IR 05000354/1998080
| ML20249A218 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/10/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20249A198 | List: |
| References | |
| 50-354-98-80, NUDOCS 9806160189 | |
| Download: ML20249A218 (46) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50-354 License Nos:
NPF-57 Report No.
50-354/98-80 Licensee:
Public Service Electric and Gas Company Facility:
Hope Creek Nuclear Generating Station Location:
P.O. Box 236 Hancocks Bridge, New Jersey 08038 Dates:
February 23 - March 19,1998 April 27 - April 30,1998 Inspectors:
A. L. Della Greca, Team Leader D. A. Dt.mpsey, Reactor Engineer G. W. Mori's, Reactor Engineer D. T. Moy, n'eactor Engineer J. G. Schoppy. Sr. Resident inspector - Oyster Creek R. Quirk, NRC Contractor Approved by:
William H. Ruland, Chief Electrical Engineering Branch
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Division of Reactor Safety I
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EXECUTIVE SUMMARY Hope Creek Nuclear Generating Station NRC Inspection Report 50-354/98-80 During the periods of February 23,1998 to March 19,1998, and April 27-30,1998, the NRC conducted an engineering team inspection at the Hope Creek Nuclear Generating
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Station. The overall objective of the inspection was to assess the effectiveness of the engineering functions in providing for the safe operation of the plant. The assessment consisted of a safety system engineering inspection that focused on the Class 1E battery j
system and the heating and ventilation of selected safety related areas. The assessment j
also addressed the quality of the safety evaluation program and implementation and the -
effectiveness of the licensee's controls in identifying, resolving, and preventing problems.
'The team conclusions are identified below.
Maintenance i
e The licensee's failure, in 1991 and 1996, to properly evaluate the results of the
- battery capacity performance tests, resulted in two examples of a violation of the NRC test control requirements. in the first case, the licensee failed to correct the
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.tes,t results for the actual test temperature.-In the second case, they failed to,
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calculate the average capacity of previous tests to determine if the test frequency-should be increased. In another example, the licensee failed to test periodically the
- automatic features of the control equipment room supply and the control area
. battery exhaust systems, as described in the FSAR. (Sections M1.1 and M1.2)
e The material condition of the selected HVAC systems reflected an effective implementation of HC preventive maintenance program. (Section M1.2)
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- The licensee's actions to address NRC information notice 97-53 were inadequate in that they failed to evaluate whether the racked-out de breakers represented a seismic hazard to other equipment in the area, as they had previously done for the (-
480Vac breakers, and failed to evaluate the seismic capability of the ac and de switchgear with breakers in other then the full-in position. Also, they failed to
adequately address the cause of the repetitive failures of the battery charger
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overvoltage shutdown (Sections M2.1 and M3.1)
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For'the battery chargers and the battery room in-duct heater controllers, the l
licensee did not follow the manufacturer's recommended preventive maintenance j
program and had not developed an alternate program. The preventive maintenance l
l that is required to maintain operability for this equipment is a c.oncern and will be
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evaluated further during a future NRC inspection. (Section M1.2)
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Engineering The licensee had not clearly designated the calculation of record, as in the setpoint e
calculation for a temperature switch and in the voltage drop calculation for motor-
-m operated valves, and had not ensured that the results of a calculation were reflected into another calculation. (Section E1.2)
e in the design control area, the selection of thermal overload devices for the HPCI and RCIC loads powered from the dc motor control center was inadequate in that the associated calculation did not properly account for the operating temperature.
Similarly, the calculation for the 20 kVA safety-related inverter full load current neglected to consider the minimum dc voltage and the efficiency of the equipment.
(Section E1.1)
e Also in the design control ares, the justification provided in the UFSAR change notice to reduce the minimum capacity margin from 5% to C% was inadequate in that it assumed that the margin was consumable and did not take into consideration.
the less than optimum operating conditions of the battery, as permitted by the
- technical specifications and by the plant surveillance procedures. In addition, for
- the safety-related panel room and the control room chilled water systems, the water temperature was not being maintained within the limits specified in the HVAC -.
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design calculations and in the UFSAR. (Section E2.3 and E2.5)
,e-In general, the safety evaluation process was acceptable and the procedure
>. contained adequate guidance for proper program implementation. The lesson plan
,n was also acceptable. For more recent changes, the safety evaluations were better-written and adequately addressed the 10 CFR 50.59 requirements.'(Sections E2.1 and E2.2)
=e For modifications that received only a safety evaluation applicability review (AR),
the documentation was generally sufficiently detailed to permit an evaluation of the adequacy of the conclusion, but some weaknesses remained. For instance, in relocating the temperature sensors m the HPCI and RCIC rooms, the licensee judged that a safety evaluation was not required and failed to evaluate whether the change invalidated TS-specified setpoints for system isolation. (Section E2.2)
e No significant concerns were identified with safety evaluations for temporary facility modifications. However, the installation of temperature and humidity instruments in the control room and in the remote shutdown panel room was not included in the design modification process and did not receive a safety evaluation to assure that it did not adversely impact the safety functions of nearby safety-related equipment.
(Section E2.2)
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The actions to address a licensee-identified deficiency regarding the potential loss of both HVAC trains in the event of a loss of control air, indicated them to be
-. acceptable and,to restore the design to its intended performance basis.
(Section E2.4)
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The HVAC systems reviewed were capable of performing their safety functions.
e Also, the applicable operating procedures generally reflected design and operating limits that were consistent with the UFSAR and design documents. (Sections E2.5 and E2.6)
The design calculations for the safety-related HVAC systems generally were
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acceptable,' although some inconsistencies were identified. For instance, a calculation for safety and turbine auxiliaries cooling system required flows and heat i
l loads, recently performed by the licensee, used more refined and better justified methods than the original calculation. However, some incorrect design
assumptions, also identified by the licensee, resulted in a non-cited violation.
(Section E2.5)
The design of several HVAC subsystems are subject to single failure due to e
inadvertent closure of fire dampers. (Section E2.7)
Plant staff maintained good material condition and housekeeping in the auxiliary and e
reactor buildings. Personnel demonstrated a willingness to initiate corrective
- actions for identified deficiencies and documented deficiencies at a low threshold of significance and in a timely manner. Management encouraged and supported
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. employee identification of deficiencies. (Section E7.1)
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The engineering staff did not always identify non-compliances with procedural and e
regulatory requirements. For instance, Design Engineering did not adequately review the potential effect of an RHR cross-tie modification on the 'D' RHR room cooler rated flow. The 18-inch pipe added during this modification was installed in close proximity of the room ~ cooler inlet flow area, blocking approximately 1/4 of this area. This was another example of inadequate design control. (Section E7.1)
e Although the overall problem resolution was generally acceptable, the plant staff did not always adequately implement the program requirement..- For instance, they failed to address the reagent gas pressure requirements in the hydrogen / oxygen analyzer and to ensure that the reagent remained above the required pressure, in j
two other examples, they failed to evaluate the ability of potentially degraded Struthers-Dunn relays to perform their safety functions during and following a seismic event, when the degradation mode was first discovered, and failed to evaluate the ability of Agastat relays in harsh environment to perform their safety functions, following an accident, using approved industry methods. (Sections E7.2, E8.2 and E8.3)
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Contrary to the fire protection program requirements, fire protection technicians failed to fully evaluate and report to the shift supervisor five failed batteries in emergency lighting units (ELUs). Subsequently, licensee QA personnel identified approximately 43% of inoperable ELUs. (Section E7.2)
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. Failure to maintain all the service water intake structure watertight perimeter flood doors under administrative controls, during a rising tide level, resulted in a violation of the technical specifications requirements for those doors. (Section E7.2)
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. The licensee maintained the corrective action program procedures current and continued to improve the processes. Sample revicws showed the instructions to be clear and comprehensive. Personnel contacted during the inspection were knowledgeable of the corrective action process and indicated that they would not u hesitate to use it to document and correct adverse conditions. (Section E7.3)
e Indicators developed by the corrective action group (CAG) showed acceptable performance in the average time to complete level 1 and level 2 evaluations. The CAG identified the average age of open condition resolution reports as an area needing improvement. (Section E7.3)
e Quality Assessment (QA) audits of the corrective action process were typically thorough and broad-scope and documented several areas where personnel inconsistently implemented the program requirements. QA had good access to management and provided them with meaningful insights into station performance.
QA was actively involved in plant activities. (Sections E7.3 and E7.5)
e Inspection activities indicated acceptable oversight by the Nuclear Review Board, the Station Operations Review Committee, and the QA Onsite Independent
. Reviewers. (Section E7.6)
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.Self-assessment reports from operations and maintenance indicated value-added
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evaluations of their respective activities with appropriate actions for identified weaknesses. (Section E7.5)
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.The Operating Experience Group, in the most part, acceptably evaluated and
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- dispositioned incoming industry operating experience information. Albeit, some examples were found when the evaluation of the issues were insufficiently detailed.
Screening of incoming items to the Business Process in the Action Request Process potentially contributed to a less timely review of some issues. (Section E7.4)
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TABLE OF CONTENTS PAGE EX EC UTIVE S U M M ARY.............................................. ii ll. M aint e na nc e................................................... 1 M1 Conduct of Maintenance........................................ 1 M1.1 Battery Performance Capacity Tests
..........................1 M1.2 Ventilation System Testing................................. 2
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M2 Maintenance and Material Condition of Facilities and Equipment............ 4 M2.1 Walkdown of DC System.................................. 4 i
M3 Maintenance Procedures and Documentation.........................5 M3,1 Manufacturer's Recommended Maintenance.........
.......... 5 111. En gi n e e rin g.................................................... 6 E1 Conduct of Engineering.......................................... 6 E1.1 Selection and Contro8 of DC Protective Devices................... 6 E1.2 Calculation Control....................................... 7 E2
. Engineering Support of Facilities and Equipment
.......................8 E2.1 Procedures and Training for Evaluating Changes to the Facility........ 8 i
E2.2 Temporary and Permanent Modifications
.......................9 E2.3 Battery Sizing Calculation Revision........................... 12 E2.4. Control Room Chiller SACS Control Valves..................... 13 E2.5 HVAC Design Calculations and System Configuration............. 14 E2.6 HVAC System Procedures................................. 17 E2.7 Ventilation System Single Failure............................ 18 i
E7 Quality Assurance in Engineering Activities.......................... 20 E7.1 Effectiveness of Licensee Controls in Identifying Problems.......... 20 E7.2 Effectiveness of Licensee Controls in Resolving Problems........... 22 E7.3 Effectiveness of Licensee Controls in Preventing Problems
.........26 E7.4 Operating Experience Feedback............................. 27 E7.5 Self-Assessment.......................................29
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E7.6 Safety Review Committee Activities
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E8 Miscellaneous Engineering issues (92903)'.......................... 31 E8.1 -(Closed) Violation 50-354/97-07-04: Failure to Perform a Safety
- Evaluation per 10 CFR 50.59 for a Design Modification............ 31 E8.2L(Closed) Unresolved item 50-354/97-07-05: Service Life of Mild Environment Struthers-Dunn Relays.......................... 31 E8.3 (Closed) Violation 50-354/97-07-008: Service Life of Agastat Relays in Harsh Environment.................................... 3 2 E8.4 Review of Related UFSAR Sections.......................... 33 V. Managem e nt Meeting s........................................... 34 X1 Exit Meeting Summary........................................ 34 f
PARTI AL LIST OF PERSONS CONTACTED............................... 35 ITEMS OPENED, CLOSED, AND DISCUSSED.............................. 36 l
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" LIST O F 'AC RO N YM S U SED.......................................... 3 8
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Report Details During the periods of February 23,1998, to March 19,1998, and April 27-30,1998, the NRC conducted an engineering team inspection at the Hope Creek Nuclear Generating
. Station. The overall objective of the inspection was to assess the effectiveness of the engineering functions in providing for the safe operation of the plant. The assessment consisted of a safety system engineering inspection that focused on the Class 1E battery system and the heating and ventilation of selected safety related areas. The assessment also addressed the quality of the safety evaluation program and implementation and the effectiveness of the licensee's controls in identifying, resolving, and preventing problems.
