuming all the filter material did reach the !
sump screens, the amount of screen blockage due to the filters l would not adversely affect the ECCS performanc J l
On June 21, during an inspection in Unit I lower containment, the j inspectors examined the air filters associated with the CACFU l The inspectors observed that several of the retaining clips used I to hold the air filters to the CACFU housing were broken off, making it easier for the filters to be dislodged. This problem i was brought to the licensee's attention; the licensee reported that the clip repairs would be scheduled for the next reactor i trip. The inspectors considered this to be adequat I The inspectors also discussed with engineering personnel whether or not the air filters to the CACFUs were necessary to the i performance of the system and the possibility that they c.,uld be removed. At the end of the report period the licensee was evaluating this as a potential long-term corrective actio Pending completion of this review and the implementation of long- t term corrective action, this issue will be carried as an IFI 413, 414/93-18-01: Review Licensee Evaluation of Removal of CACFU Air Filter ,
No violations or deviuions were identifie ,
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4. Engineered Safety Features System Walkdown (71710) l'
l The inspector performed a walkdown of the Unit 1 NW System. The purpose of this walkdown was to independently verify that the system was operable as required by olant TS Using the licensee's NW system valve checklist in procedure OP/1/A/6200/19, Containment Valve Injection Water System, the inspectors verified that main system flow path valves and assorted drain and vent valves were in their proper positions. When this walkdown was conducted
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Unit I was in Mode 4 for repair of the leaking expansion joint in the steam extraction line; therefore, the inspector also was able to verify the position of selected manual NW system valves inside containmen Valves were found to be in their correct positions with no bent stems or ;
improper labelin Selected process instrumentation was examined to ensure proper installation and functioning. Hangers and supports were properly installed. Outstanding work requests on NW system components were reviewed to determine that no maintenance that affected the system's ,
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performance was outstanding. In addition, portions of the as-built system configuration were compared against the most current plant drawings to ensure that they reflected the correct configuratio The following observations were generated from the walkdown, review of '
TS, and review of the Updated FSAR for the NW system:
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While all valves were in their correct position, four globe valves _
located inside containment did not have valve handles. These valves were INW-51, -52, -55, and -174. The licensee's checklist required that INW-51 be open and the other three valves be close The inspector was unable to determine the position of these valves. However, when the licensee was notified of the missing handles an operator verified the valve positions. All four valves ,
were correctly positioned. The licensee initiated a work requcst to replace the valve handle Level in the "B" train NW Surge Chamber indicated 55 gallon Normal volume in the chamber is 75 gal'ons. This was still above the minimum level of 32 gallons required for operability. The 1 operators indicated that they were having to makeup inventory to the chamber in some cases as muc!i as once per 12-hour shif Further discussions with the operations sup:,ct * personnel revealed i
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that some troubleshooting had already beer ccncucted to determine which valves in the system were leaking. Tha raspected leaking valves were in the RN makeup line to the s arge chamber. The inspector reviewed the Unit 1 Forced Outaos list and noted that a work item had been added to repair these 54F,es. The inspector considered the licensee's actions to resols this problem to be adequat In summary, the inspector found the NW system to be fully operable and
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adequately maintaine No violations or deviations were identifie . Surveillance Observation (61726)
During the inspection period the inspectors verified that plant operations were in compliance with various TS requirements, such as those for reactivity control systems, reactor coolant systems, safety injection systems, emergency safeguards systems, emergency power
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systems, containment, and other important plant support systems. The inspectors verified that surveillance testing was performed in accordance with approved written procedures; test instrumentation was l calibrated; limiting conditions for operation were met; appropriate
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removal and restoration of the affected equipment were accomplished; ;
l test results met acceptance criteria and were reviewed by personnel other than the individual directing the test; and any deficiencies
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identified during the testing were properly reviewed and resolved by appropriate management personne ,
t The following surveillances were witnessed and/or reviewed by the
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I inspectors: Unit 1 Pressure Boundary Valve Testing On June 21, prior to the Unit entering Mode 3, Hot Standby, the l licensee began pressure boundary valve testing in accordance with procedure PT/2/A/4200/0lN, Reactor Coolant System Pressure Boundary Valve Leak Rate Test. The purpose of this testing is to verify that leakage from any NC system pressure boundary valve does not exceed the TS limit of I gpm. Testing is performed by pressurizing across the valves and then measuring any leakage using a graduated cylinder. The inspectors witnessed portions of !
the valve testing and reviewed the results. Testing was performed .
at an NC pressure of opproximately 1300 psig and the licensee-appropriately adjusted the measured valve flow rates to full reactor coolant system pressure. The inspector noted that, although none of the valves tested exceeded their limit, several check valves were extremely close (e.g. 0.9916 gpm was obtained for valve INI-70, and 0.9924 gpm for INI-171). The inspector '
questioned the test engineer as to whether or not these values should have been rounded to 1.0. The engineer determined that, based upon the degree of accuracy required in the calculated flow rate, the measured values were correct to three decimal place If the measured values were rounded to three figures, the TS limits would still be met. The engineer indicated that +hese valves did not have a history of excessive leakage and might be worked on during the upcoming outage to reduce the leakag Following plant startup, the inspectors reviewed the calculated NC ;
system unidentified leaPage and noted that it was well within the .
