IR 05000302/1993006
| ML20036A076 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/22/1993 |
| From: | Freudenberger, Holmesray P, Landis K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20036A075 | List: |
| References | |
| 50-302-93-06, 50-302-93-6, NUDOCS 9305100014 | |
| Download: ML20036A076 (14) | |
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[/* "'%g UNITED STATES f
NUCLEAR REGULATORY COMMISslON
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REGION 11
,7l g -)., oE 101 MARIFTTA STREET, N.W.
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' f ATLANTA, GEORGI A 30323 Report No.:
50-302/93-06 Licensee: Florida Power Corporation 3201 34th Street, South St. Petersburg, FL 33733 Docket No.:
50-302 License No.: DRP-72 Facility Name: Crystal River 3 Inspection Conducted: Februa 21 - Marcn 27, 1993
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Inspect r:
f 4u O
t P. Holmes-Ray,~ Senior Resident Inspector Datd, Signed Inspecto :
4 u; q7 d. Freudenberge',
es' dent inspector Dats Signed
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Approved by:
44t*h
K. Lan'dM, Section Chief Date Signed
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Division of Reactor Projects SUMMARY Scope:
This routine inspection was conducted by two resident inspectors in the areas of plant operations, maintenance, surveillance, and licensee action on previous inspection items. Numerous facility tours were conducted and facility operations observed. Backshift inspections were conducted on February 23, 27, March 4, 6, 9, 11, 16, 17 and 27.
Results:
In the area of plant operations, the following was identified:
An excessive plant cooldown occurred while initiating Decay Heat Removal. ** Unresolved Item 50-302/93-06-01, Engineering evaluation and failure analysis of cooldown while initiating Decay Heat Removal was opened.
(paragraph 3.a.)
Shutdown operations, including draining of the reactor coolant system, were observed to be performed in a controlled fashion.
(paragraph 3.b.)
Actions to mitigate the effects of a strong storm which passed through the area on March 13 were appropriate.
(paragraph 3.c.)
9305100014 930422 PDR ADOCK 05000302 G
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In the area of maintenance / surveillance, A Non-cited violation was identified regarding improper use of a Request for Engineering Assistance and failure of electrical maintenance personnel to recognize the out of tolerance condition during the March 16 performar.ce of SP-525A.
(paragraph 4.a.)
In the area of engineering / technical support, The licensee's engineering evaluation in response to a small area of reactor building spalling was timely and addressed relevant potential safety issues.
(paragraph 3.d.)
- Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviations.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- J. Alberdi, Manager, Nuclear Plant Operations
- G.
Boldt, Vice President Nuclear Production
- W. Conklin, Director, Nuclear Operations Materials and Controls
- R. Davis, Manager, Nuclear Plant Maintenance
- E. Froats, Manager, Nuclear Compliance
- F. Fusick, Manager, Design and Modifications
- G. Halnon, Manger, Nuclear Plant Systems Engineering
- B. Hickle, Director, Nuclear Plant Operations
- S. Johnson, Manager, Nuclear Chemical and Radiation Protection
- H. Koon, Nuclear Maintenance Supervisor
- P. McKee, Director, Quality Programs
- R. McLaughlin, Nuclear Regulatory Specialist B. Moore, Manager, Nuclear Integrated Scheduling W. Neuman, Supervisor, Inservice Inspection
- S. Robinson, Manager, Nuclear Quality Assessments W. Rossfeld, Manager, Site Nuclear Services R. Widell, Director, Nuclear Operations Site Support
- K. Wilson, Manager, Nuclear Licensing Other licensee employees contacted included office, operations, engineering, maintenance, chemistry / radiation, and corporate personnel.
NRC Resident Inspectors
- P. Holmes-Ray, Senior Resident Inspector
- R. Freudenberger, Resident Inspector
- Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragraph.
2.
Plant Status and Activities The plant was in power operation (Mode 1) until March 4, when Mid-cycle 9 maintenance outage started and continued for the duration of this inspection period.
On February 26, NRC Commissioner de Planque, the commissioner's Technical Assistant, and the Director, Division of Reactor Projects, Region II were on site for a plant tour and visit.
During the week of March 8, a Security inspection was conducted by a specialist inspector from Region II.
The results of the inspection will be documented in NRC Inspection Report 50-302/93-07.
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On March 12, Thomas Novak, Director, Division of Safety Programs, Office for Analysis & Evaluation of Operational Data, was onsite for a site visit.