During the inspection periods, the plant remained at or near full rower.
II. Maintenance M1 Conduct of Maintenance M 1.1 Batterv Performance Capacity Tests a.
Scooe of insoection (IP 93809)
The team reviewed the licensee's implementation of the battery surveillance-
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procedures to assess the capacity of the safety-related batteries and to verify their-compliance with the technical specification (TS) requirements, b.
Observations and Findinas The battery performance test requirements are specified in the TS. The team determined that the licensee conducted battery performance tests in accordance with their surveillance procedure and the technical specification requirements. The tests were also in conformance with the recommendations of IEEE Standard 450-1975, " Recommended Practice for Testing Large Lead Acid Stationary Batteries."
The team identified no concerns with PSE&G's implementation of the procedural requirements for the capacity performance tests. These tests indicated capacities of 100% or higher. The team found two deficiencies during their review of the surveillance tests performed since 1991.
In November 1991, the performance capacity test for battery 1DD447 was interrupted because of a loss of the load bank. The team found that the licensee had failed to account for this interruption and the consequent apparent capacity increase. This resulted in the licensee's incorrectly recording the battery capacity as 134.7%. Furthermore, although erroneous, this value should have been corrected for the actual test temperature, as required by the test procedure. This adjustment would have yielded an apparent battery capacity of approximately 138%. The licensee's failure to evaluate the test results and correct them is an example of inadequate test control and a violation of 10 CFR 50, Appendix B, Criterion XI, Test Control. (VIO 50-354/98-80-01)
In December 1995, when the licensee conducted the capacity test for the same battery (1DD447), the licensee found that the capacity had dropped from the 134.7% recorded in 1991 to 113.1 %. In accordance with the guidance of IEEE
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Standard 450-1975,a decrease in battery capacity of more than 10% from the AVERAGE of the previous tests should result in the licensee increasing the test afrequency from five years to 18 months. A review of the December.1995 test
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results showed that the licensee had not calculated the average capacity from the previous tests'and, hence, not determined whether the test frequency should be
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increased. For the other batteries the average capacity similarly had not been calculated. However,'only in the case of battery 1CD447, the capacity had dropped more that 10% from the previous test.
In response to the team's. concern, the licensee calculated the average of the previous two performance tests conducted during the operational phase (1991 and 1985) plus the results of the pre-operation acceptance test, also performed in 1985, and found the average of previous tests to be 120.4%. This calculation brought the change in capacity from the average of all previous tests down to less than the i
10% criteria. Therefore, an increase in the test frequency was not required.
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Nonetheless, the licensee's failure to identify this test requirement is another
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example of inadequate test control and a violation of 10 CFR 50, Appendix B, Criterion XI. (VIO 50-354/98-80-01)
Present industry standards, recognizing the need for improved monitoring of battery
degradation, require that the 10% decrease in battery capacity be based on the
previous capacity test rather than on the average of all previous tests. The licensee i
had not updated their procedure to conform with the revised standard. The team -
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considered this a weakness of the Hope Creek test procedure.
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Conclusions The licensee's failure, in 1991 and 1996, to properly evaluate the results of the
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battery capacity performance tests, resulted in two examples of a v' Ation of the NRC test control requirements. In the first case, the licenses failed to correct the.
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' test results for the actual test temperature, in the second case, they failed to calculate the average capacity of previous tests and to determine if the test frequency should be increased.
M1.2 Ventilation System Testina a.
Insoection Scooe (IP 93809)
The team reviewed the testing performed by the licensee to assure periodically the design basis functionality of the emergency core cooling system (ECCS) pump room cooling systems in the reactor building and of the heating, ventilation and air
' conditioning (HVAC) systems in the auxiliary building diesel generator and control areas. The review included the licensee's conformance with the test control (
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. requirements.of 10 CFR 50, Appendix B, and UFSAR commitments.
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Observations and Findinas
. None.of the HVAC systems reviewed were included in the Hope Creek Technical
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Specifications, and no periodic performance or surveillance testt were conducted.
Rather, system performance was monitored, during normal operation, on a continuous basis, and material condition was evaluated through a program of regularly scheduled preventive maintenance of individual components.
The team noted, during system walkdowns, that the material condition of the systems was good. Preventive maintenance procedures pertaining to temperature
- and flow instrumentation, room heater controllers, time delay relays, circuit breakers and starters, damper actuators, and air handling units provided the necessary detail
. to identify degrading conditions and to test and maintain component functionality.
Preventive maintenance periodicity was appropriately assigned. Where applicable,
. automatic system actuation in response to engineered safeguards features demands t
were verified during periodic tests of the emergency diesel generators and load sequencers.- Except for the two systems described below, the team considt, red the
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. licensee's test practices to be consistent with the UFSAR commitments.
' Table 9.4-6, " Control Room and Control Building HVAC Systems Tests and
- Inspections," describes the tests and inspection that are performed for the
- applicable systems. Items b., and e. of this table state that: (b) "For systems that :
must perform a safety-related function, periodic inservice testing of fans, valves, controls, and instrumentation in the systems is performed. Motor-operated valves ~
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and dampers are tested by opening and closing the valve or damper. Temperature,
- differential pressure readings, and flow capacity are recorded"; and (e) " Standby units are tested at periodic intervals to verify the operation of essential features.
Periodic tests of the actuation circuitry and the system components are conducted
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during normal operation."
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-The team's review of the licensee's practices for testing and inspecting the control equipment room supply (CERS) and the control area battery exhaust (CABE)
i systems determined that inservice testing of these systems was limited to checking
~ the functionality of individual components under the preventive maintenance -
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program and that the automatic standby features of these system, as described in j
section 7.3.1.1.11.6.3 of the FSAR, were not being periodically tested as specified
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in Table 9.4-6 of the UFSAR. This is a violation of 10 CFR 50, Appendix B, Criterion XI, Test Control. (VIO 50-354/98-80-01)
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Conclusions The material condition of the selected HVAC systems reflected an effective implementation of HC preventive maintenance program. However, the test and
. _ inspection practices were, in general, limited to checking the functionality of
~ individual components under the preventive maintenance program.
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Failure to test periodically the automatic features of the control equipment room supply and the control area battery exhaust systems, as described in the UFSAR,
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resulted in a violation of the Appendix B test control requirements.
l M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Walkdown of DC System a.
Scoce of Insoection (IP 93809)
l The team performed a walkdown of the dc system equipment to assess the material condition of the de system components, b.
Observations and Findinas i
DC Switchaear in response to multiple industry concerns, including NRC Information Notice 97-53,
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i racked-out position,- the licensee verified, by engineering judgement, that the
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480 Volt circuit breakers did not represent a seismic hazard to other safety-related equipment in the area. They did not, however, evaluate the impact of the withdrawn circuit breakers on the seismic qualification of the switchgear itself.
During the walkdown of the 125 and 250 Volt de switchgear, the team found numerous examples of spare circuit breakers in the racked-out position. The licensee had not evaluated the seismic hazard represented by these breakers and had failed to evaluate whether the circuit breakers in their current position challenged the seismic qualification of the dc switchgear itself. To address the team's concerns in this area the licensee placed all breakers in their connected (analyzed) position and issued a performance improvement request (PIR 980311265) to evaluate the issue. Nonetheless, this example of incomplete evaluation of industry operating axperience is a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. (VIO 50-354/98-80-02)
l Batterv and Batterv Racks The team observed that the safety-related batteries were in exceptionally good condition, despite the fact that they were approximately fifteen year old. Therefore.
they had no concerns about their physical condition.
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Conclusion
-.The actions to address NRC information notice 97-53 were inadequate in that the licensee failed to evaluate whether the racked-out de breakers represented a seismic hazard to other equipment in the area, as they had previously done for the 480Vac breakers, and failed to evaluate the seismic capability of the ac and de switchgear witn breakers in other then the full-in position. This was a violation of the NRC corrective action program requirements.
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M3 Maintenance Procedures and Documentation M3.1 Manufacturer's Recommended Maintenance a.
Scoce of inspection (IP 93809)
The team reviewed manufacturer's instruction and operations manuals end qualification reports of equipment within the scope of the inspection to assess the licensee's implementation of recommended predictive and preventive maintenance.
b.
Observations and Findinos I
l Battery chargers include components such as electrolytic capacitors and printed circuit cards that require periodic replacement. The team's review of the preventive maintenance program for this equipmen determined that the above components
had not been replaced within the manufacturer's recommended replacement intervals and that the licensee had neither provided a justification for not following such recommendatlons nor developed an alternate predictive maintenance plan.
The team had two concerns with this equipment: (1) the licensee did not measure -
the battery charger output ripple which could be indicative of degraded component; and (C) the licensee apparently had not calibrated the overvoltage shutdown feature
. since the charger installation in 1983. Actuation of this feature trips the charger
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actuation of an overvoltage shutdown devic ? following the onset of an accident, l
would hamper the ability of the operator to mitigate its consequences. The team
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determined that examples of overvoltage shutdowns of the battery chargers had been previously documented (Section E7.2 of this Report). The loss of the battery
charger due to overvoltage shutdown is a significant condition adverse to quality and the licensee's failura to preclude repetition is a violation of the 10 CFR 50, Appendix B, Criterion XVI, corrective action requirements. (VIO 50-354/98-80-02)
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Each Class 1E battery room is equipped with in-duct electric heaters. For these heaters, the heater controller manufacturer recommends a periodic replacement of the air flow switch, contactors, and door gaskets. The team similarly found that no plan existed for the periodic replacement of these components.
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Conclusions For the battery chargers and the battery room in duct heater controllers, the licensee did not follow the manufacturer's recommended preventive maintenance program and had not developed an alternate program. The preventive maintenance
.that is required to maintain operability for this equipment is a concern and will be evaluated further during a future NRC inspection. (IFl 50-354/98-80-03)
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The licensee's actions for the repetitive failures of the battery charger overvoitage shutdown did not adequately address the potential cause of the failures and were in
., violation of the corrective action program requirements.
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i E1 Conduct of Engineering E1.1 Selection and Control of DC Protective Devices a.
Scope of Inspection (93809)
The team reviewed selected protective devices used in the de circuits to assess the interface between the de system and supported systems.
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Observations and Findinos The team's review of calculation E7.9, "125/250 Vdc System Protective Relay,"
determined that the thermal overload (TOL) devices for the HPCI and RCIC loads -
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- powered from the de motor control centers had been incorrectly selected. The team found that, although the ambient temperature specified in the UFSAR for the de motor control center (MCC) areas, in the reactor building, was 148 F, the selection process for the TOL heaters was based on a 104*F ambient. This difference in temperature would require a 75% correction factor for the nonambient-compensated thermal overload relays.
The licensee stated that recently performed ventilation calculations ;ndicated a maximum area temperature of only 117*F. In this case, the required correction factor would be only 93%. However, the calculations were preliminary. In addition to the lack of ambient temperature compensation, the team found that the calculation had not addressed the different TOL response to de currents versus three-phase ac currents.
i The team's reviev, of the protective device setting for the 20 kVA safety-related inverters found that, in calculating the full load current, the licensee had neglected to assume a minimum de voltage of 106 Volts and the inverter efficiency estimated to be 70%. This total error of 50% was, however, offset by the present calculated inverter loading of 10 kVA.
