IR 05000255/1992002
| ML18058A188 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/10/1992 |
| From: | Jorgensen B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18058A187 | List: |
| References | |
| 50-255-92-02, 50-255-92-2, NUDOCS 9202180057 | |
| Download: ML18058A188 (16) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-255/92002(DRP)
Docket No. 50-255 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 License No. DPR-20 Facility Name:
Palisades Nuclear Generating Plant Inspection At:
Palisades Site, covert, MI Inspection Conducted: December 17, 1991 through January 27, 1992 Inspectors:
J. K. Heller J. R. Roton B. E. Holian Approved By: ~~hief
.
.
Reactor Projects Section 2A
-'-spection summary Inspection from December 17, 1991 through January 27. 1992 (Report No. 50-255/92002CDRPl)
Areas Inspected:
Routine unannounced safety inspection by the Resident Inspectors and the Licensing Project Manager of plant operations, immediate notifications per 10 CFR 50.72, radiological controls, maintenance, surveillance, reportable events, plant review committee and.NRC Region III request No Safety Issues Management System (SIMS) items were reviewe In addition, a routine management meeting held at the site on December 19, 1991, occurred during the perio *
Results: Of the 8 areas inspected, no violations or deviations were identified in any are Strengths, weaknesses and open items are discussed in paragraph 11,
"Management Interview."
In summary:
Strengths included the response to the cooling tower pump trip, technician knowledge of a work activity, prestaging of tools, and knowledge of local leak rate test requirement Weaknesses included non-specific exit points for off-normal procedures, an additional example of inadequate controls of a contractor, and an individual entering a contamination area without wearing the proper protective othin PDR ADOCK PDR G
DETAILS Persons Contacted Consumers Power Company
- D. P. Hoffman, Vice President, Nuclear Operations
- G. B. Slade, Plant General Manager
- R. M. Rice, Plant Operations Manager
- R. D. Orosz, Nuclear Engineering and Construction Manager
- P. M. Donnelly, Safety & Licensing Director
- K. M. Haas, Radiological Services Manager
- J. L. Hanson, Operations Superintendent
- R. B. Kasper, Maintenance Superintendent
- K. E. Osborne, System Engineering Superintendent L. J. Kenaga, Health Physics Superintendent c. S. Kozup, Technical Engineer R. J. Frigo, Operations Staff Support Supervisor w. L. Roberts, Senior Licensing Analyst R. W. Smedley, Staff Licensing Engineer
- T. A. Buczwinski, Reactor & Thermal Hydraulic Engineering Manager
- T. J. Palmisano, Administrative & Planning Manager Nuclear Regulatory Commission CNRCl
- C. J. Paperiello, Deputy Regional Administrator
- W. D. Shafer, Chief, Reactor Projects Branch 2 B. L. Jorgensen, Chief, Reactor Projects Section 2A
- J. K. Heller, Senior Resident Inspector
- J. R. Roton, Resident Inspector
- Denotes some of those present at the Management Meeting on December 19, 199 *Denotes some of those present at the Management Interview on January 31, 199 Other members of the plant staff and several members of the contract security force were also contacted during the inspection perio.