On March 16, a meeting was conducted in the NRC Region II Office to discuss the scope and scheduling of the Electrical Distribution System Functional Inspection at Crystal River.
3.
Plant Operations (71707, 93702, & 40500)
Throughout the inspection period, facility tours were conducted to observe operations and maintenance activities in progress. The tours included entries into the protected areas and the radiologically controlled areas of the plant. During these inspections, discussions were held with operators, health physics and instrument and controls technicians, mechanics, security personnel, engineers, supervisors, and plant management.
Some operations and maintenance activity observations were conducted during backshifts.
Licensee meetings were attended by the inspector to observe planning and management activities.
The inspections confirmed FPC's compliance with 10 CFR, Technical Specifications, License Conditions, and Administrative Procedures.
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a.
Reactor Coolant System Cooldown On March 5, the plant was in Mode 4, Hot Shutdown, in the process of cooling down and transferring to decay heat removal system cooling.
When the "A" decay heat train was placed in service, valve DCV-177, a flow control valve in the decay heat closed cycle cooling system, failed to control properly.
The failure of the flow control valve resulted in an uncontrolled cooldown of the RCS. Operators placed the "B" decay heat train in service and secured the "A" train which terminated the uncontrolled cooldown.
The licensee reported this event to the NRC in accordance with 10 CFR 50.72 (reference EN 25194).
A cooldown of 47' F occurred within a half hour period (as measured at the outlet of the decay heat removal heat exchanger).
Bulk fluid temperatures of the RCS (as measured by hot leg, cold leg and incore detectors) indicated a cooldown of approximately 23" F.
Operators complied with the actions of EP-220, Pressurized Thermal Shock, and entered TS 3.4.9.1.
At the end of the report period, one of the actions required by the TS remained outstanding. An engineering evaluation to determine the effects of the out of limit condition on the fracture toughness properties of the RCS had not been completed.
The licensee planned to complete the evaluation prior to plant heatup. A failure analysis of the decay heat temperature control valves was also underway.
This issue remains unresolved pending completion of the engineering evaluation in accordance with TS 3.4.9.1 and the failure analysis of the decay heat temperature control valves and is identified as
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URI 50-302/93-06-01, Engineering evaluation and failure analysis of cooldown while initiating Decay Heat Removal.
b.
Shutdown Operations The inspector observed portions of two outage related operations evolutions to assess the adequacy of associated procedures and operator performance.
The evolutions observed were the establishment of 500 KV backfeed in accordance with OP-703A, Isolating, Grounding & Preparing CR-3 Transformers / Buses for Backfeeding, and the RCS drain in accordance with OP-301, Operation of the RCS.
In both cases the procedures provided sufficient information for the evolutions to be performed in a controlled manner and operations personnel involved demonstrated an appropriate focus on plant safety.
In preparation for the Crystal River Unit 3 Mid-Cycle Outage (9M),
the resident inspectors reviewed the licensee's administrative controls for operation of the RCS in reduced inventory and midloop conditions. At Crystal River Unit 3, plant administrative procedures consider the plant to be in a reduced RCS inventory condition when the RCS is not filled and vented. Midloop is defined as water level lower than four inches below the top of the flow area of the hot legs at the junction of the hot legs to the reactor vessel.
This corresponds to an elevation of 129 feet 6 inches. The licensee's planned schedule for the outage includes a period of approximately one week in reduced RCS inventory conditions and no midloop operations.
The reduced RCS inventory condition was planned to allow for the following maintenance activities:
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replacement of the pressurizer safety valves, RCV-8 and 9;
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repair of the "B" hot leg RTDs;
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repair of a manual valve in the pressurizer spray line (RCV-36); and
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potential replacement of RCP-IA seal package "0"-rings.
Prior to plant operation with RCS partially drained, the following items were completed by the resident inspectors:
1)
Review Generic Letter 88-17.
Generic Letter 88-17 and associated licensee correspondence was reviewed.
2)
Ensure the licensee has reviewed their controls and administrative procedures governing mid-loop operation.
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The licensee has established an administrative procedure, Al-504, Guidelines for Reduced Reactor Coolant System (RCS)
Inventory Operations, which applies during plant shutdowns when the RCS is not filled and vented or the fuel transfer canal is not filled to greater than 156 feet, with fuel in the reactor. The purpose of AI-504 is to provide a level of safety in excess of TS requirements.