The incorrect temperature used in the selection of the TOL heaters and the incorrect inverter voltage and efficiency in the design setting of the inverter protective device are example of inadequate translation of design bases into design specifications and are a violation of 10 CFR 50, Appendix B, Criterion lil, Design Control.
1(YlO 50-354/98-80-04)
in response to a Notice of Violation, documented in inspection report 50-354/94-19, PSE&G indicated that the review of the safety-related fuses would be completed by the end of the seventh refueling outage. Hope Creek came out of that outage at the
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end of 1997. The team review of drawing E-0118-0," Schematic Meter and Relay l
Diagram, 250 Volt DC System," found several instrument and control fuses,
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. connected directly to the main dc bus, for which only the. rating (6 Amps) was identified. The licensee initiated action request (AR) 980312261 to address the missed fuses and to ensure that the fuse critical voltage and fault interruption characteristics, as well as the use of ac fuses in de applications were reviewed.
This failure constitutes a violation of minor significance and is not subject to nr rmal enforcement action.
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Conclusions The selection of thermal overload devices for the HPCI and RCIC loads powered from the de motor control center was inadequate in that the associated calculation did not properly account for the operating temperature. This was a violation of the NRC design control requirements. Similarly, the calculation for the 20 kVA safety-related inverter full load current neglected to consider the minimum de voltage and the efficiency of the equipment. This was another example of inadequate design control.
E1.2 Calculation Control i
a.
Scope of Insoection (IP 93809)
The team reviewed selected calculations to assess the interface between those
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s calculations.
b.
Observations and Findinas The team's review of calculations related to the de system determined that the licensee had multiple calculations for the same purpose and the calculation of record was not always evident, e.g., the setpoint calculation for temperature switch TSHL-9558. In another example, the de voltage drop calculation for the dc motor
, operated valves was not superseded by the motor operated valve (MOV) voltage drop calculations performed to address the NRC generic letter (GL) 89-10, " Safety-Related Motor-Operated Valve Testing and Surveillance," program. The team considered this a weakness of the document control process.
The team also found an example where the results of one calculation were not factored into another calculation. This example pertained to the revision of the load profile in the battery sizing calculations which was not evaluated for impact on the voltage drop calculations.- This failure to revise the voltage drop calculation constitutes a violation of minor significance and is not subject to normal enforcement action.
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c.
Conclusions l
sThe team identified examples where the licensee had not clearly designated the calculation of record, as in the setpoint calculation for a temperature switch and in the voltage drop calculation for motor-operated valves, and had not ensured that the results of a calculation were reflected into another calculation. This was a weakness of the licensee's document control process.
E2 Engineering Support of Facilities and Equipment E2.1 - Procedures and Trainina for Evaluatina Chanaes to the Facility a.
Insoection Scooe (37001)
The primary objective of this portion of the inspection was to evaluate the Hope Creek process and staff training for implementing the 10 CFR 50.59 requirements.
This was accomplished by reviewing selected sections of the plant procedures and tiaining lessons and the training records of the personnel who prepared, reviewed, f.
s and approved safety evaluations (SEs).
-
b.
Observations and Findinas The licensee uses procedure NC.NA-AP.ZZ-0059(Q), "10 CFR 50.59 Applicability Reviews and Safety Evaluations," to implement the requirements of 10 CFR 50.59.
Guidance to individuals performing 10 CFR 50.59 reviews and evaluations is included in procedure NC.NA-AS.ZZ-0059(O), "10 CFR 50.59 Program Guidance."
The team's review of selected portions of these documents found them to contain adequate guidance for the proper implementation of the 10 CFR 50.59 requirements.
In response to earlier NRC findings, e.g., inspection report (IR) 50-272/96-01,the licensee concluded, in early 1996, that personnel preparing, reviewing, and approving 10 CFR 50.59 documents should receive initial formal training and refresher training every two years. Also, changes were made to the procedures and j
training was conducted on them. In general, the quality of SEs and SE Applicability Reviews (ARs) written since their revicion of the above documents was better than those written under earlier revisions.
The lesson plan for 10 CFR 50.59 training was included in document No. 0905-300.20N-5059ZZ-06. The team's review of this document found it generally acceptable. They did notice, however, that " increased probability of occurrence of an accident or malfunction of equipment important to safety" was defined, per NSAC-125 guidance, as a change from one ANSI frequency class to another. The
. lesson plan did note that this definition was different from the NRC interpretation of the requirement, but it was not clear what guidance AR and SE preparers should
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ese. The team identified no cases where the definition was incorrectly applied. The licensee indicated that the lesson plan would be updated to comply with current NRC guidance.
The team reviewed the training records of personnel who prepared, reviewed, and approved selected SEs and ARs. They identified no training record exceptions.
c.
Conclusions in general, the safety evaluation process was acceptable. The procedure contained adequate guidance for proper program implementation. The lesson plan for the 10 CFR 50.59 process was acceptable.
E2.2 Temocrarv and Permanent Modifications a.
Insoection Scooe (37001)
The objective of this review was to evaluate the licensee's implementation of the 10 CFR 50.59 requirements regarding permanent and temporary alterations to the-plant. The team's review included scope, safety evaluations (SEs), SE applicability -
reviews (ARs), and generic SEs. This review focused on changes to the Class IE battery and ventilation systems, but also evaluated changes to other systems and components made during the last two years. Of the 51 design change packages (DCPs) reviewed,16 had received only a 10 CFR 50.59 applicability review (AR),
eight were supported by a generic 10 CFR 50.59 safety evaluations (equivalent component replacements), and 27 had undergone complete safety evaluation.
b.
Observations and Findinas General The team's review of the DCPs determine that, in general, the changes to the facility properly addressed the concerns for which the DCP had been prepared. The safety evaluations for design changes prepared before 1996 were typically brief and lacking detail. Nonetheless, the team identified no unreviewed safety questions (USQ) with these changes. For more recent changes, the team found that the SEs were better written and adequately addressed the 10 CFR 50.59 requirements.
Applicability Reviews For those modifications that received only a safety evaluation applicability review, the ARs were generally sufficiently detailed to permit an evaluation of the adequacy of the conclusion that a safety evaluation was not required. Not in all cases, however, did the team reach the same conclusion as the licensee. For instance, items 5.f. and 6.f. in Table 3.3.2-2 of the HC technical specifications state that signals are generated to isolate the reactor core isolation cooling (RCIC) and high pressure coolant injection (HPCI) turbine steam supply, if the difference between the room exhaust and supply air temperatures exceeds 70*F. The same TS table also
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specifies an allowable differential temperature of 80*F. Further, UFSAR section
7.6.1.3.2 states that the temperature sensors "are strategically located in the
_equipinent area."
To resolve some concerns regarding spurious HPCI and RCIC isolation signals, probably r'ue to the inadequacy of the setpoints under certain conditions (e.g.
. winter time), the licensee issued Engineering Change Authorization (ECA) 4HE-00204, Packages 1 and 2. These engineering changes involved the relocation of temperature elements 1SKTE-NO28A and C [HPCI) and 1SKTE-NO218 and D [RCIC]
in the inlet ventilation ducts to downstream of the in-duct electric heaters.
Although the relocation of the sensors effectively resolved the isolation issue, their removal to downstream of the heaters constituted, in effect, a change of the differential temperature setpoints stated in the technical specification. This was clearly true for the winter months when the heater is on and the temperature of the sensors is raised by the heater.
t The team recognized that a steam leakage would raise the temperature in the room, shutdown the air inlet heaters and, hence, restore the setpoint to their intended uvalue.. However,.it was not evident how the inlet temperature. sensors would be--:.
,, _
- affected by the room temperature increase, for instance, and the subsequent
"
potential pressure buildup. The licensee had not prepared, during the design phase of the change, a transient analysis and stated that calculations 11-22(O),
" Temperature Transient for Leak Detection Criteria" and 11-85(O), " Leak Detection
Temperature Setpoints," had not been reviewed for the impact of these changes on
>
the conclusions contained therein and on the TS-specified setpoints.
Because UFSAR section 7.6.1.3.2 was not specific regarding the physical location of the temperature sensors (nor was it evident to the team that the exact location of the sensors was shown on any of the drawings listed in Section 1.7 of the UFSAR)
the licensee concluded that a safety evaluation was not required. Therefore, they moved the sensors without further review. Design verifications to determine the acceptability of the design change were inadequate in that they failed to evaluate the impact of the move on the TS specified setpoints. This is a violation of the 10 i
CFR 50, Appendix B, Criterion lil, Design Control. (VIO 50-354/98-86-04) The licensee initiated a level 2 Condition Report (CR) to address the issue.
Generic Safety Evaluations Regarding the use of generic safety evaluations for equivalent replacements, the team identified no specific concerns or inadequate applications of the program. The j
team did notice a discrepancy between the generic safety evaluation and the i
procedure that allows its use. Generic Safety Evaluation A-O-VARX-NSE-0727 l
" Equivalent Replacement and Document Update Generic Evaluation" requires that the replacement components " meet or exceed all the original operating
,,
characteristics, specifications, and design requirements." This requirement is more restrictive than the requirements of procedure NC.NA-AP.ZZ-OOO8(O) " Control of Design and Configuration Change, Tests, and Experiments" which permits i
justification of the differences between existing and replacement items.
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in reviewing the application of generic safety evaluations, the team identified several instances where the replacement components did not " meet or exceed all
.the original operating characteristics, specifications, and design requirements," but the justification for using the new equipment was acceptable.
Temocrary Modifications l
The team reviewed the 10 CFR 50.59 evaluations for several temporary facility modifications. No significant concerns were identified with the evaluations, but a minor discrepancy was observed between the installed and the approved configuration of a temporary modification (T-MOD). TMOD 96-028," Ventilation i
Ducting for Chem Lab ICP Exhaust Hood," showed the flexible duct hood resting on the inductively Coupled Plasma (ICP) machine fume exhaust port, but the installed configuration had the hood propped up on one side. The licensee indicated the
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. propping was done because, when the hood rested on the equipment, excessive j
exhaust flaw interfered with the analysis. The equipment was nonsafety-related
'
and installed in a non-seismic area. The licensee took appropriate corrective action, and revised the T-MOD and associated SE.
During a plant walkdown, the team observed unrestrained temperature and humidity j
recorders in the Main Control Room and Remote Shutdown Panel Rooms. The on-shift operating staff reported that the monitors had been in place for an extended
'
period and that their use had been incorporated into plant operating procedures. For instance, procedure HC.OP-DL.ZZ-0003(Q), Revision 24, " Log 3 Control Console j
Log Condition 1,2,3," Attachment 1, requires recording control room temperature l
and humidity once every 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift. Also, HC.OP-DL-ZZ-0016(Q), " Sunday Shift Routine Log," Attachment 1, requires operators to " Change Control Room Humidity
and Temperature Chart." The HVAC System Manager confirmed that the recorders were being used to complete the shift log. The team found that the licensee had issued neither a T-mod nor a DCP for the long-term use of recorders in those areas.
Because the instruments were unrestrained, the team was concerned that they might fall or be propelled during a seismic event and impact the safe shutdown of the plant during that event. Section 3.5.1.1.3 of the UFSAR states that " equipment and components installed in safety-related plant areas outside the primary containment are designed and installed so that they do not present gravitational missile hazard to safety-related structures-systems and components during or after
. a SSE" and that "non-permanently installed equipment is either removed from the safety-related areas or secured in place before reactor operation to ensure that it does not become dislodged and present a missile hazard." Because no design modification documents existed for the installation of the recorders in the above specified areas, no evaluation had been made of the required restraints or of the impact of the recorders interaction with other safety-related equipment located in
. areas in question. This is a violation of the 10 CFR 50.59 requirements.