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TS limit of I gpm. This provided further evidence that the '
pressure boundary valves were not leaking excessivel "A" Train Nuclear Service Water (RN) Pump Testing On June 21 the licensee attempted to test both the 1A and 2A RN .
pumps in accordance with inservice test procedure PT/0/A/4400/22A, '
Nuclear Service Water Pump Train A Performance Test. The pumps ,
are tested individually with an alignment through various Train A heat exchangers to attain a flow rate between 19,264 and 19,654 gp Pump operating parameters are measured and recorded when ,
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i 9 f this flow rate is attained, and they must meet acceptable limits '
for pump operability. When tested, neither pump could achieve the minimum flow rate described above. At the flow rate that was !
achieved, pump discharge pressure was higher than that specified in the procedure indicating that the system had developed some additional friction losse :
The inspectors compared the RN pump data obtained at the lower flow condition with the RN pump head curves. Based on this review the inspectors determined that pump performance had not degraded, >
indicating that the system flow resistance had increased and/or system leakage had increase The following day a flow balance test was performed. This was ,
done to resolve concerns that the system flow resistance may have ;
increased since the last flow balance, possibly rendering one A l train pump incapable of providing sufficient flow to plant ,
equipment during certain design basis events. The RN train is normally set up for a single pump flow balance, meaning that a single pump is capable of providing cooling to shut down one unit in a LOCA condition while maintaining the other unit in Mode 5, '
cold shutdown. The results of this flow test proved that a single
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pump was incapable of meeting the minimum flow requirement This, however, does not necessarily render the train inoperable '
because two pumps are available in the train. The RN train was tested with two pumps operating. Portions of this test were witnessed by the inspectors and the results were reviewed. The test results proved that the two operating RN pumps could provide the necessary design basis flow requirements. This condition is acceptable by T While conducting the two-pump test, the licensee identified that valve leakage through the clam backwash line was the major cause of the low pump flows. The licensee planned to make repairs to this backflush valve during the week of July 19, 1993. In the interim, procedures controlling the RN system configuration were revised and an operations Technical Memorandum was issued to :
ensure additional operator actions are performed if either unit enters Mode 5 with a Diesel Generator or RN pump inoperable. The purpose of these actions was to ensure that adequate flow via one RN pump could maintain the shutdown unit in Mode 5 and provide the i required accident flow to the safety systems of the operating unit. The operator actions necessary to accomplish this involved isolating flow to those components rendered inoperable by removing '
the Diesel Generator or RN pump from service. The inspector verified the implementation of these procedures and the Technical Memorandum. The inspectors determined that the licensee's corrective actions for this RN flow problem were adequat .
During review of the single pump flow balance procedure, the l inspectors noted that the licensee was not accounting for the '
potential flow reduction when the opposite RN pump's backwash
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strainer was operating. These strainers are designed to backwash to prevent them from clogging or restricting RN flow. Based on discussions with the engineering staff on the operation of the strainers, it was reported that the backwash cycle is automatically initiated on high differential pressure as well as cycled periodically regardless of differential pressure. The inspectors questioned the licensee as to whether or not the :
reduction in RN flow resulting from operation of an additional backwash strainer could adversely affect the single RN pump flow l balance requirements. At the end of the report period the !