AI-504 provides the framework for scheduling of plant outages. Schedules generated include systems unavailability as well as a " mirror" schedule which identifies systems availability for electrical power supplies and decay heat removal capability.
The " Modes 5 & 6 Daily Plant Status Report" includes electrical power supply, reactor cooling method, and containment penetration status, and is reviewed on a daily basis by plant management.
3)
Prt, ?dures are active and in use for the following requirements:
a)
CONTAINMENT CLOSURE CAPABILITY FOR MITIGATION OF RADI0 ACTIVE RELEASES.
Major containment penetrations (Equipment flatch, Personnel Hatch and Purge Valves) status is reviewed on a daily basis via the Modes 5 & 6 Daily Plant Status Report with other containment penetrations open to atmosphere logged in a control room log with responsibility for timely closure capability assigned to responsible maintenance personnel.
AI-504 requires that prior to reducing RCS level below elevation 129 feet 6 inches, the equipment batch must be temporarily installed in accordance with MP-II4, Removal and Reinstallation of Equipment Hatch.
b)
RCS TEMPERATURE - AT LEAST TWO INDEPENDENT, CONTINU0US INDICATIONS THAT ARE REPRESENTATIVE OF CORE EXIT CONDITIONS, ARE OPERABLE.
SP-301, Shutdown Daily Surveillance Log, requires recording two incore instruments temperature indication, at increasing frequencies as the RCS level is decreased. Al-504 references the SP-301 requirements.
c)
RCS LEVEL INDICATION - AT LEAST TWO INDEPENDENT, CONTINUOUS WATER LEVEL INDICATIONS OPERABLE (CALIBRATED).
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SP-301, Shutdown Daily Surveillance Log, requires recording of two reactor vessel level indications.
One from instrument RC-201-LII, RC-201-LI2, RC-202-LI or computer point R329/R330, the other based on Tygon tubing indication.
These levels are recorded at increasing frequencies as the RCS level is decreased.
Al-504 references the SP-301 requirements.
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RCS PERTURBATIONS SHOULD BE AVOIDED Al-504, Guidelines for Reduced Reactor Coolant System (RCS) Inventory Operations; OP-301, Operation of the Reactor Coolant System; and OP-404, Decay Heat Removal System, include precautions to minimize RCS perturbations when the plant is in reduced inventory or midloop operation.
e)
RCS INVENTORY ADDITION - AT LEAST TWO ADDITIONAL MEANS OF ADDING INVENTORY TO THE RCS MUST BE AVAILABLE, IN ADDITION TO THE PUMPS THAT ARE PART OF THE NORMAL RHR SYSTEMS - VERIFY OPERABILITY.
Al-504 requires two available means of adding inventory to the RCS, one of which shall be a Makeup Pump and flow path through one injection valve.
Acceptable methods of adding inventory to the RCS other than the makeup pump flow path are included as Enclosure 2 to Al-504.
f)
N0ZZLE DAMS / LOOP STOP VALVES - REASONABLE ASSURANCE IS OBTAINED THAT ALL HOT LEGS ARE NOT BLOCKED SIMULTANE0USLY UNLESS A VENT PATH IS ESTABLISHED.T0 PREVENT PRESSURIZATION OF THE UPPER PLENUM OF THE RV.
This requirement does not apply to Crystal River Unit 3, a Babcock and Wilcox NSSS, due to the configuration of the RCS.
g)
LICENSEE HAS CONTINGENCY PLANS TO REPOWER VITAL BUSSES FROM ALTERNATE SOURCE IF PRIMARY SOURCE IS LOST.
AI-504 delineates Power Supply Requirements for all modes of reduced inventory operation.
Enclosure 1 to AI-504 requires that the primary power source during reduced RCS inventory operation be backfeed from the 500 KV switchyard, with backup power sources from the 230 KV switchyard.
If the 500 KV source is not available, both 230 KV sources are required.
Both emergency diesel generators are preferred as emergency power sources per AI-504, Enclosure I, with the plant in reduced RCS inventory conditions. Operations in reduced inventory conditions are acceptable with one
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diesel available if the 500 KV backfeed is established and one 230 KV power source is available.
Deviation from this Al-504 requires PRC and Plant Manager approval.
In summary, the licensee's administrative controls contained in AI-504, Guidelines for Reduced Reactor Coolant System (RCS)
Inventory Operations, placed system availability requirements beyond those outlined in Generic Letter 88-17 on the operation of
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Crystal River Unit 3 in reduced RCS inventory operations.
c.