Operational Safety Verification (71707, 71710, 42700)
Routine facility operating activities were observed as conducted in the plant and from the main control roo Plant power operation and system lineup and operation were observed as applicabl The performance of reactor operators and senior reactor operators, shift engineers, arid auxiliary equipment operators was observed and evaluate Included in the review were procedure
! * * use and adherence, records and logs, communications, shift/duty turnover, and the degree of professionalism of control room activitie Evaluation, corrective action, and response for off normal conditions were examine This included compliance to any reporting requirement Observations of the control room monitors, indicators, and recorders were niade to verify the operability of emergency systems, radiation monitoring systems, and nuclear reactor protection system Reviews of surveillance, equipment condition, and tagout logs were conducte Proper return to service of selected components was verifie General The plant operated at essentially full power during this reporting period except for an approximately 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> period while the plant recovered from a cooling tower pump trip that is discussed in paragraph 2.f of this repor Alternate Steam Supply to the Turbine Driven Auxiliary Feedwater CTDAFW) Pump, CV-0521
~nspection Report No. 50-255/91024(DRP) details the initial and second failure of CV-0521 to stroke open during surveillance testin Also discussed was the evaluation conducted by the Plant Review Committee (PRC) to understand the problem and evaluate available option On December 14, 1991, CV-0521 failed to stroke for the third tim The valve was on a daily stroking schedul In accordance with a course of action previously approved by the PRC, a steam vent path was established by backing off on the packing ring to prevent a "hydraulic lock" which had been theorized as a possible proble To date, CV-0521 remains on a daily stroke schedule and has not failed to ope The lic'ensee will repair the valve during the upcoming refueling outag Currently, the TDAFW pump is continuously rolling due to leakage past CV-0521, a condition which has been analyzed previously and determined not to degrade the operational capability of the pum Control Rod Drive Mechanism CCRDM) 17 Inoperable Inspection Report No. 50-255/91018(DRP) discusses the licensee's response to an elevated seal leak-off temperature for CRDM 1 Following the reactor trip of December 9 *,
1991, the leak-off temperature of CRDM 17 returned to the normal temperature ban Based on this fact, CRDM 17 once again was declared operabl Following rod manipulations conducted while returning the reactor to power operations, CRDM 17 once again experienced a high leak-off temperature and was declared inoperabl The licensee has again
- implemented action previously discussed in Inspection Report No. 50-255/91018(DRP).
Tours (1)
During a tour of the diesel generator rooms, the inspector observed a sign glued to a lube oil heat exchanger suppor The sign listed the hazardous chemicals used and stated precautions for breaching system integrit The precautions were applicable to both diesel generators but the sign was only posted at the 1-1 diesel generato The sign was a typed 8 1/2" x 11" sheet of paper enclosed in a plastic folde The name of the sponsor was typed on the sheet and the words "Do Not Remove" were handwritten in re The inspector discussed this item with the sponsor, who stated that state law requires posting of this information when hazardous chemicals are in us The inspector questioned if the appropriate administrative controls were in plac For example, the sign had no indication that it was reviewed or approved for use, the typed sheet did not indicate that the sign consisted of only one page, and the sign did not appear permanent as indicated by the method of attachment and the fact that the sign for the 1-2 diesel generator was missin These observations were discussed with the sponsor and at the management intervie (2)
During a tour of the component copling water room, the inspector noted that some of the nut to stud engagements for check valves CC-916 and 914 - both heat exchanger outlet valves - had less then one full thread of the stud extending above the nu This was discussed with maintenance personnel who indicated that a minimum of one full thread of the stud should extend above the nu The valve numbers were provided to the mechanical maintenance department scheduler who reviewed this item and found three additional valves with questionable thread engagemen Work requests were issued to address the proble Maintenance history was reviewed to determine if these valves were the subject of recent maintenance activities; none were foun This observation appears to be administrative in nature and did not affect operabilit System Walkdowns The inspector verified operability of the systems listed below by using the applicable checklist (CL) and confirming that major flowpath valves were in their correct positio No items were found that degraded the system (1)
Control Room "Black Dot" Checklist for Power Operation **
(2)
Main Generator Seal Oil System, CL 8.2.
