IR 05000346/1999010
| ML20217F845 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/08/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20217F841 | List: |
| References | |
| 50-346-99-10, NUDOCS 9910210096 | |
| Download: ML20217F845 (15) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGIONlil a-Docket No:.
50-346 License No:
Report No:
50-346/99010(DRP)
Licensee:
Toledo Edison Company Facility:
Davis-Besse Nuclear Power Station Location:
5501 N. State Route 2 Oak Harbor, OH 43449-9760 Dates:
August 2 through September 13,1999 Inspectors:
K. Zellers, Senior Resident inspector Christine Lipa, Senior Resident inspector, Perry Approved by:
' Thomas J. Kozak, Chief '
Reactor Projects Branch 4 Division of Reactor Psojects
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9910210096 991008
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PDR ADOCK 05000346 l.,
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EXECUTIVE SUMMARY
Davis-Besse Nuclear Power Station NRC Inspection Report 50-346/99010(DRP)
This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 6-week period of resident inspection.
Ooerations l
Overall, the facility was operated in a conservative and conscientious manner
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(Section 01.1).
The inspectors concluded that plant personnel exhibited a lack of sensitivity to the j
control of doors important-to-safety throughout the plant as evidenced by the
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identification that doors had either been left open or had been blocked open on several j
occasions (Section O1.2).
Operators responded promptly and thoroughly to a loss of cooling water flow to the
hydrogen cooling system (Section 01.3).
The inspectors cone'uded that shift turnovers were more thorough than in past years.
- Detracting from this was a failure of operators to activate the CCW system inoperability status light when the system was inoperable, the failure of the oncoming shift to recognize the light was not activated, and the failure of operators to print and place shift logs in the unit log book on two occasions (Section 01.4).
The restricted change process did not require that the body of a procedure be changed
which could result in procedures not being performed as intended (Section O3.1).
Maintenance Activities were planned and performed in a risk-informed manner. Pre-evolution briefs
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heightened personnel awareness of the potentialimpact of work activities. Engineering personnel provided support to maintenance activities and coordinated the more complex activities (Section M1.4).
The impact of long-term scaffolding was not being rigorously reviewed (Section M1.4).
- Fiberglass ladders in battery rooms were not fully insulated to maximize protection to
personnel and equipment (Section M1.4).
Enaineerina A Non-Cited Viohtion of Technical Specifications occurred when the licensee failed to
perform an engineering evaluation of the pressurizer after a cooldown of 160 degrees in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> occurred and prior to exceeding 500 psig reactor coolant system pressure. The apparent root causes were unclear procedural guidance and untimely corrective actions (Section E8.1).
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Report Details
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Summary of Plant Status The plant was operated at nominally 100 percent power throughout the inspection period, except for brief, small reductions of power for testing.
1. Operatig_na
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Conduct of Operations 01.1 General Comments (71707)
-The inspectors attended management meetings, reviewed Condition Reports (CRs),
attended shift briefs, and questioned plant personnel on a continuing basis. Operators controlled plant maintenance and testing activities in an effective manner. Degraded i
conditions were placed into the corrective action system. Operations managers effectively communicated pertinent concems to operations personnel in a timely manner. Plant management was aware of and prioritized efforts to address adverse conditions. Control room operators were alert and cognizant of plant activities. An
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example of good questioning attitude and attention-to-detail was exhibited when an i
equipment operator identified mis-labeled drain system valves while hanging a tagout.
I The inspectors concluded that, overall, the facility was operated in a conservative and conscientious manner.
01.2 Control of Imoortant Plant Doors a.
Insoection Scooe (71707)
The inspectors reviewed the licensee's program for control of important-to-safety doors at the station.
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Observations and Conclusions The boric acid addition tank (BAAT) room door, which is a high energy line break door, serves to protect important plant equipment from the effects of a high energy line failure.
This door was found open on July 23 and September 1,1999. These events were documented in CRs 1999-1262 and 1999-1471, respectively. The apparent cause for the door being left open was that its closing mechanism did not overcome the resistance of the floor sweep when plant employees passed through the door. Contributing to these two instances was that plant employees failed to check that the door was closed s
be_ hind them after they passed through it.
The inspectors also noted that there have been several other occasions of poor door control in the past, including negative pressure boundary doors and fire doors being left open, and important doors being blocked open for convenience. Additionally, the licensee recently submitted licensee event report (LER) 1999-002, "Both Trains of
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Emergency Ventilation Rendered inoperable Due to Unattended Open Door," which was another example of improper door usage. These items in the aggregate, indicate an insensitivity of plant personnel towards door usage.
Operations management has recently heightened employee sensitivity tc door control by
' instructing equipment operators to challenge other station personnel to self-check that doors are closed behind them and by challenging managers to improve performance in this area. Additionally, a memorandum concerning door cor, trol was distributed to all site personnel to heighten the sensitivity of the importance of proper door control.
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Conclusions The inspectors concluded that plant personnel exhibited a lack of sensitivity to the control of doors important-to-safety throughout the plant as evidenced by the
. identification that doors had either been left open or had been blocked open on several occasions.