March 13 Storm Related Issues U
On March 13, a strong storm passed through the vicinity of the facility.
Sustained hurricane force winds in the early morning hours of March 13 resulted in a storm surge and some localized flooding of the site (vital equipment was not affected).
Several reportable events occurred as a result of the storm.
A safeguards system degradation related to detection and alarms was caused by the severe weather (EN 25240). The storm surge caused flooding of the turbine building but no damage to safety related equipment (EN 25241) and a significant portion of the offsite notification system was determined to be inoperable following the storm (EN 25259).
Eighteen of forty sirens were out of service due to storm damage and twenty two (55%) remained operable.
By the end of the report period all but two were returned to service.
A sufficient number of sirens were returned to service by March 18 to return the offsite notification system to operable status. While the system was in a degraded condition, notification was available through route alerting in accordance with the county response plans.
On March 17, light rain, combined with salt deposits on insulators from the storm, resulted in a degradation of the 230 KV switchyard. Arcing across insulators caused a loss of 230 KV power to the plant. Offsite power remained available through the 500 KV switchyard backfeed.
The "A" emergency diesel generator was utilized to power the operable decay heat removal train.
The licensee took the conservative action of separating the safeguards busses from non-safeguards busses.
The licensee made a Notif'. cation of Unusual Event at 8:40 a.m. on March 17, due to the degraded offsite power supplies.
The licensee manned the TSC, and the NRC manned the Region II Incident Response Center in a monitoring mode.
Event notifications (EN 25273) were completed in a timely fashion and accurately updated as plant condition changes warranted. The Unusual Event was terminated at 7:05 p.m. after washdown of the 230 KV switchyard and return of the offsite power and emergency diesel generators to normal alignments.
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On March 24, the licensee conducted a critique of their i
performance in response to the storm. The inspector reviewed the
items for improvement identified during the critique. They
included improvements to procedure enhancements to storm
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preparations, event declaration criteria,-communications, and
emergency equipment availability. The inspector noted that the
critique was effective at identifying details for improved performance of the organization in the future.
I In addition to the critique mentioned above, a management team was-established with representatives from Energy Control, Nuclear Operations, Crystal River Coal Plant, Combustion Turbine and Steam Plant Operations, Telecommunications, System Protection and Control, Transmission and Substation Design, and Substation Maintenance and System Facilities. The purpose of the team was to perform an in-depth critique and identify followup actions to strengthen the FPC electric distribution system against similar future conditions.
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A reactive inspection by a specialist inspector from the NRC l
Region II Office was scheduled for the week of March 29.
The purpose of'the inspection was.to assess the adequacy of the licensee's actions to protect offsite power availability prior to and following the storm.
The results of that inspection will be documented in NRC Inspection Report 50-302/93-10.
d.
Reactor Building Spalling On February 23, a security officer on a routine patrol identified a corner section of the post tensioning buttress for the horizontal containment tendons (in the vicinity of tendon 62 H 27)
in pieces on the Intermediate Building Roof.
The section was approximately 36 inches by 6 inches with a triangular cross-section on one end that tapered to a point at the opposite end.
Problem Report 93-0042 and an operability determination accordance with N0D-14, Determining Operability, were initiated. The resolution relied upon the performance of an engineering evaluation of the spalled concrete.
The engineering evaluatics was completed by call out of personnel during backshift hours and concluded that the spalled section of concrete was not in the bearing area of the tendon anchor plate and in no way affected the structural portion of the reactor building.
NRC assessment noted that the licensee's engineering evaluation was timely and addressed relevant potential safety issues.
e.
Radiological Protection Program Radiation protection control activities were observed to verify that these activities were in conformance with the facility i
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policies and procedures, and in compliance with regulatory
requirements.
These observations included:
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Entry to and exit from contaminated areas, including step-i off pad conditions and disposal of contaminated clothing;
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Area postings and controls;
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Work activity within radiation, high radiation, and contaminated areas;
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RCA exiting practices; and
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Proper wearing of personnel monitoring equipment, protective j
clothing, and respiratory equipment.
The implementation of radiological controls observed during this inspection period were proper and conservative.
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Security Control In the course of the monthly activities, the inspector included a review of the licensee's physical security program. The performance of various shifts of the security force was observed
in the conduct of daily activities to include:
protected and vital areas access controls; searching of personnel, packages, and
vehicles; badge issuance and retrieval; escorting of visitors; patrols; and compensatory posts.