(3)
Steam Generator Feed Pump P-lA System, CL 1 Cooling Tower Pump Trip On January 25, 1992, one of the two cooling tower pumps tripped because of a failed water flow transmitter for cooling water to the pum Loss of the pump resulted in a reduced condensate cooling capability and a drop in condenser vacuu The crew properly diagnosed the problem and reduced power to 60 percen The licensee evaluation determined the crew responded properly and the plant responded as per desig However, the licensee did determine that the controlling off-normal procedure -
ONP 14
"Loss of Condenser Vacuum" - did not clearly identify the exit point In fact, the licensee determined that, in general, the alarm response procedures and off-normal procedures did not clearly identify the exit point This item was identified at the end of the inspection report period and was still under evaluation by the license The inspector's review will continue during the next reporting perio No violations, deviations, unresolved or open items were identifie Immediate Notification Per 10 CFR 50.72 During this reporting period the licensee made the following 10 CFR 50.72 telephone notifications to the NR The inspector performed an initial review to determine if the problem affected operability of the identified equipment, immediate corrective action was implemented and the problem was correctly identified to the NRC Operations Center duty office On January 7, 1992, while performing the Escape Airlock Test (S0-4b), the licensee discovered the inner door equalizing valve stuck in the open positio This meant that containment integrity had been violated when personnel entered the escape airlock while installing the inner door strong backs to facilitate testin Additional reviews are documented in Paragraph 6.b of this report and will be conducted when the Licensee Event Report is issue On January 20, 1992, the licensee declared an Unusual Event (UE) when Safeguards Contingency Procedures were activate This was caused by the discovery of a toy plastic hand grenade on top of the suction piping to Condensate Pump P-2 The security force determined that the device was inadvertently left behind following security drills which had been conducted the previous wee The licensee
discussed the event with the contractor who conducted the drills, confirmed that it was a prop used during the drills
- and that there were no additional unaccounted props.*
In accordance with the Palisades Site Emergency Plan, any activation of the Safeguards Contingency Procedures requires entrance into the Site Emergency Pla The inspector
reviewed this event with a Region III security specialist and had no additional question The inspector noted that control of other kinds of contractors was a previously identified weaknes During a configuration control group review of the shutdown cooling system, Temperature Indicator TE-351B, "Outlet of the Shutdown Cooling System Heat Exchangers" was identified as not qualified per design specification of the FSA The FSAR stated that certain categories of Regulatory Guide 1.97 identified equipment will be environmentally qualifie TE-351B was classified as equipment that required environmental qualificatio This instrument, which was original plant equipment, was not upgraded to the environmental standard The licensee has declared the instrument inoperable and documented a justification to permit plant operations until the next refueling outag The inspector discussed this item with a Region III Regulatory Guide 1.97 specialist and concluded that there was not an immediate operability concer No violations, deviations, unresolved or open items were identifie.
Radiological Controls (71707)
During routine tours of radiologically controlled plant facilities or areas, the inspector observed occupational radiation safety practices by the radiation protection staff and other worker Effluent releases were routinely checked, including examination of on-line recorder traces and proper operation of automatic monitoring equipmen Independent surveys were performed in various radiologically controlled area During the performance of Q0-17, "Inservice Test Procedure, Charging Pumps," the inspector observed an Auxiliary Operator (AO) enter the posted contaminated area at the P-55C charging pum The radiological work permit (RWP) controlling this activity was RWP P-92-11 The inspector noted that the AO was not dressed in accordance with this RWP in that, he was not wearing the prescribed rubber overshoe (The AO was, however, wearing shoe covers).
Shortly after entering the contaminated area, the AO realized he was not wearing the overshoes and asked the AO assisting him for a pai This was don The inspector observed the two AOs as they exited the contaminated are In each case, the AOs demonstrated appropriate techniques for
removing and disposing of their protective clothin The inspector also noted that once they had removed their rubber overshoes, they had difficulty maintaining their footing on the inclined floor. -The step-off pad could have been located on level flooring and precluded the potential for an accidental slip or fal The inspector informed both the Health Physicist on duty and the Control Room Supervisor regarding failure of the AO to dress in the required protective clothing prior to entering the posted area. It was the inspector's understanding that such deviations were reported in accordance with the guidelines provided in Health Physics (HP) Procedure 1.3, Revision 5, "Investigation of Radiological Incidents and Deficiencies."
However, no report was initiate The inspector was informed that the nature of this incident did not meet the threshold to warrant a Radiological Deficiency Repor *
The inspector discussed this issue with both the Health Physics Supervisor and Radiological Services Manager and is satisfied that not reporting this incident does not conflict with the guidance set forth in HP The Radiological Service Manager stated at the exit interview that this item was included in the plant trending progra No violations, deviations, unresolved or open items were identifie Maintenance (62703, 42700)
Maintenance activities in the plant were routinely inspected, including both corrective maintenance (repairs) and preventive maintenanc Mechanical, electrical, and instrument and control group maintenance activities were included as availabl The focus of the inspection was to ensure the maintenance activities reviewed were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with Technical Specification The following items were considered during this review: the Limiting Conditions for Operation were met while components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures, and post maintenance testing was performed as applicabl The following work order (WO) activities were inspected: WO 24105598, "Chemical Addition Tank." WO 24105871, "Calibrate K-6A Lube Oil High Temperature Switch." WO 24105229, "Thermal Margin Monitor Channel c. 11 J
- The inspector noted that the lead technician demonstrated a thorough working knowledge of the processing functions integral to the operation-of the Thermal Margin Monito Also noted was the manner in which the lead technician provided a.detailed description of these processing functions to the technician assigned to assist in the troubleshooting of the variable high power trip setpoin WO 24200096, "PPAC TGS032, DPI-1419. 11 WO 24200042, "C-lB, Gland Steam Condenser Exhauster."