licensee was evaluating this issue. Pending completion of the evaluation this issue will be carried as an IFI 413,414/93-1B-02:
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Review Licensee Evaluation of Potential RN Flow Balance Proble ,
c. Residual Heat Removal Pump Testing i
On June 30 the inspector witnessed the licensee's testing of the '
IB ND pump in accordance with procedure PT/1/A/4200/10B, ND Pump 1B Performance Test. The purpose of this test is to verify pump operability in accordance with the requirements of ASME Boiler and :
Pressure Vessel Code,Section XI. The inspector verified that 1 (1) instrumentation was calibrated, (2) all required data were taken in accordance with the test procedure, and (3) test results were satisfactory. The inspector observed good communication and coordination of activities between the test and operations '
personnel. The inspector observed that test personnel had to access several different areas of the auxiliary building to manipulate valves, check instrumentation, etc. Following their exit from these non-contaminated areas, they still used a hand and foot monitor to check for potential contamination. The inspector i considered this a good radiation protection practice to prevent ;
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the spread of potential contamination. No discrepancies were identifie i d. Diesel Generator Operability Testing On July I the inspectors witnessed testing of the 2B Diesel Generator in accordance with procedure OP/2/A/4350/02B. The insp:ctors monitored the diesel operation from the diesel generator room and reviewed the completed test procedur No +
discrepancies were note l The inspectors noted good coordination of test activities by the work scheduling group because two other TS surveillances were performed in conjunction with this diesel test. One of these surveillances involved the inservice valve stroke time test of the diesel's fuel oil day tank fill valve, 2FD-2 Following this valve stroke test the diesel is required to be operated to verify proper valve operation. The second surveillance involved testing ;
of the ESFAS automatic actuation logic and actuation relays for safety injectio It was also necessary to operate the diesel for
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this test. By coordinating these activities together, diesel generator unavailability time was reduce No violations or deviations were identifie . Ma...tenance Observations (62703)
During the reporting period the inspectors reviewed maintenance activities to assure compliance with the appropriate procedures and T Methods used in this inspection included direct observation, personnel interviews, and retnrds review. The activities associated with the following W0s were reviewed by the inspectors: WO 93044227-01: Investigate / Repair Valve 1NI-393 Not Opening On June 21, during Unit 1 NC pressure boundary valve testing, NI test header valve 1NI-393, which is used to test NC pressure boundary check valve 1NI-181, would not open to allow pressurization across the check valve. During troubleshooting inside containment, the licensee discovered the top to the air ,
filter regulator for valve 1NI-393 had failed and was blown of The failed cap was made of plastic and the replacement cap was metal, The inspectors witnessed the replacement of the air 4 regulator top and reviewed the WO documentation of the work. No discrepancies were identifie Since the inspectors recalled similar failures of air regulators ,
that had occurred the previou year, this problem was discussed with the engineering staff. In addition, PIP 2-C92-135, which documented this generic issue, was reviewed. The licensee determined that the failure of these regulator tops was due to heat and radiation induced hardening of the material. The inspector noted that operations personnel had previously identified all critical valves that used this type of regulator, and the regulators to these valves had been replaced. The licensee had planned to replace the same regulators for non-critical valves such as INI-181 in conjunction with any future maintenance on the valve The inspectors determined that the licensee's corrective actions for this issue were adequat WO 93045842-01: Investigate / Repair Check Valve 2RN-9 On June 27 a non-licensed operator discovered that the idle 2A RN pump was rotating backwards. The pump was declared inoperable and the back-flow was stopped after the discharge of the pump was isolated with a manual isolation valve. This back-flow indicated that swing check valve 2RN-9, located in the pump discharge, was not seating properly. When maintenance personnel removed the valve bonnet they discovered the disc cocked in the open position due to excessive wear at the hinge-to-disc assembly. Thr. disc ,
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assembly was replaced and the hinge was repaired. Following these repairs, the pump was successfully tested to verify operabilit The inspector witnessed portions of this maintenance, reviewed all documentation associated with the repair work, and concluded that it was properly performe No violations or deviations were identifie . Exit Interview The inspection scope and findings were summarized on July 6,1993, with '
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those persons indicated in paragraph 1. The inspector described the areas inspected and discussed in detail the inspection findings listed ;
in this section. No dissenting comments were received from the i licensee. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this l inspectio ;
Item Number Description and Reference IFI 413, 414/93-18-01 Review Licensee Evaluation of Removal of CACFU Air Filters (paragraph 3.e).
IFI 413, 414/93-18-02 Review Licensee Evaluation of Potential RN Flow Balance Problem (paragraph 5.b).
8. Acronyms and Abbreviations ACOT - Analog Channel Operational Test ASME - American Society of Mechanical Engineers i BDMS - Baron Dilution Mitigation System i l CACFU - Containment Auxiliary Carbon Filter Unit 1 CFR -
Code of Federal Regulation ECCS - Emergency Core Cooling System ESF -
Engineered Safety Features
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ESFAS - Engineered Safety Feature Actuation System l 'F -
degrees Fahrenheit i
FSAR - Final Safety Analysis Report l GL -
Generic Letter gpm -
gallons per minute l
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IAE -
Instrumentation and Electrical IFI -
Inspector Followup Item I/R -
Intermediate Range LER -
Licensee Event Report LOCA - Loss of Coolant Accident NC -
Reactor Coolant ND -
Residual Heat Removal NI Safety Injection
NIS -
Nuclear Instrumentation System NRC -
Nuclear Regulatory Commission NUREG - Nuclear Regulatory Guide NW -
Containment Penetration Valve Injection Water System
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OMP -
Operations Management Procedure OP- -
Operating Procedure PIP -
Problem Investigation Process (report) !
psig - pounds per square inch gauge ,
PT -
Periodic Procedure i RCCA - Rod Cluster Control Assembly RN -
Nuclear Service Water &
R&R -
Removal and Restoration TS -
Technical Specifications .;
WO -
Work Request ;
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