(VIO 50-354/98-80-05) In the licensee's engineering judgement, during a seismic event, no damage to safety-related equipment is likely to occur, but issued PIR 980318240 to further review the issue.
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12 Weaknesses in the application of the 10 CFR 50.59 process were identified
. previously by the Offsite Safety Review (OSR) and the Nuclear Review Board (NRB).
!
. ; Based on the result of their review, the licensee instituted a process where all SEs
,,
are reviewed by a multi-discipline safety evaluation review team (SERT). This intermediate review should have a positive impact on their quality of the safety evaluations. However, because the SERT only reviews SEs, their impact on potentially incorrect AR conclusions is expected to be minimal.
c.
Conclusions
.For more recent changes the safety evaluations were better written and adequately
addressed the 10 CFR 50.59 requirements.
. For modifications that received only a safety evaluation applicability review (AR);
the documentation was generally sufficiently detailed to permit an evaluation of the j
adequacy of the conclusions but some weaknesses remained. For instance, in
' relocating the temperature sensors in the HPCI and RCIC rooms, the licensee judged
that a safety evaluation was not required and failed to evaluate whether the change invalidated TS-specified setpoints for system isolation. nThis was a violation of the a H,
.-,
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-design control program.
The review of safety evaluations for temporary facility modifications identified no
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- significant concerns.. However, a non-permanent installation of temperature and humidity !nstruments in the control room and in the remote shutdown panel room
-
was not included in the design modification process and did not receive a safety
. evaluation to assure that it did not adversely impact the safety functions of nearby
' safety-related equipment. This was a violation of the 10 CFR 50.59 requirements.
E2.3 ' Batterv Sizina Calculation Revision
a.
Scoos of Inspection (93809)
The licensee had revised the battery sizing calculations to address the NRC observation (inspection report 50-354/96-80)that a discrepancy. existed between the TS-specified minimum battery temperature (60*F) and the minimum temperature assumed in the battery specification and original sizing calculations (72'F). The team reviewed the calculations to assess the method and assumptions used by the licensee and adequacy of the results.
b.
Observations and Findinas The calculations for battery capacity were performed in accordance with IEEE
' Standard 485-1975,as invoked in the UFSAR. The team's review of these
.
,
calculations determined that the licensee had met the 25% capacity margin criteria l
set by the standard to account for aging degradation and that the current rating of the batteries was sufficient to supply adequate voltage to the loads. However, the L z
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team also observed that, in the 10 CFR 50.59 safety evaluation, the licensee had reduced the minimum design margin stated in the UFSAR from 5% to 0%.
IEEE 485 defines design margin as the combination of future load growth and battery condition less than optimum. NUREG 1048, the Hope Creek Safety
,
Evaluation Report, also defines design margin as less than optimum battery condition. Consistent with these documents, the UFSAR states, "This margin
,
[25%)... is in addition to a 5 to 10 percent design margin allowed for load growth and/or for less than optimum operating conditions of the battery."
.
.
' Contrary to the above statements,-in the safety evaluation they prepared for the UFSAR change notice, the licensee justified changing the design margin range from
!
5-10% to 0-10% on the assumption that the capacity margin was a consumable margin and the basis that design controls were in place to restrict load growth.
!
.This justification failed to recognize that margin was also needed for those periods when the batteries operate at less than optimum conditions. For ins.ance, the Hope Creek batteries were purchased witn cells with a fully charged electrolyte specific gravity of 1.215. Since the specific gravity drops as the sulphuric acid reacts with
- <the lead plates, during discharge, the technical specifications and the related -
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c --surveillance procedures permit the average battery specific gravity to be as low as-1.195 for 31 days and 1.205 continuously. These reductions in electrolyte specific
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gravity correspond to a battery capacity losses of approximately 6% and 3%,
respectively.
The licensee's failure to assure that the UFSAR-specified battery capacity margin
,
was maintained to compensate for loss of capacity during less than' optimum i
battery condition is a violation of 10 CFR 50, Appendix B, Criterion lil, Design Control. (VIO 50-354/98-80-04)
s
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> c.
Conclusions j
j The battery sizing calculations were performed in accordance with applicable
<
standards and the current battery capacity was sufficient to supply the required loads.
- The justification provided in the UFSAR change notice to reduce the minimum capacity margin from 5% to 0% was inadequate in that it assumed that the margin was consumable and did not take into consideration the less than optimum operating conditions of the battery, as permitted by the technical specifications and j
by the plant surveillance procedures. Failure to assure that the capacity margin was
'
properly maintained was a violation of the NRC design control requirements.
E2.4 Control Room Chiller SACS Control Valves a.
Insoection Scope (37001)
The team reviewed design change package 4EC-3662-2," Addition of Backup Air Supply to the Control Area Chiller SACS Control Valves." The review also included
' a walkdown of the completed modification.
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b.
Observations and Findinas l
uThe. licensee determined that, in the event of a loss of the nonsafety-related control air when ultimate heat sink temperature (UHS) river water temperature was less than 55'F, both HVAC trains for the control room and Class-1E panel room could
be lost. ' The licensee reasoned that, upon loss of. control air, the safety auxiliary cooling system (SACS) control valves would fail open and that the SACS heat exchanger bypass control valves would fail closed. This would result in maximum
- cooling of the SACS concurrent with maximum SACS flow through the control room
. and Class-1E panel rooms chillers. This co.idition could result in the chillers tripping
- and, hence, in the loss of the control area ventilation cooling.
To address this concern, during the 1997 refueling outage, the licensee designed K
DCP 4EC-3662-2 which involved. the installation of two backup nitrogen bottles on
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each of the four chiller package outlet SACS control valves. The safety evaluation took credit for replacing the backup air bottles every four hours.
The team verified that procedure HC.OP-AB.ZZ-0131(O) " Loss of Instrument Air
+
.- And/Or Service Air" had been revised to address monitoring the pressure in the-u-
. bottles'and replacing them when the pressure decreased to 35 psig. The team also
- verified that replacement of the bottles was practical, and that sufficient spare bottles were maintained to ensure continuous operation of the system. However, -
because questions remained about this modification, including testing of the check 1 valves that isolate the safety from the nonsafety-related portion of the system, the -
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followup of outstanding issues was transferred to the resident inspection team.
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. c.
Conclusions A preliminary review of the actions to address a licensee-identified deficiency
'regarding the potential loss of both HVAC trains in the event of a loss of contrcl air, indicated them to be acceptable and to restore the design to its intended performance basis. Followup was ongoing by the resident staff to further evaluate the adequacy of this modification.
E2.5 HVAC Desian Calculations and Svstem Configuration
- a.
Insoection Scope (93809)
The team reviewed design calculations pertaining to the selected HVAC systems.
The review addressed area heating and cooling loads, maximum design basis area temperatures, battery hydrogen generation rates, cooling coil and heater sizing, and
.' instrumentation loop accuracy and set points. The purpose of the review was to
.- verify.that the systems were sized and configured to perform their intended safety ifunction w
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b.
Observations and Findinas JMost of the calculations reviewed by the team had been performed by the plant j
architect / engineer (A/E) and accepted by PSE&G. The licensee recently had begun
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- to review and was in the process of revising the HVAC system calculations to i
support raising the plant ultimate heat sink temperature limit, but these calculations were stillincomplete and not available during the inspection. The A/E calculations I
conformed to industry standards and practices for HVAC systems. instrument set i
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points and loop tolerances were calculated using the methods contained la instrument Society of America (ISA) Standard S67.04-1982," Set points for Nuclear Safety-Related instrumentation Used in Nuclear' Power Plants." Design inputs and assumptions were consistent with the UFSAR and the Hope Creek design,
installation, and test specifications (DITS).
I
- Control Eauioment Room Sucolv (CERS)
. The team found minor inconsistencies, as described below, in the calculations and
-operating procedure for the CERS system. The CERS system provides heating, I
~ ym Hvy ventilation < and air conditioning to the auxiliary building 'lVAC equipment room, and
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- all elev_ations within the auxiliary building except the main control room. The areas -
- serviced by the system also include inverter and computer rooms, the high pressure l
coolant injection and reactor core isolation cooling battery rooms, and the cable
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spreading room.
l Procedure HC.OP-SO.GK-OOO1(Q), " Control Area Ventilation System Operation,"
directs operators to set the system supply air temperature controller per Table GK-001, which specifies a temperature range of 60 to 64*F. The procedure was unclear.as to whether the specified range represented the acceptable operating limits, or whether the controller could be set anywhere within the band. If the latter, then actual supply air temperature could exceed slightly the design limit of
,
64.4*F considering the instrument loop tolerance of 1 *F. Also, the instrument j
set point register (19855-JO402)and set point calculations appeared to confuse cooler leaving air temperature with the supply air _ temperature limit. For example,
' Attachment I of calculation GK-1(O), " Auxiliary Building - Relay Area Heating and Cooling Requirements " incorrectly assumed a cooler leaving air temperature band -
of 60 to 64*F, wnile the main body of the calculation established a maximum leaving air temperature of 62'F. The apparent errors had no adverse l'
consequences, however, since the cooling coil is sized for a leaving air temperature well below (at approximately 56*F) the maximum design assumption.
Chilled Water Svstems l
The safety-related panel room chilled water (PRCW) system supplies chilled water to
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, the air handling units for the Class 1E panel rooms, the. technical support center,
. and the remote shutdown panel room. The control room chilled water (CRCW)
l system supplies chilled water to the cooling equipment of the main control room, j
the auxiliary equipment (battery and computer) rooms, emergency switchgear rooms l
(including the emergency diesel generator panels control panels), and the SACS I
pump rooms in the reactor building.
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The design bases for the control area chilled water system (CWS), comprising the PRCW and CRCW systems, are described in Section 9.2.7.2.1 of the UFSAR. In L
. paragraph 2 of this section it is stated that, " Chilled water at 45*F from the safety-l related panel room chillers and 47 F from the control room chillers is supplied by l
the control area CWS to maintain the design ambient air temperatures in the areas
[
served." Consistently with the UFSAR statement, the design calculations of the
HVAC systems serving these areas (e.g, GM-3(O), and GK-1(O)) assumed the L
chilled water temperatures to be 45 F and 47*F, respectively.
.On March 10 and 12,1998, during control room walkdowns, the team observed that the PRCW and CRCW system chiller outlet temperatures were 47*F and 48*F, respectively which exceeded the temperatures assumed in the design calculations and stated in the UFSAR and DITS.
i To show that the systems were functioning as designed, the licensee provided i
,
calculation GJ-OOO3(Q), " Chillers and Chilled Water System Process Set points,"
L dated May 7,1985. The calculation showed that the chilled water controllers for
,
the PRCW and CRCW systems were set to maintain temperature bands of 43 -
j
. 47*F and 45 - 49*F, respectively..While the calculation justified the control room t l
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readings, it did not correctly reflect the HVAC design requirements. Furthermore,-
the calculation did not account for instrument loop tolerances.
.The incorrect chilled water temperatures were not an HVAC system operability
f concern because the heating loads assumed in the cooler sizing calculations
]
l contained large (10 - 20 percent) safety factors and the cooling coils thus had
'
excess design capacity. However, the licensee's failure to properly translate the design requirements in the setpoint calculation and implementing procedure is a violation of 10 CFR 50, Appendix B, Criterion lil, " Design Control."