WO 24105697, "PPAC, Verify calibration on the D/G K-6B Local Meters."
WO 24200324, "C-2C, Instrument Air Compressor."
WO 24200095, "Multi-004 Group, Service Water Bay Level Instruments."
For several of the wos listed above, the inspector noted that the mechanics or technicians involved in the repair or maintenance activity demonstrated a thorough understanding of the procedur The procedures were clear and concise as to the work to be performe Tools and other documentation required to complete the task appeared to be pre-staged at the job site indicating a detailed review of the work order was performed prior to starting the jo No violations, deviations, unresolved or open items were identifie.
Surveillance (61726, 42700)
The inspector reviewed Technical Specifications required surveillance testing as described below and verified that testing was performed in accordance with adequate procedure Additionally, test instrumentation was calibrated, Limiting Conditions for Operation were met, removal and restoration of the affected components were properly accomplished, and test results conformed with Technical Specific?tions and procedure requirement The results were reviewed by personnel other than the individual directing the test and deficiencies identified during the testing were properly reviewed and 1 resolved by appropriate management personne The following activities were inspected: Partial Q0-17, "Inservice Test Procedure, Charging Pumps."
The inspector observed _the breaker and valve alignment of Charging Pump P-55C in preparation for the partial performance of Q0-1 Standard Operating Procedure Number 2 was used in the alignmen The Auxiliary Operators
performing the line-up appeared to be very familiar with the system arid the procedur Good face-to-face communications were noted and independent verification of valve and breaker positioning was observe S0-4b, "Escape Air Lock Penetration Leak Test."
Paragraph 3.a, "Immediate Notification Per 10 CFR 50.72," of this report identified that during the performance of S0-4b the inner door equalizing valve stuck ope This resulted in loss of containment integrity when the outer door was ope The inspector reviewed the test, the test results, vendor manual file c-53 (SH73) "Operating and Maintenance Instructions for the Escape Airlock" and procedure CLP-M-4
"Air Lock Strong Back Installation and Personnel Lock Equalizing Valve Leak Check and Adjustment."
In addition, the inspector interviewed the crew that performed the test, the system engineer and the chairman of the corrective action review board (CARB) that initially evaluated the proble The inspector found that the crew performing S0-4b notified the system engineer when the test results indicated an containment escape air lock problem; the system engineer responded.to the plan The crew confirmed that containment integrity was not compromised by snoop testing the outer door seals and outer door equalizing valv With the outer door operable and Technical Specifications pertaining to containment integrity satisfied, the crew entered the containment via the other personnel airlock and found the inner door equalizing valve stuck ope The crew confirmed, by telephone conversation with operations management, that the performance of S0-4b provided the appropriate pre-maintenance test data to permit manual closure of the equalizing valv The crew manually closed and exercised the valve a number of times before satisfactorily completing the tes Following post test restoration the valve was visually verified close The CARB required an evaluation to determine if the valve had failed in the open position when S0-4b was performed approximately 6-months ag The equalizing valve is spring loaded and opened when the hand wheel actuates the cam and lever ar The evaluation determined that the valve was shut when the outer door was first opened since interviews with the operator entering the airlock did not indicate that he heard noise that would be characteristic of air passing by the stuck open equalizing valv In addition, the equalizing valve was operated a number of times as a result of air lock entries to establish the proper test condition The inspector compared the vendor manual to CCP-M-The vendor manual specified a test clamp bolting torque value of 320 ft-lbs, whereas the maintenance procedure specified a torque value of 260 ft-lb The lower torque value appears
- acceptable as evidenced by the satisfactory test results.