In addition, the inspector observed the operational status of protected area lighting, protected and vital areas barrier integrity, and the security organization interface with operations and maintenance.
No performance discrepancies were identified by the inspectors.
4.
Maintenance and Surveillance Activities (62703 & 61726)
Surveillance tests were observed to verify that approved procedures were being used; qualified personnel were conducting the tests; tests were adequate to verify equipment operability; calibrated equipment was utilized; and TS requirements appropriately implemented.
The following tests were observed and/or data reviewed:
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SP-605, Emergency Diesel Generator Engine Inspection / Maintenance;
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SP-169D, Diesel Generator Instrumentation Calibration; i
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SP-525A and SP-525B, 500 KV Switchyard Battery Testing The inspector observed maintenance activities to verify that correct equipment clearances were in effect; work requests and fire prevention work permits, as required, were issued and being followed; quality control personnel performed inspection activities as required; and TS requirements were being followed.
Maintenance was observed and work packages were reviewed for the following maintenance activities:
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WR 308177, Bench Calibration and Replacement of Temperature Switch DL-030-TS;
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WR 303204, Remove Check Valve SWV-764 and Install Spool Piece (TMAR 92-11-06-01);
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WR 243872, Air Operators for Valves SWV-353 and SWV-354 rebuild; l
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WRs 301294, 301298, and 292462 Motor Operated Valve differential pressure testing of FWV-30, FWV-36, and BSV-16.
In addition, the inspector reviewed details associated with several worker concerns regarding maintenance activities performed primarily in the turbine building during the Refuel 8 Outage.
No safety significant issues were identified by this review.
The following item was considered noteworthy.
500 KV Switchyard Buttery Testing In order to facilitate maintenance on the electrical distribution system during long term outages, the licensee utilizes backfeed from the 500 KV switchyard through the main output transformers and the auxiliary transformers.
TS 3.8.1.2 requires one circuit between the offsite transmission network and the onsite Class lE distribution system, and one battery / charger combination supplying DC control power to the 230 KV switchyard breakers.
The TS does not require the circuit being used originate in the 230 KV switchyard, although it implies that it does.
To address this discrepancy, the licensee generated a TSI 90-02 which requires l
similar surveillance testing of the 500 KV switchyard DC control
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power in order to consider that offsite circuit operable.
Permanent resolution of this discrepancy is addressed in the I
Revised Standard Technical Specifications scheduled for implementation in the fall of 1993.
Surveillance procedures SP-525A, Weekly Battery Check (500 KV Switchyard Batteries), and SP-525B, Quarterly Battery Check 500 KV Switchyard, implement the TSI 90-02 requirements. On February 17, 1993, the corrected specific gravity for battery 1, cell 18 was slightly less than SP-525B acceptance criteria.
Electrical maintenance initiated REA 93-0178 with a proposed solution of
" accept this as adequate for this battery application".
The REA described a method to use tubing to take specific gravity readings using samples from lower in the cell to compensate for c+mtification and accepted the cell for "use as is".
. currently, cell 18 was assigned as the pilot cell for battery On review of the REA, the inspector noted that the proposed 2.
solution provided by electrical maintenance may have inappropriately predispositioned the resolution as "use as is".
Also, the REA provided a revised method for performance of the procedure and altered the acceptance criteria without revision of
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i the surveillance procedures. These issues were discussed with-i licensee engineering management.
Reiteration of the importance of not changing surveillance procedure methods and acceptance
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criteria was the subject of interoffice correspondence to all NPTS-r l
personnel and discussions during monthly NPTS staff meetings.
Problem Report 93-0078 was initiated on March 23, 1993, to' address.
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an out of tolerance reading on the pilot cell (cell 18) of battery
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During his review of the problem report, the system engineer noted that the cell was also slightly low on specific gravity
during the performance of SP-525A on March 16. The out of tolerance condition was not identified during the performance or
review of the procedure. The battery was placed on an equalizing
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charge, and the specific gravity readings improved but did not
meet acceptance criteria. The Problem Report included corrective
actions to address the improper use of the REA and the failure of
electrical maintenance personnel to recognize the out of tolerance
condition of the pilot cell on the March 16 performance of SP-
525A.
It is noteworthy that between March 16 and March 23, the 500 KV circuit was relied upon as the only offsite circuit while the i
specific gravity of the pilot cell was slightly out of toleranca.