(VIO 50-354/98-80-04)
i Safety and Turbine Auxiliaries Coolin9 System Reauired Flows and Heat Loads
,
I L
. PSE&G calculation EG-0020(O), "STACS - Required Flows and Heat Loads,"
i l
Revision 6, dated November 18,1997, recalculated and summarized design heat l
- loads of safety and nonsafety-related components in the auxiliary and reactor buildings serviced by the safety and turbine auxiliaries cooling system. The team found that the methods in the calculation were more refined and better justified and referenced than those in the original A/E calculations, and that the assumptions were consistent with those intended to maximize the heat loads described in
- Section 3.11.4 (Loss of Ventilation) of the UFSAR. In most cases, the large safety factors added to the room heat loads by the A/E were eliminated. However, using later editions of the American Society of Heating, Refrigerating, and Air Conditioning
!:
Engineers (ASHRAE) Fundamentals Handbook, the licensee calculated higher base
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_ cooling loads such that the cooling requirements were greater.than the original values, but less than the original values plus safety factors. Thus, the calculation
' confirmed the acceptability of the HVAC systems existing capacities.
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Notwithstanding the improvements, the team identified several weaknesses in the calculation, including: (1) The pump / motor heat gain was calculated using the brake horsepower (bhp) at the pump operating points rather than the design bhp;
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(2) the air flow rate through the filtration, recirculation, and ventilation system units was assumed to be 30,000 cubic feet per minute (cfm) instead of the minimum value of 27,000 cfm permitted by the Technical Specifications; (3) the chiller heat loads were based on design chilled water temperatures rather than on the less conservative current field settings; (4) room 4104 heat loads were assumed incorrectly to bound the ones in the four core spray pump rooms; and (5) actual rather than the less conservative design room cooler air flow rates were assumed.
)
The reduced flow rates justified by the calculation had not yet been implemented.
Therefore, there was no impact on the functionality of the areas cooled. Also, the team found that, prior to the inspection, the calculation originator already had identified these issues and was revising the calculation. However, the licensee's independent review of the original calculation had been mostly editorial indicating a less than effective review of the calculation. This licensee-identified and in the process of being corrected violation of the design control requirements is being h.a
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streated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC
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Enforcement Policy. (NCV 50-354/98-80-06)
c.
Conclusions The HVAC systems reviewed were capable of performing their safety functions.
The design calculations pertaining to the safety-related HVAC systems generally were acceptable, although some inconsistencies were identified.
For the safety-related panel room and the control room chilled water systems the team found the water temperature was not being maintained within the limits specified in the HVAC design calculations and in the UFSAR. This was a violation of the NRC design control requirements.
A calculation for safety and turbine auxiliaries cooling system required flows and heat loads, recently performed by the licensee, used more refined and better justified methods than the original calculation. However, some incorrect design l
assumptions, also identified by ttm licensee, resulted in a non-cited violation.
E2.6 HVAC System Procedures L
a.
Insoection Scooe (93809)
The team reviewed operating and alarm response procedures for the auxiliary L
building control area and diesel generator area heating, ventilation,.and air conditioning (HVAC) systems. The purpose of the review was to verify that the procedures were consistent with the plant design and licensing basis as described in l
the Final Safety Analysis Report (UFSAR) and station design, installation, and test specifications. The procedures reviewed applied to the safety and nonsafety-related l
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HVAC systems in the control and diesel generator areas of the auxiliary building including: (1) diesel area battery exhaust; (2) switchgear room cooling; (3) diesel
.arca panel room supply; (4) diesel area battery room supply heaters; (5) control area
,
l battery exhaust; (6) control equipment room supply; and (7) control area battery duct reheat.
b.
Observations and Findinas The procedures contained adequate instructions for placing in and removing the systems from service, including required independent verifications. Normal operating parameters (e.g. air flow rates) matched design values as were verified by the team. The procedures also specified maximum design room temperatures limits that were consistent with the values contained in design, installation, and test specifications (DITS) D7.5, " Hope Creek Station Environmental Design Criteria,"
D3.51, " Auxiliary Building Diesel Area Heating, Ventilation, Cooling Systems," and UFSAR Sections 9.4.1.1.4 (Control Area) and 9.4.6.1 (Diesel Area).
The team identified minor discrepancies in the operating procedures that the
.alicensee noted for correction. For instance, the team found that the diesel area 1E.
. panel room, and control area and diesel area battery room exhaust low flow alarm response actions prescribed by procedures HC.OP-AR.GM-0002(O) and HC.OP-AR.GM-0001(Q) could not be performed as written. The procedures directed certain balancing damper positions to be checked in accordance with the
"... applicable TRIS lineup." TRIS is a computerized system that generates safety tagouts for system maintenance and repair. Since the balancing dampers are not used to tag out the systems, they do not appear in the TRIS. In addition, other components that could cause a low flow condition, such as fire dampers, are not listed in TRIS. The licensee initiated AR 980225220to address the deficiency.
This failure constitutes a violation of minor significance and is not subject to normal enforcement action.
c.
Conclusions With minor exceptions, the HVAC system operating procedures reflected design and operating limits that were consistent with the UFSAR and design documents.
E2.7 Ventilation System Sinale Failure a.
Insoection Scone (93809)
The team reviewed selected portions of the HVAC systems d6 sign to ascertain their conformance with single failure requirements described in Appendix A (General Design Criteria) of 10 CFR 50 and the Final Safety Analysis Repor _ _ _ _ _ _ _ _ _ _ _ _ _
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b.
Observations and Findinas
,
, mThe. team's review of the Control Area (CABE) and Diesel Area battery exhaust-(DABE) and the Diesel Area Class 1E Panel Room Supply (DAPRS) systems determined that the failure (in the closed direction) of certain fire dampers in the supply or exhaust ducts could isolate air flow to or from safety-related equipment / battery rooms without initiating system low flow alarms. Failure of fire dampers was not considered in the system failure modes and effects analyses described in Tables 9.4-5 and 9.4-17 of the UFSAR. Specific examples are discussed below:
For the CABE system, section 9.4.1 of the UFSAR states that the duct from each battery room is connected to a common seismic Category I tornado-protected exhaust duct system. The single failure criterion for active safety-related equipment is met by using redundant equipment and controls and automatically switching from an operating system to a standby system upon detection of a single active failure.
However, closure either of damper FPD-D614 or FPD-D615 would interrupt exhaust from the reactor core isolation cooling system or high pressure coolant injection
. u.cc, a. ' maystem battery. rooms, respectively/ and the condition may.not be annunciated due A
'to the remaining system flow and the established alarm set point.
Section 9.4.6 of the UFSAR pertaining to the DABE system states that the standby ediesel generator area ventilation systems maintain a suitable operating environment
-
' for the safety-related battery rooms during all modes of operation, and that a low flow computer input actuates an alarm in the main control room upon loss of
- airflow, and starts the redundant fan automatically. However, failure of damper FPD-D151 or FPD-D998 would interrupt exhaust flow from their respective battery rooms (5609 or 5614) without causing a low flow alarm.
In the DAPRS system, failure of any one of several fire dampers could result in loss of air flow to redundant safety-related equipment roams without causing a low flow alarm. : For example, closure of damper FPD-D178 would isolate flow (approximately 5900 cfm) to inverter rooms 5615 and battery rooms 5609 and 5614 and the remaining system flow (35,100 cfm) would be above the alarm set point of L
.34,000 cfm. The inverter room and battery room high temperature design limits are 85*F and 80'F, respectively. No temperature alarms exist for the individual rooms, but a common system high return air temperature alarm annunciated in the control room. The team believed the high temperature in the affected rooms would not suffice to initiate a return air alarm.
The General Design Criteria of 10 CFR 50, Appendix A apply to Hope Creek.
.However, the licensee noted that Appendix A did not specify the conditions under which single failures of passive components in a fluid system (including HVAC)
should be considered in the design of that system.,The licensee also stated that
,.g they considered closure of a fire damper to be a passive failure, citing American National Standard ANSI /ANS-58.9-1981," Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems."
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' The standard defines " passive failure" as the failure of a component to maintain its structural integrity _or the blockage of a process flow path, due to the separation of a. valve disk from its stem, for instance. The standard also states that operator s
actions for mitigation of the consequences of single failures shall be allowed if
- suitable time and means for detection, diagnosis, and correction of single failures are provided.- The team noted that ANSl/ANS-58.9-1981 was not part of the Hope l
Creek design or licensing basis.
The team recognized that a loss of ventilation to or from the affected rooms would cause abnormal air flow patterns, abnormal differential pressure across doors, or unusually high room temperatures that could be detected by operators during.
. normal rounds. In addition, the licensee already had procedures in place and
,
temporary ventilation equipment designated for loss of normal ventilation events.
!
Lastlyf the team's evaluation of the battery room hydrogen concentration under -
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worst-case charging conditions determined that such concentration would remain well below the design limit of two percent. Therefore, they did not consider the
-
. issue to be an immediate safety concern. The adequacy of the system design
against single failures will be evaluated later, pending (1) clarification of the design j
s
_ <and licensing basis of the HVAC systems and (2) the NRC. verification of the seismic
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qualification of the fire dampers. (IFl 50-354/98-80-07)
. c.
Conclusions i
i The design of the Control Area Battery Exhaust, Diesel Area Battery Exhaust, and Control Area Class 1E Panel Room Supply systems are subject to single failures of
. fire dampers. No immediate safety concern existed regarding this issue, but an inspection followup item was opened by the NRC to verify the design basis of these HVAC systems and to evaluate their conformance to the single failure criterion and
- to the design control requirements of 10 CFR 50, Appendix A.
E7-
. Quality Assurance in Engineering Activities
'
- E7 1 - Effectiveness of Licensee Controls in Identifvino Problems
. a.
Inspection Scone (40500)
The team evaluated the effectiveness of licensee controls in identifying issues that degrade the quality of plant operations or safety. For this purpose, the team conducted plant tours, interviewed personnel, and reviewed applicable documents.
b'.
' Observations and Findinos During tours of the auxiliary and reactor buildings, the team noted that plant staff
. maintained good housekeeping standards and generally good equipment material condition in these areas. Interviews with plant personnel indicated a willingness to use the corrective action process to document deficiencies. Ample evidence of this positive attitude toward problem identification was available in the review of reactor and equipment operator logs, Quality Assessment (QA) audits, maintenance work j
. orders, surveillance, and the action request (AR) data base.
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Personnel documented deficiencies at a low threshold of significance, in a timely-manner, and in sufficient detail. For example, during 1997 the Hope Creek staff
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_.. averaged approximately 130 condition resolution (CR) reports and 220 corrective maintenance (CM) work orders per month. Management fully encouraged and supported employee identification and proper documentation of allissues, large and small.
The team reviewed the licensee's Employee Concern Program (ECP) and observed that ECP staff proactively communicated with plant personnel, conducted meaningful cultural surveys, and allowed easy and confidential alternate methods to raise concerns. The ECP received a relatively low number of safety / quality concerns and was generally viewed in a positive light by plant personnel. The licensee maintained current NRC Form 3 postings.
Despite their overall good performance in problem identification, the engineering
- staff did not always identify non-compliances with procedural and regulatory requirements,'as described below.
,Desian Control During an ECCS room cooler walkdown, the team observed that engineering had
- routed the residual heat removal (RHR) cross-over pipe in the 'D' RHR room within 2 inches of the room cooler inlet and blocked approximately 1/4 of the ir.let flow area.