However, it was not clear why the vendor manual and maintenance procedure specified different torque value This was identified to the system engineer and discussed at the management intervie DW.0-13, "LLRT Inner /Outer Personnel Air Lock Seal."
MI-27E, "Functional Check of PCS Low Temperature overpressure Protection LTOP."
No violations, deviations, unresolved or open items were identifie.
Reportable Event {92700, 92720)
The inspector reviewed the following Licensee Event Report {LER)
by means of direct observation, discussions with licensee personnel, and review of record The review addressed compliance to reporting requirements and, as applicable, that immediate corrective action and appropriate action to prevent recurrence had been accomplishe {Open) LER 92003:
Inadvertent Auto Start of both Diesel Generators On December 15, 1991, the licensee reported the inadvertent auto-start of both emergency diesel generators when an undervoltage condition was detected by the undervoltage relays of the "lC" and
"lD" safety related busses due to the starting of a non-safety related load on the "lE" non-safety related bu During the course of a normal plant start-up with reactor power at 35 percent, the second Heater Drain Pump {P-lOB) was starte At 800 HP, the Heater Drain Pumps are the largest loads on the 2400 VAC syste The starting transient caused the voltage on the 2400 VAC system to drop below the setpoints of the second level undervoltage relays on the "lC" and "lD" busse Actuation of the undervoltage relays caused both diesel generators to start. The logic for the undervoltage relays requires three out of three actuations for the associated bu The 127-7XY relay on bus "lC" apparently did not actul:tt However, actuation of the relays on bus "lD" is sufficient to start both diesel generator Observations made by the*licensee during evaluation of this incident include:
{l)
Amperage indications for the P-lOB motor during and after the start were norma Starting current exceeded the range of the meter which has a maximum reading of -
300 amp After a few seconds the amperage indications reduced to a normal running current of 143 amp The other Heater Drain Pump, P-lOA, was running at 145 amp **
(2)
Pump packing leakoff for P-lOB was norma It had last been adjusted following the plant shutdown in July 199 (3)
Both diesel generators started and were at full operating speed and voltage by the time the control room operator walked over to the C-04 control panel to check on diesel generator respons (4)
The Control Operator #1 reported that the "2400 V BUS lC AND/OR lD UV" alarm had not been receive However, the shift Supervisor said that he might have seen the alar This alarm is actuated by the same auxiliary relay that actuates one of the diesel generator start circuit (5)
The second level undervoltage relays are set to actuate at 91.8%, 2204 VAC, on the respective bus "lC" or "lD."
The relays on bus "lC" (127-7's) were last checked on 12/12/91 and found to be within allowable toleranc The relays on bus lD (127-8's) were last checked on 10/24/90 and found to be within allowable toleranc The relays were set to actuate after a 0.5 second time dela (6)
If bus voltage had remained degraded an additional 6 seconds, then both offsite power breakers to the associated bus would have tripped and a load shed signal generate (7)
The recorded voltages were:
Safeguards bus -
2.4 KV, Bus "lC" -
2.38 KV, and Bus 1110 11
-
2.42 K This incident is similar to one in which the diesel generators started during an 1989 performance of Technical Specification surveillance Test Q0-1, "Safety Injection System."