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Also, on March 29, after a failure of control cables in the 500 KV switchyard, the battery charger was deenergized and the batteries i
supplied sufficient power to trip the breakers in the 500 KV switchyard.
l The safety significance of these issues was minimal because the
specific gravity was out of specification by a small margin and
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the battery was capable of performing its intended function as
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demonstrated by its performance when challenged on March 29.
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Therefore, this licensee-identified violation-is not being cited because criteria specified in Section VII.B of the NRC Enforcement
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Policy were satisfied. NCV 50-302/93-06-02, Improper use of an-REA and failure of electrical maintenance personnel to recognize
the out of tolerance condition during the March 16 performance of SP-525A.
Otherwise, surveillance and maintenance activities observed and discussed above were performed in a satisfactory manner in accordance with procedural requirements and met the requirements of the TS.
5.
Licensee Action on Previously Identified Inspection Findings (92702 &
92701)
(Closed) URI 50-302/92-30-02: Adequacy of engineering evaluation of degraded studs.
As documented in NRC inspection report 50-302/92-30, detail 4.b., spent fuel system valve SFV-25 was rebuilt as the result of a seat leak in the valve.
Procedure MP-ll8, Valve Maintenance for Bolted and Screwed Type
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11 Bonnets, included an inspection of the bonnet studs prior to reinstallation. The inspection required, in part, that the threads of
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the studs be chased with the nuts the full length of the stud. When this portion of the inspection failed, and no replacement studs were available, system engineering was contacted for assistance. An REA, number 930039, was processed to disposition the use of the studs.
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I the description of the request, the REA stated that the body to bonnet studs had minor deformation (necking down) between the portions of the stud that engaged the nuts. Necking down would indicate that the stud material had been stressed beyond yield point at some time. Behavior of the material under load would then be unpredictable.
The response to I
the REA stated that the portion of the studs that engage the nuts was unaffected and no signs of cracking were present in any of the studs, therefore they were acceptable for use as is.
No detailed visual or non-destructive testing of the studs was performed.
Following reassem-bly, the body to bonnet joint leaked, apparently due to difficulties with the gasket.
Replacement studs of the proper material were avail-able to replace the damaged studs during the rework of the valve.
The inspector was concerned that insufficient technical justification to support the "use as is" disposition was included in the REA documentation and that the lack of detail was apparently due in part to pressure to complete the valve reassembly to return the system to service. To prevent similar concerns in future similar situations, Problem Report 93-0025 was issued; this issue was used as a case study for engineering personnel; during the March 1993 staff meeting, the
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manager of NPTS discussed the importance of not compromising engineering
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justification or judgement to satisfy scheduling goals; and interoffice correspondence was issued outlining the limitations of using REAs for
"use as is" dispositions. The inspector concluded that these actions appeared sufficient to prevent recurrence. A violation did not. occur since the valve was not returned to service in a degraded condition.
This item is closed.
6.
Exit Interview The inspection scope and findings were summarized on March 29, 1993 with l.
those persons indicated in paragraph 1.
The inspectors described the l
areas inspected and discussed in detail the inspection results provided l
in this report.
Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
Item Number Status Description and Reference URI 50-302/93-06-01 Open Engineering evaluation and failure analysis of cooldown while initiating Decay Heat Removal..
NCV 50-302/93-06-02 Closed Improper use of an REA and failure of electrical maintenance personnel to recognize the out of tolerance l
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condition during the March 16 l
performance of SP-525A.
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URI 50-302/92-30-02 Closed Adequacy of engineering evaluation
of degraded studs.
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Acronyms and Abbreviations
AI
- Administrative Instruction a.m.
- ante meridiem CFR
- Code of Federal Regulations DC
- Direct Current EN
- Event Notification EP
- Emergency Procedure l
F
- Fahrenheit
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- Florida Power Corporation
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KV
- Kilovolt MP
- Maintenance Procedure NCV
- Non-cited Violation N0D
- Nuclear Operation Department Manual NPTS - Nuclear Plant Technical Support
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NRC
- Nuclear Regulatory Commission
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NSSS - Nuclear Steam Supply System
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OP
- Operating Procedure p.m.
- post meridiem PRC
- Plant Review Committee RCA
- Radiation Control Area i
- Reactor Coolant System REA
- Request for Engineering Assistance RHR
- Resistance Temperature Detector RV
- Reactor Vessel SP
- Surveillance Procedure TS
- Technical Specification TSC
- Technical Support Center TSI
- Technical Specification Interpretation URI
- Unresolved Item WR
- Work Request
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