The 18-inch pipe had been installed, in March 1996, to provide a discharge cross-tie
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e between the 'B' and 'D' RHR heat exchangers (DCP 4EC-3411). The licensee stated that they had not evaluated the impact of the potential flow reduction on the room cooler performance and, on March 18,1998, issued CR 980318160to evaluate such impact. Subsequently, licensee's calculations determined that the partial blocking of the cooler inlet had reduced the flow rate from 31,284 scfm to -
31,000 scfm. The required design flow was 30,000 scfm. The team found the
' calculation conservative and acceptable. However, prior to the installation of the-pipe, the licensee's design control measures were inadequate in that the licensee failed to evaluate the impact of the design change on the cooler performance and the design verification process did not ensure the adequacy of the revised design.
This is a violation of 10 CFR 50, Appendix B, Criterion 111. (VIO 50-354/98-80-04)
Test Control Nuc! ear Administrative Procedure NC.NA-AP.ZZ-0001, Nuclear Procedure System, step 5.9.5.H requires technicians to correct errors by drawing a single line through the entry, such that the incorrect entry remains legible, and to initial and date the entry. CJ December 16,1997,in implementing procedure MD-ST.PJ-OOO2 on battery 10-0-421, technicians changed the recorded temperature, in step 5.2.1,.
g from 73* F to.74*F without initialing and dating the change. Also, on.
December 16,1997,in step 5.5.1 of procedure HC.MD-ST.PK-0002 for battery 1 A-D-411, technicians changed the recorded cell No. 5 voltage from 2.12 Vdc to 2.13 Vdc without initialing and dating the chang _ _ _ _ _
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22 This f ailure to follow procedure constitutes a violation of minor significance and is not subject to formal enforcement action.
c.
Conclusions Based on the walkdowns conducted, plant staff maintained good material condition and housekeeping in the auxiliary and reactor buildings. Personnel demonstrated a willingness to initiate corrective actions for identified deficiencies and documented deficiencies at a low threshold of significance and in a timely manner. Management encouraged and supported employee identification of deficiencies.
Personnel did not always identify non-compliances with procedural and regulatory requirements. For instance, Design Engineering did not adequately review the potential effect of an RHR cross-tie modification on the 'D' RHR room cooler rated flow. The 18-inch pipe added during this modification was installed in close proximity of the room cooler inlet flow area, blocking approximately 1/4 of this area.
This was a violation of the design control process requirements.
E7.2 Effectiveness of Licensee Controls in Resolvino Problems
=
a.
Inspection Scope (40500)
The team performed a sampling inspection of the licensee's efforts to resolve identified deficiencies through various plant processes. For this purpose, the team reviewed condition resolution (CR) evaluations and corrective maintenance (CM)
work orders.
b.
Observations and Findinos The team reviewed cms and CRs involving issues of varying significance level and covering areas like human performance, equipment reliability, maintenance rework, procedural weaknesses, and design basis verification. Based on the documents reviewed, the team found that the licensee had acceptably resolved the identified deficiencies. Management supported the timely and correct resolution of such deficiencies.
Operations provided team leadership in the development of troubleshooting plans, taking into consideration design basis requirements, technical specification limiting conditions for operation, extent of condition, and root cause. They did not, however, demonstrate good ownership of and involvement in the workaround process. The workaround list appeared to be stagnant, one of the five workarounds listed did not hava an open work order number assigned, and a potential candidate (a valve leakage in the RHR sriutdown cooling supply header) was not on the list.
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Although the licensee demonstrated acceptable overall problem resolution, they did not always adequately implement the program requirements and effectively resolve
. identified deficiencies. For example, the HVAC and battery system managers did not meet managements expectations for performance trending in that they could not produce evidence of any organized trending analysis for their systems. The team also identified the following examples of inadequate problem resolution.
Emeraency Liahtina Units During a plant walkdown, the team observed EMIS tags on several emergency lighting units (ELUs). These tags related to dead batteries identified by the fire protection technicians in March 1997. The team's review of this issue determined that, although the tags were still hanging, the batteries had been rep %ced. This, however, had not occurred until after Quality Assessment (QA) identified, in August 1997, programmatic weaknesses in the Fire Protection's testing and maintenance of ELUs that resulted in several significance level 2 CRs and two level 1 CRs regarding inoperable ELUs and inadequate supply of replacement batteries in the store room.
- Although the emergency lighting unit deficiencies were eventually identified and corrected by the licensee, the team had several concerns with this issue, including:
(1) fack of followup by fire protection technicians (in March 1997) despite five ELU dead batteries had been found; (2) the quantity of inoperable ELUs found by QA (approximately 43%); (3) the fire protection technicians' failure to notify the Nuclear or Senior Nuclear Shift Supervisor, as required by the fire protection procedure; (4) lack of inventory control (no batteries were available in the storeroom); and (5) the licensee's root cause evaluation had not identified the missed opportunities for prior identification of this process weakness. This is a violation of the Hope Creek License Condition regarding the fire protection program requirements.
-(VIO 50-354/98-80-08)
SQ, Analyzers During operator log and CR database reviews, the team noted at least five cases, since July 1996, when the in-service reagent gas (H or O ) supply bottles were
2 found either " empty" or at " low pressure". In another instance, the licensee found that the number of bottles in the storeroom had dropped below the required quantity. UFSAR section 6.2.5.1 states that, following a LOCA, the Containment Atmosphere Control System (CACS) is designed to continuousiv monitor and, if necessary, alarm upon high concentration of hydrogen or oxygen in the primary containment. The hydrogen oxygen analyzer system (HOAS)is part of the CACS.
Reagent gas at the required pressure is necessary for the operability of the HOAS.
Team discussions with Engineering determined that they did not have an analysis that calculated the required minimum reagent gas pressures to ensure the proper operation of the HOAS for the post-accident operability time. Operators do receive a low reagent gas pressure alarm in the control room at 20 psig. However, based on an engineering estimate (200 psig minimum required pressure), by the time operators receive that alarm the bottle pressure is well below that which is needed to allow the analyzers to operate continuousiv following a LOCA.
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Insufficient reagent gas pressure in the supply bottles was a significant condition adverse to quality in that it could have prevented the H 0, analyzers to perform
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. wtheir design basis post-accident safety function..The licensee's failure to' assure
'
that the corrective action taken precluded repetition is a violation of 10 CFR 50,
-)
_
Appendix B, Criterion XVI (Corrective Action). (VIO 50-354/98-80-02)
Batterv Charaers On February 27,1996, the licensee determined that battery chargers tripped on
- high voltage shutdown, after restoration of the supply power, following a load shed.
l A subsequent informal operability review by the licensee determined that the condition affected only a backup charger operating in parallel with a load carrying charger. In August 1996, engineering decided to correct the condition by increasing the time delay on the battery charger circuit card, but later, the implementing procedures were placed on hold pending procedure writer editing.
t U
' On March '3J1997, the 1C-D-414 Lattery charger tripped on high voltage shutdown on the start of the 'C' service water pump. Maintenance closed the corrective e ->
-
.x : maintenance work order based on the procedure being on hold to address the '
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February 27,1996 condition. In addition, Engineering did not perform an operability determination even though the initiating event (start of the 'C' service water pump)
was different than the previous circumstance. Later, on September.12,1997, a maintenance engineer closed a request to verify the status of the time delay adjustment procedure revisions even though procedures were still on hold. Then, he closed the original CR even though the corrective actions were still pending.
On October 16,1997, the 1C-D-414 experienced another high voltage shutdown following the start of the 'C' service water pump. The licensee initiated CM 971016067to release the procedures from hold and schedule the work activities.
However, they did not evaluate the untimely corrective actions.
During the current inspection, as a result of the concerns expressed by the team relative to the adequacy of corrective action process in this matter, the licensee
!
initiated another CR to re-evaluate the 1C-D-414 high voltage shutdown trips.
I However, the licensee's failure to take timely corrective actions to prevent the high j
voltage shutdown of the safety-related battery chargers is a violation of 10 CFR 50, i
Appendix B, Criterion XVI (See also Section M3.1 of this Report).
!
- Service Water intake Structure Watertiaht Perimeter Flood Doors l
At 6:13 a.m. on March 9,1998, operators entered operating procedure HC.OP-AB.ZZ-0139, Acts of Nature, due to the tide level being at 94 feet and PSE&G anticipating it to reach 95 feet. At 95 feet,' Technical Specification (TS) 3.7.3
_ requires operators closing of all service water intake structure watertight perimeter flood doors within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Once closed, the TS requires that "... all access through the doors shall be administratively controlled."
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At 6:40 a.m., operators logged "all service water watertight doors secured," but at 6:42 a.m., they reopened them and logged them to be "under administrative
~ control.". At 7:34 a.m., the tide level reached 95 feet and, at 7:35 a.m., a log entry stated, "all watartight doors secured. Service water and south end personnel l
access door :e-opened und6r administrative control."
l
- While the tide was still above 95 feet, the team's review of the log entry questioned
!'
how the re-opening the doors complied with TS requirement considering that an L
individual had not been stationed at the *
.srtight doors to control access. Also, l
operability determination (CROD) 960 o128, dated June 5,1996, required i
operators to close the service water doors, dog them, then tighten the dogs with a wrench. The same requirement was listed on the operator's daily turnover sheet.
Based on their review of the CROD, the operating staff ordered all watertight doors closed, dogged, and tightened. Then they initiated a CR to evaluate the issue.
They later discovered that Engineering had previously re-evaluated the service water
' watertight door operability and determined that operators no longer needed to
- tighten the dogs down after closing the doors, but had failed to inform Operations
~ rof this change and to update the CROD..The re-opening of the watertight doors m
-.
without establishing administrative controls is a violation of the Technical'
Specification 3.7.3 requirements. (VIO 50-354/98-80-09)
'
c.'
Conclusions Within the scope of the inspection, the team found that the licensee acceptably resolved the identified deficiencies. Although the overall problem resolution was acceptable, the plant staff did not always adequately implement the program requirements.
,
Fire protection technicians failed to fully evaluate and report to the shift supervisor five failed batteries in emergency lighting units. Subsequent review by the licensee QA identified approximately 43% inoperable ELUS. This is a violation of the Hope Creek License Condition regarding the fire Protection Program requirements.
Failure to address the reagent gas pressure requirements in the hydrogen / oxygen analyzer and to ensure that the reagent remained above the required pressure resulted in a violation of the NRC corrective action requirements.
Failure to maintain all the service water intake structure watertight perimeter flood doors under administrative controls, during a rising tide level, resulted in a violation of the technical specifications requirements for those doors.
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E7.3 Effectiveness of Licensee Controls in Preventina Problem.e
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a.
Inspection Scoce (40500)
!
The team evaluated the effectiveness of the licensee's controls in preventing issues that degrade the quality of plant operation and safety. For this purpose, the team evaluated the processes that were in place ane'.nterviewed plant personnel.
b.
Observations and Findinas The licensee established NC.NA-AP.ZZ-0000 (NAP 0), Action Request Process, for reporting conditions requiring corrective actions. The licensee instituted NC.NA-AP.ZZ-0006 (NAP 6), Corrective Action Program, to ensure plant staff dispositioned and corrected conditions adverse to quality in accordance with NRC requirements.
The revision summary indicated that the licensee maintained the procedures current and continued to improve the processes. Based on the sample review of these procedures, the team found the instructions developed by the Corrective Action Group (CAG) for these two processes to be clear and comprehensive.
Quality Assessment (QA) performed periodic audits of the corrective action process.
A sample review of these audits indicated that they were thorough and broad-scope and documented several areas where personnel did not consistently implement program requirements. QA initiated appropriate corrective actions for these deficiencies. The team's review of several recently completed ARs found no evidence of continued weaknesses in the areas of improperly rejected ARs, inappropriate significance level assigned, level 2 cms without CRs, and items closed with open corrective actions.