During that test all Engineered Safeguards Systems pumps were started simultaneously which also causes a momentary voltage drop on the 2400 VAC syste Licensee actions planned or taken include:
(1)
Perform an analysis of the voltage transient associated with performance of Q0-1 and provide operators recommendations for operation of the 2400 VAC system to preclude diesel generator star (2)
Validate the computer model for the voltage transient during performance of Q0- (3)
Verify the impedance of the 2500-foot cable run between Safeguards Transformer 1-1 and the Safeguards bu It has been determined that the calculated impedance for the
safeguards transformer cable was incorrec The safeguards transformer incorporates an automatic under-load tap changer to maintain 2400 VAC on the "lC" and "lD" busse The tap changer uses a line drop compensator setting to account for the cable impedance and voltage drop between the transformer and the bu Because of the incorrect calculated cable impedance, the compensator setting was also incorrec Calculations have shown that with the incorrect compensator setting, the nominal bus voltage operates within a 30 VAC band and the actual voltage could be lower than nomina Computer modeling of the voltage transient which was caused by starting a Heater Drain Pump revealed that, at the reduced starting voltage, the voltage drop was down to the setpoint of the second level undervoltage relay The proximate cause of the inadvertent diesel generator start is a voltage transient caused by starting a Heater Drain Pump which caused the 2400 VAC system to drop to the setpoints of the bus
"lC" and "lD" second level undervoltage relay Both diesel start events resulting from voltage transients occurred after the installation of Safeguards Transformer 1-1 during the autumn of 198 The transformer was installed as part of Facility Change FC-800 to improve offsite power supplie "Palisades Safe Guard Transformer 1-1 Electrical Specification Study", RP-00-88-65, was used to determine the specifications for the transforme The study used a cable impedance value provided by Consumers Power Company Distribution Department to determine transformer specification The impedance value was not verified by the project enginee The licensee indicated that the value would not normally be verified since it was provided by "experts in their field."
The transformer specifications were also bas*ed on all expected load conditions including Loss-of-Coolant-Accident and normal operation The response of Safeguards Transformer 1-1 to a step load increase was expected to be less of a voltage drop than what was the case for Station Power Transformer 1-2, the former normal power source for busses "lC" and 1110. 11 Since voltage transients had not started the diesel generators before installation of Safeguards Transformer 1-1, it was not expected that the diesel generator starts would occur after installatio Post modification testing did not include measurement of the cable impedance or overall system dynamic response testing which might have uncovered the deficienc No problems were observed during the first performance of Q0-1 after the modification, but special instrumentation was not use The plant was shutdown
during the test, however, and the 2400 VAC system was probably not loaded as heavily as during plant operation In both voltage transients the voltage quickly recovered within the 6 second time delay before further automatic actions were initiate Therefore, there was no concern about operability of the 2400 VAC system or of the emergency diesel generator Proposed remedial corrective actions are:
(1)
Change Safeguards Transformer 1-1 line drop compensator setting per the recommendation of System Protection Engineerin (2)
Verify actuation of the 11 2400 V BUS lC AND/OR lD UV" alar Proposed actions-to prevent recurrence are:
(1)
Measure voltage and current during start of a Heater Drain Pump to validate the computer mode (2)
Evaluate Post Modification Testing requirements with respect to overall system response testing or dynamic response testin (3)
Evaluate bus 111C" and bus 111D" second level undervoltage relay setpoints and time delays to accommodate expected transients for starting the largest load or simultaneous starting of several loads during routine testin (4)
Establish a periodic activity to check the tap changer compensator settin This event is being evaluated by a Region III inspection specialist as part of the recently completed Engineering Design Safety Functional Inspection (EDSFI).
No violations, deviations, unresolved or open items were identifie.
Plant Review Committee (37702. 40500)
The inspector attended the regularly scheduled monthly Plant Review Committee (PRC) meeting on November 21, 199 The inspector observations of four items are discussed belo The PRC discussed a new procedure which would allow Technical Specification basis changes without using the formal Technical Specification amendment proces This is permitted, since the basis section of a Technical Specification is not legally part of the Technical Specification However, the inspector cautioned the plant staff to ensure that they do not inadvertently change the licensing basis while making a Technical Specification Basis chang Implementation of the change was postponed until
- the plant staff discussed the new process with the NRC Office of Nuclear Reactor Regulation (NRR).
The inspector asked one PRC member how many days in advance of the meeting he received the PRC agenda with attachment The individual stated that he had received the package the previous afternoo The inspector questioned whether one working day notice was sufficient time to review the agenda and attachmen Log No. 91-1246 provided a safety evaluation for an FSAR change which had been included in the last FSAR update pertaining to containment sump level, volume and alarm setpoint change This evaluation was necessary because there was not a written justification to support the FSAR change. NRR's review (NRC TAC number M80204) of the last FSAR change continue The inspector noted that this log number was not discussed during the PRC meeting and it was not clear if the PRC membership was aware that the evaluation was required to support a previously submitted FSAR chang Log No. 91-1210 involved a change to the Technical Specification Limiting Safety System Settings - Reactor Protective system, Basis sectio The change had been recommended for approval by the subgroup designated by the PRC to review safety evaluation However this change contained incorrect information, and was not approve The identification of incorrect material was the result of an inter-discipline PRC membership and a strong discussion of the proposed chang This was considered a PRC strengt However, this item was one of only a few which were discusse The Technical Specification basis change was an example of inadequate technical review. _Although this item was not significant, it raises a question regarding the review proces This will be the subject of an upcoming NRR 10 CFR 50.59 inspectio No violations, deviations, unresolved or open items were identifie. Region III Requests (92705)
By memorandum dated December 20, 1991, NRC Region III management requested the inspector to meet with plant management and emphasize the seriousness of two recent loss-of-power events caused by crane movement Both events were caused when the boom came in contact with over head power cable On December 20, 1991, the inspector discussed the memorandum with the Plant Manager who indicated he was aware of the events, the j
potential safety significance, and stated that additional measures were in place during crane movement to support the recently completed steam generator replacement outag The inspector also discussed this item with the outage planning staff who stated that crane movements were planned while the work orders are planne They also stated that the plant does not have. a formal administrative program for crane movemen No violations, deviations, unresolved or open items were identifie.
Quarterly Management Meeting A Quarterly Management meeting was held at the Palisades site on December 19, 1991, with the personnel indicated in Paragraph in attendanc Topics of discussion included: reactor vessel life, pressurized thermal shock, flux reduction, plant operating performance, spent fuel storage, ALARA performance, and shutdown ris.
Management Interview The inspectors met with licensee representatives - denoted in Paragraph 1 -
on January 31, 1992, to discuss the scope and findings of the inspectio In addition, the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection was also discusse The licensee did not identify any such documents or processes as proprietar Highlights of the exit interview are discussed below: Strengths noted:
(1)
Operations personnel properly responded to a cooling tower pump trip and the plant responded as per design (paragraph 2.f, "Operations - Cooling Tower Pump Trip").
(2)
A technician demonstrated a through working knowledge of a component and clearly communicated the repair activity to an assisting technician (paragraph 5.c;
"Maintenance").
(3)
There were positive indications that procedures were reviewed and that tools were prestaged (paragraph 5,
"Maintenance").
(4)
Plant personnel demonstrated a good understanding of local leak rate testing requirements and the need to perform pre-maintenance testing before performing a repair (paragraph 6.b, "Surveillance -
Escape Air Lock Penetration Leak Test"). Weaknesses noted:
(1)
The off normal procedure did not specify the exit points - this is a repeat item (paragraph 2.f,
"Operations - Cooling Tower Pump Trip").
(2)
A contractor left a prop behind following a drill indicating that sufficient controls of the contractor were not in place - similar problems were previously identified in other contractor performance areas (paragraph 3.b, "Immediate Notif~cation Per 10 CFR so. 72 11 ).
(3)
An individual entered a posted contamination area without wearing the proper clothing (paragraph 4,
"R~diological controls"). Observations pertaining to the installation/control of hazardous chemicals signs (paragraph 2.d.(1), "Operations -
Tours") * * Observations pertaining to the installation of fasteners and how many threads must extend above the nu The inspector concluded that the components were not degraded and there was insufficient information to classify this as a weakness (paragraph 2.d.(2), "Operations - Tours"). The three 10 CFR 50.72 notifications were discussed~
Two of the items will be reviewed when the LERs are issued (paragraph 3, "Immediate Notification Per 10 CFR 50.72 11 ). The CARB requested an evaluation to determine if a problem affected past operabilit This is an indication that lessons learned from a previous violation are being implemented (paragraph 6.b, "Surveillance -
Escape Air Lock Penetration Leak Test"). An example was discussed where a vendor manual and maintenance procedure were not in agreemen The example did not appear to affect the component operability
(paragraph 6.b, "Surveillance - Escape Air Lock Penetration Leak Test").
- The Region III request pertaining to loss of offsite power events caused by crane movement was discussed (paragraph 9,
"Region III Requests").