The CAG develops real-time performance indicators and ensures management awareness of corrective action program performance. The team's review of current indicators showed acceptable performance in the average time to complete level 1 and level 2 evaluations. The CAG identified the average age of open condition resolution reports as an area needing improvement.
Personnel contacted during the course of the inspection appeared knowledgeable of the corrective action process and indicated that they would not hesitate to use the process to document and correct adverse conditions. The team also observed that the CAG did not provide continuing training on program requirements and expectations.
The Corrective Action Review Board (CARB) provides the final concurrence on root cause analyses and corrective actions for significance level 1 evaluations. On February 26,1998, the CARB reviewed three level 1 evaluations involving a high
. pressure coolant injection (HPCI) system failure, a missed surveillance on a diesel air system check valve, and a missed reactor protection trip functional test. The team observed that board members came to the meeting prepared, demonstrated a good questioning attitude and safety focus, challenged the presenters, and verified proper implementation of NAP 6 program requirements. The CARB rejected two out of threo reports because the documented analyses and corrective actions did not meet the program requirements.
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The team's review of this issue indicate the two rejections to represent isolated cases. Apparently, an evaluator had been individually assigned to prepare a root
.
_cause analyses for two different significance level.1 condition resolution reports, concurrently, and at a time when the individual had important and time consuming outage responsibilities. The team also determined that the root cause evaluator did not have.the requisite training and qualification to perform level 1 root cause analyses.
. Regarding evaluators assignments the team found that, of the 35 level 1 analyses.
. conducted in 1997, qualified root cause mcmbers participated in 32 evaluations, and many teams consisted of 3 or more qualified specialists. For the remaining three evaluations, involving just equipment issues, managers had assigned a single -
unqualified individual to perform the root cause evaluation. Through the review of a randomly selected root cause analysis report and the evaluator's interview, the team found that the evaluator had done a comprehensive cause and effect analysis of the condition, despite his lack of formal root cause training.
c.
Conclusions -
,The licensee maintained the procedures current and continued to improve the'
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processes. Based on a sample review of the applicable procedures, the instructions developed by the Corrective Action Group (CAG) were clear and comprehensive.
Quality Assessment performed periodic audits of the corrective action process. A sample review of these audits indicated that they were thorough and broad-scope j
and documented several areas where personnel did not consistently implement the program requirements.
i Current indicators prepared by the CAG indicated acceptable performance in the
. average time to complete level _1 and level 2 evaluations. The CAG identified the average age of open condition resolution reports as an area needing improvement.
I Personnel contacted during the course of the inspection appeared knowledgeable of
- the corrective action process and indicated that they would not hesitate to use the process to document and correct adverse conditions.
x E7.4' Ooeratina Experience Feedback a.-
Insoection Scope (40500)-
l The team evaluated the processes that provide for the incorporation of operating
. experience feedback.
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b, Observations and Findinas
[
uThe Operating Experience (OE) Group analyzes both industry and in-house (Salem l.
Station) operating experience for applicability at Hope Creek. Based on the sample i
review of applicable documents, the team found that the OE group, in the most part, acceptably dispositioned incoming industry operating experience information.
,
l In 1997, the OE staff initiated approximately 200 Action Requests (AR). At the time of the inspection,77 ARs remained open and 7 were listed as overdue. The team considered the overdue number to be potentially misleading since 75% of the OE initiated ARs are of the Business Process (BP) variety for which the due date can be extended many times by the BP action item owner.
' The team's review of OE initiated action items involving Vendor Reports, INPO Reports, NRC Generic Letters and information Notices, and other industry information also identified several examples where OE items had not been adequately reviewed and properly dispositioned. One such item involved the seismic qualification of 480 volt breakers in the " racked-out" position (NRC Info Notice 97-53). This issue is also addressed in Section M2.1 of this report. Another f
~ example pertained NRC Information Notice (IN) 95-36, Potential Problems With
~ ~ v-
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Post-Fire Emergency Lighting. A more timely and thorough evaluation of the IN concerns, could have identified some of the emergency lighting unit deficiencies that QA discovered in August 1997 (See also Section E7.2 of this report).
Administrative procedure NC.NA-AP.ZZ-OO54," Operating Experience (OE)
Program," requires the program effectiveness be monitored by means of performance indicators and that quarterly self-assessments be conducted to assure continued program effectiveness and improvements. The team noted that the OE Group did not provide monthly reports to site managers since July 1997 and that the last assessment was performed in August 1997. This assessment was done to address INPO 96-006," Performance Objectives & Criteria."
c.
Conclusions Based on the sample review of applicable documents, the Operating Experience
. Group, in the most part, acceptably evaluated and dispositioned incoming industry operating experience information. Some examples were found when the evaluation of the issues were insufficiently detailed.
Resolution of OE items was handled through the Action Request Process, albeit the screening of the incoming items to the Business Process might contribute to a less timely review of some issue.
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.E7.5 Self-Assessment a.
' Insoection Scooe - (40500)
- The teem evaluated the licensee's self-assessment program do determine their
[.
ability to evaluate and correct performance deficiencies. This was accomplished
'
.through a review of portions of applicable documents and interviews of appropriate personnel.
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[.
b.' - ' Observations a'nd Findinas The team interviewed Quality Assessment (QA) personnel, observed QA members L
interact with personnel in the plant and in daily management meetings, and p
reviewed several QA audits and monthly reports. The monthly reports provided good assessment of licensee performance, effectively used trending analysis, and, i
'
when applicable, initiated appropriate corrective actions. Based on the documents I
reviewed, QA conducted thorough and well-focused audits and surveillance.
- < con February 24,1998, the team attended the QA exit meeting for a Training anda H
<
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. Qualification Audit and a Hope Creek Operational Readiness Surveillance. The exit +
,
meeting was well attended by senior and department managers and by line supervisors. QA provided management with meaningful insights into station i-
- performance.
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The team reviewed several QA Barrier Trending Analysis reports for NRC identified violations. The QA Barrier Trending Analysis program is a QA-initiated tool to identify weaknesses that may exist within the QA oversight program. The reports
- were detailed and included recommended future QA activities to improve the assessment process.
The team reviewed recent self-assessments in the operations and maintenance areas. The respective departments conducted periodic performance based L
assessments of work practices, processes, and procedures. The reports reviewed indicated value-added evaluations by these departments with appropriate actions for j
identified weakness 3s.
')
l No recent self-assessments were available from Engineering. This was also noted in QA monthly reports (July'- December 1997)in which it was stated that, "There were no formal self-assessments scheduled or performed during this reporting j
period in Hope Creek System Engineering." However, in December 1997, QA
- observed that Hope Creek and Salem design engineering had made a positive l
attempt at an initial self-assessment of the adequacy, effectiveness, strengths, and i
weaknesses of the Design Change Program. QA also noted that, due to the-l
.. engineering department's reorganization, the self-assessment program had been
,
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placed on hold.
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in their January 1998 monthly report QA stated that the most significant weakness in the corrective action program implementation was the backlog of activities that
. existed in the engineering department, On January 30,1998, engineering developed a backlog reduction plan to address the nearly 9,000 open activities.
c.
Conclusions Based on the reports reviewed, Quality Assessment provided management with meaningful insights into station performance. They were actively involved in plant activities and had good access to management.
Self-assessment reports from operations and maintenance indicated value-added evaluations of their respective activities with appropriate actions for identified weaknesses. No recent self-assessments were available of engineering programs.
E7.6 Safety Review Committee Activities a.
Inspection Scope (40500)
The team reviewed the Safety Review Committee activities to evaluate their impact on the safety operation of the plant.
L b.
Observations and Findinos The Nuclear Business Unit's Independent Review Program consists of three advisory groups that review and evaluate items related to nuclear safety: (1) the Nuclear Review Board (NRB) which is responsible for independently reviewing, auditing and evaluating both technical and organizational matters related to safe plant operation;
'(2) the Station Operations Review Committee (SORC) which is responsible for advising the General Manager on operational matters related to nuclear safety; and (3) QA which is responsible for performing onsite independent reviews.
The team's review of selected NRB meeting reports, SORC meeting minutes, identified no areas of concern. Also interviews of three QA Onsite Independent Reviewers indicated that they were experienced and knowledgeable of the plant processes and procedures.
c.
Conclusions
,
inspection activities indicated acceptable oversight by the Nuclear Review Board, the Station Operations Review Committee, and the QA Onsite Independent Reviewers.
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E8 Miscelinneous Engineering issues (92903)
.E8.1 (Closed) Violation 50-354/97-07-04: Failure to Perform a Safety Evaluation per i
10 CFR 50.59 for a Design Modification Design change package (DCP) 4HE-OO170-2, Revision 2, did not include a safety evaluation to address control logic change to the ta tor core isolation cooling
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(RCIC) system. In the safety evaluation applicability review (AR) the licensee had concluded that the modification did not change the facility as described in the
- UFSAR. A violation was issued because the RCIC flow control diagram (FCD)
describing the system control logic was included in the UFSAR and the changes should have been reflected in the FCD.
.In their letter LR-N97800, dated January 12,1998, the licensee concurred with the finding and attributed the root cause to personnel error. To correct the error, they a
prepared safety evaluation (SE) H97-090 and UFSAR change notice H98-OO7. In the SE they concluded a USQ did not exist. To prevent recurrence, the licensee
' addressed the NRC finding in an internal newsletter and stated that they would
.c-
-issue procedural guidance regarding UFSAR changes and the 10 CFR 50.59 aa process.
The team reviewed the status of licensee's actions to address the violation and -
confirmed that the planned actions had been completed. The team also determined
that added guidence had been provided to the staff regarding the level of UFSAR review and the need for maintaining the level of UFSAR detail. The licensee had addressed this need in the internal memorandum as well as in Section 4.2 of -
procedure NC.NA-AS.ZZ-OO59(O), Revision 1, "10 CFR 50.59 Program Guidance,"
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approved February 25,1998. The team concluded that the licer.see's corrective action including the UFSAR change and revised program guidance adequately addressed the concerns. This violation is closed.
E8.2 (Closed) Unresolved item 50-354/97-07-05: Service Life of Mild Environment -
Struthers-Dunn Relays During a review of the actions to address thermally-inducad aging degradation of
. Struthers-Ounn relays in a mild environment, the NRC inspectors questioned the licensee's basis for concluding that the relays that had not been replaced when L
degradation was originally discovered, could perform their safety function during a
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in addressing their finding, the licensee had initiated a wa!kdown of safety-related panels, but had ' considered degraded only those relays with visible pieces of bearing pad material at the bottom of the relay case. The basis for this was their understa'nding that a degraded relay would first begin to chatter and that several
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due to contact sealing (arcing) or spring fatigue. The inspectors' concern was that, l
during a seismic event, degraded, but stillintact bearing pad material might fall
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apart and cause the relay to chatter. In this case, contact bouncing could I
potentially initiate unwanted system and component actions.
During the current inspection, the team reviewed the actions taken by the licensee to address the NRC concern. This review determined that the licensee had issued an AR (971117262)to document and evaluate the NRC observation. In the evaluation of the issue, however, the licensee had concluded that the operability of the 14 relays that had not yet been replaced was not an immediate concern. The basis for the licensee's conclusion was, in part, their engineering judgement that "a seismic event will not accelerate the degradation of the bearing pad material. No additional analysis had been done to support the conclusion. The team's evaluation of the qualification test report revealed that relays with aged, but apparently not disintegrating baring pad material (the report did not indicate that was the case) had failed to meet the seismic test acceptance criteria, due to contact bounce.
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The replacement of normally-energized Struthers-Dunn relays was nearly complete I-
- with three relays remaining and with work orders already drafted. These relays did-not have a system or component actuation function. Therefore, they were not a safety concern. However, the licensee's failure to identify the safety functions of the potentially degraded relays and their ability to perform such functions during and following a seismic event, when the degraded relays were first discovered was a violation of the NRC corrective action requirements,10 CFR 50, Appendix B, Criterion XVI. (VIO 50-354/98-80-02)
E8.3 (Closed) Violation 60-354/97-07-008: Service Life of Agastat Relays in Harsh Environment During a review of the qualification status of normally-energized Agastat E7000 relays in harsh environment, the NRC determined that their qualification had been based on the assumption that the coil temperature rise was only 34*F. Because results of coil temperature rise tests performed with a variety of other relays l
indicated much higher values than the 34*F assumed by the licensee, the inspectors expressed a concern that the operating temperature of the relays that remained energized for an extended penod following a design basis accident might not be enveloped by either the " aging" (212 F) or the " hostile" (172*F)
qualification tests performed by the relay manufacturer.
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Followup of the actions taken by the licensee to address the NRC concerns determined that they had tested the energized E7000 relay and found that the coil temperature rise was 110*F for the de relay and 87 F for the ac relay and, hence, well above the 34*F temperature rise that had been assumed. When the licensee
.added this va!ue to the design ambient temperature (148 F) at the mounting locations of the relay panels and the anticipated temperature rise of the panels in which the relays were mounted (20*F) as well as the margin (15'F) specified by IEEE Standard 323, they found that the resulting operating temperature was well above the qualification temperature.
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The ensuing evaluations determined that the maximum ambient temperature at the panel locations was approximately 116*F, but this value was preliminary and no
.docurnents were available, during the inspection, to support this conclusion. Also, even at this ambient temperature, the operating temperature of the relay of some relays was above the tested conditions. Rather than evaluating the safety functions and, hence, the qualification status of each relay, the licensee used operating experience (lower temperature) and the Arrhenius equation to establish qualification of the relays at the higher accident temperature. The analysis was included in the
. resolution of an AR (971121293)and had received approval by Operations and other engineering and supervisory personnel, in accordance with the operability determination process. The analysis of record, however, had not yet been revised and had not undergone the required peer review.
The team disagreed with the method used by the licensee to establish the relay post-accident qualification because: (1) No calculation existed for the revised ambient temperature and (2) there was no industry experience or recognition of the use of the Arrhenius regression analysis in reverse. Industry has used extensively and the NRC has accepted, in the past, the Arrhenius methodology to extend the
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service qualified life of components at a lower temperature, using the result of tests performed for a shorter duration at a higher temperature, but not to qualify components for shorter service periods at higher than tested temperatures.
Preliminary evaluations by the licensee, during the inspection, regarding the safety functions of the relays in question determined that the relays would either complete their safety functions within a few seconds from the onset of the accident or remain deenergized until some time after the onset of the accident. In no case, the relays would remain energized for a sufficient time to be considered normally energized.
Qualification tests by the manufacturer had cycled (energized and deenergized) the relay several at a maximum ambient temperature in excess of the one specified for the Hope Creek relays.
Based on the licensee's preliminary deterrnination that the relays in question are not continuously energized post-accident and pending complete evaluations by the licensee, the qualification status of the E7000 relays in harsh environment was not a safety concern. However, the licensee's failure to evaluate the ability of the relays to perform their safety functions, following an a':cident, using approved industry methods is a violation of the NRC corrective action program requirements, 10 CFR 50, Appendix B, Criterion XVI. (VIO 50-354/98-80-02)
E8.4 Review of Related UFSAR Sections A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAh! description highlighted the need for e special. focused review that compares plant practices, procedures and/or parameters to the UFSAR description. While performing the inspections discussed in this report, the team reviewed the applicable portions of the UFSAR that related to the area inspected. Except as discussed in Sections E2.5 and E2.7, the UFSAR wording was consistent with the observed plant practices and operating procedures.
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V. Manaaement Meetinos X1
. Exit Meeting Summary J
The teams presented the inspection results to members of licensee management at the conclusion of the inspection on March 19,1998. Subsequently, on June 8, 1998, via telephone, the NRC presented to the licensee the results of the additional-in-office inspection confucted during the period of April 27-30,1998. The licensee acknowledged the findings presented during both meetings.
The team asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED M. Beziila General Manager Hope Creek Station G. Boerschig Manager Electrical Design Engineering G. Dawson DC System Engineer R. DeNight Supervisor, Specialty Engineering A. Fakhar Manager, Maintenance C. Fricker Manager, Quality Assessment D. Garchow Director. Design Engineering J. Hilditch Supervisor, l&C Design Engineering H. Keiser Executive Vice President, Nuclear Business Unit S. F. Kobylarz Supervisor, QA Engineering D. McHugh Electrical Design Engineer J. Mc Mahon Director, QA/ Training /EP J. Morrisen Corrective Action Group M. Mouney Assistant Operations Manager, Hope Creek S. Nevelos Corrective Actions /EP/IT J. O'Connor Supervisor, Electrical Design Engineering G. Overbeck Director System Engineering K. Petroff Maintenance Engineer
. D. R. Powell Director Licensing Regulation and Fuel J. Priest Hope Creek Licensing L. Rajkowski Supervisor Electrical Systems Engineering P. D. Roberts Manager Hope Creek Systems Engineering G. Salamon Manager Hope Creek Licensing B. Simpson Senior Vice President, Nuclear Engineering i
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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-354/98-80-01 VIO 10 CFR 50, Appendix B, Criterion XI, Test Control.
(1)
Failure to correct the battery capacity test results for the test temperature. (Section M1.1)
(2)
Failure to calculate battery average capacity and evaluate need to increase test frequency.(Section M1.1)
(3)
Failure to test control room and control building HVAC system. (Section M1.2)
50-354/08-80-02 VIO 10 CFR 50, Appendix B, Criterion XVI, Corrective Action (1)
Failure to evaluate seismic hazard and design adequacy of switchgear with breakers in withdrawn position.
(Section M2.1)
(2)
Failure to correct repetitive failures of battery charger overvoltage shutdown. (Section M3.1)
(3)
Failure to ensure sufficient reagent gas supply bottle pressure for operability of H 0, analyzers. (Sect. E7.2)
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Failure to evaluate the ability of Struthers-Dunn relays -
to perform their safety functions. (Section E8.2)
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(5)
Failure to evaluate the ability of Agastat E7000 relays to perform their post-accident safety functions.
(Section E8.3)
50-354/98-80-03 IFl Preventive maintenance required to maintain operability of battery chargers and in-duct heater controllers.
(Section M3.1)
50-354/98-80-04 VIO 10 CFR 50, Appendix B, Criterion Ill, Design Control (1)
Use of incorrect temperature in the selection of TOL heaters for de loads. (Section E1.1)
(2)
Use of incorrect input voltage and efficiency in the selection of battery charger overload protective device.
(Section E1.1)
(3)
Relocation of temperature detectors without a safety evaluation of its impact on TS setpoints. (Section E2.2)
(4)
Failure to assure that the UFSAR-spc'fied battery capacity margin was maintained. (Section E2.3)
(5)
Failure to maintain chilled water temperature within limits specified in the HVAC design calculation.
(Section E2.5)
(6)
Failure to evaluate impact on cooler performance of 18" pipe installed in close proximity of inlet flow area.
(Section E1.1)
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50-354/98-80-05 VIO 10 CFR 50.59, Changes, Tests, and Experiments -
l Unrestrained installation of recorders in tha control
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room and remote shutdown room without a safety l
evaluation of its impact on safety-related equipment.
(Section E2.2)
50-354/98-80-07 IFl Potential single failure of HVAC systems due to fire damper closure. (Section E2.7)
50-354/98-80-08 VIO Hope Creek License Condition - Failure to followup identified deficiencies of emergency lighting units.
(Section E7.2)
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50-354/98-80-09 VIO
. TS Sect. 3.7.3 - Failure to maintain administrative controls regarding opening of watertight doors.
(Section E7.2)
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50-354/98-80-06 NCV inadequate review of design calculation for turbine and reactor buildings HVAC. (Section E2.5)
Closed 50-354/97-07-04 VIO Failure to perform a safety evaluation for the RCIC control logic change. (Section E8.1)
50-354/97-07-05 URI Service life of mild environment Struthers-Dunn relays (Section E8.2)
50-354/97-07-08 URI Service Life of Agastat relays in harsh environment.
(Section E8.3)
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LIST OF ACRONYMS USED A/E Architect /Er.gineer l
ANS American National Standard AOV Air-Operated Isolation Valve
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APRM Average Power Range Monitor AR Applicability Review / Action Request ASHRAE American Society of Heating, Refrigerating, and Air Conditioning Engineers BRDR Battery Room Duct Reheat CABE Control Area Battery Exhaust CACS Containment Atmosphere Control System CAG Corrective Action Group CARB-Corrective Action Review Board
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CERS Control Equipment Room Supply l
cfm Cubic Feet per Minute CFR Code of Federal Regulations CM Corrective Maintenance CR Condition Resolution CRCW Control Room Chilled Water l
CROD Condition Resolution Operability Determination l
DABE Diesel Area Battery Exhaust DABR Diesel Area Battery Room DAPRS Diesel Area Panel Room Supply DCP Design Change Package DITS Design, installation, and Testing Specification EACS Equipment Area Cooling System ECA Engineering Change Authorization ECP Employee Concern Program EDG Emergency Diesel Generator ELU Emergency Lighting Unit EPA Electrical Protection Assemblies FCD Flow Control Diagram FSAR Final Safety Analysis Report GL Generic Letter HC Hope Creek HVAC Heating, Ventilation, aiid Air Conditioning HOAS Hydrogen / Oxygen Analyzer System HPCI High Pressure Coolant injection lA Instrument Air IEEL Institute of Electrical and Electronic Engineers IFl Inspection Followup item INPO Institute of Nuclear Power Operations IST Inservice Test kVA'
kilo Volt-Amperes LER Licensee Event Report LOCA Loss of Coolant Accident MCC Motor Control Center MOV Motor Operated Valve M&TE Measuring and Test Equipment.
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NCV Non-Cited Violation NRB
' Nuclear Review Board l
.NRC Nuclear Regulatory Commission l
NSAC Nuclear Safety Analysis Center
.NUREG Repod issued by the NRC OSR Offsite Safety Review PClG Primary Containment instrument Gas PDR Public Document Room PRCW Panel room chilled water PRNMS Power Range Neutron M.onitoring System PSE&G Public Service Electric and Gas psig Pounds per Square Inch Gage QA Ouality Assurance / Quality Assessment RBTVB Reactor Building to Torus Vacuum Breaker RCIC Reactor Core Isolation Cooling RG Regulatory Guide RHR Residual Heat Removal RPS Reactor Protection System RWCU Reactor Water Cleanup SACS Safety Auxiliaries Cooling System SBO Station Blackout scfm Standard Cubic Feet per Minute SE-Safety Evaluation SERT Safety Evaluation Review Team SORC Site Operation Review Committee SRC Switchgear room cooling SRV Safety Relief Valve SSW Station Service Water STACS Safety and turbine auxiliaries cooling system TOL Thermal Overload TS Technical Specification TSHL Temperature Switch High-Low UFSAR Updated Final Safety Analysis Report URI.
Unresolved item UHS Ultimate Heat Sink VIO Violation l
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