01.3 Operator Response to a Loss of Coolina Water to the Generator Hydroaen Coolina System a.
Inspection Scope (71707)
The inspector conducted a routine walkdown of the control room on August 12,1999.
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Observations and Findinas Annunciator 16-1-E (main generator hydrogen gas pressure high) and 16-3-E (main generator hydrogen gas outlet temperature high) came into alarm at about 11:26 a.m.
A control room reactor operator (RO) immediately checked the cooling water flow to the
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hydrogen cooling system and noted very little flow. He immediately informed control room personnel and the zone operator of the condition. The control room senior reactor operator used the annunciator response procedure to respond to the annunciators. The zone operator determined that the normally open cooling water control valve for the hydrogen cooling system was closed. At 11:29, with the outside assistant shift supervisor present, the zone operafn placed the cooling water control valve in manual
control and opened it to restore coo..:a water flow to the hydrogen cooling system.
Annunciators 16-1-E and 16-3-E reset by 11:31 after the hydrogen cooling system
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cooling flow was restored. As a result of the condition, a shift status brief was performed and testing activities were suspended. It was determined that control logic for the valve had failed which caused the valve to close.
The temperature setting for annunciator 16-3-E was previously set conservatively low and, as a result, the annunciator was often in alarm. To reduce distractions to plant operators and improve their ability to recognize degrded conditions, the alarm setpoint was raised to clear the alarm condition. The inspectors determined that this effort contributed to operator's quick response to the loss of cooling water flow to the hydrogen cooling system.
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Conclusions Operators responded promptly and thoroughly to a loss of cooling water flow to the hydrogen cooling system.
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01.4 Conduct of Tumovers (71707)
During routine observations of shift turnovers, the inspectors noted that the turnovers were conducted more thoroughly than in past years. Contributing to this improvement was that the shift briefs were conducted in the work support center instead of the control room (which minimized distractions and was a better environment), and that shift
. turnover sheets were more detailed. However, the inspectors identifie'l two instances where operators exhibited inattention-to-detail associated with shift turnovers in the first instance, tne inspectors observed that operators failed to activate the component cooling water system inoperability status light in the control room when the component cooling water (CCW) system was made inoperable. Further, although operators on the oncoming shift were aware the system was inoperable, they did not recognize that the inoperability light was not illuminated. Operations management generated CR 1999-1507 to document the observation. In the second instance, the inspectors identified that the official unit log book was missing tog entries for an entire shift on two different occasions. The inspectors were informed that the operators routinely read the previous shifts' logs on a plant computer in which they were generated. However, it was management's expectation that the logs be printed and placed in the official unit log book at the end of each shift. The licensee indicated that the operators forgot to print the logs and place them in the log book at the end of their shift. In response to the observation, shift management required that the unit log be printed out on a shift basis by including it on an operator activity log. The inspectors concluded that shift turnovers were more thorough than in past years. Detracting from this was a failure of operators to activate the CCW system inoperability status light when the system was inoperable, the failure of the oncoming shift to recognize the light was not activated, and the failure.,
of operators to print and place shift logs in the unit log book'on two occasions.
02-Operational Status of Facilities and Equipment O2.1 System Walkdowns (71707)
The inspectors walked down the accessible portions of the following engineered safety features (ESF) and important-to-safety systems during the inspection period:
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component cooling water
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low voltage switchgear
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makeup pumps
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high pressure injection
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low pressure injection
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auxiliary feedwater-
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No substantive concerns were identified as a result of the walkdown?. System lineups and major flowpaths were verified to be consistent with plant procedures / drawings and the Updated Safety Analysis Report (USAR). - Pump / motor fluid levels were within their i
normal bands. - Minor oil and fluid leaks were noted on occasion. The inspectors informed plant inanagement of poor housekeeping in the turbine building truck bay.
Management stated that the area was not up to their standards and initiated actions to improve the cleanliness in the area.
- 03 Operations Procedures and Documentation O3.1 Temoorary Procedure Chanaes
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Insoection Scooe (71707)
The inspectors questioned control room operators knowledge of a temporary
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modification that had been made to a control room annunciator.
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. Observations and Findinas
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Annunciator 15-5-B (electro hydraulic control panel trouble), was in alarm due to a failed i
input. The licensee implemented Temporary Modification (TM) 99-0024 which
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authorized lifting a lead from the failed input to allow the annunciator to operate normally In response to other alarm conditions. The TM was to remain in place until the failed input was repaired. The inspectors questioned control room operators as to how the actions taken in accordane,e with the TM were reflected in the annunciator response procedure. An operator informed the inspectors that the change was not referenced in the body of the alarm response procedure but that it was accounted for in the procedure by using the restricted change process. The restricted change process is used for procedure changes that are temperary and is accomplished by attaching the change to the front of a procedure. However, the process does not require that changes be made to the body of the procedure. Plant management expected that plant personnel would check the restricted changes to a procedure prior to its use. However, the inspectors have observed that control room operators have not always checked for a restricted procedure change prior to using annunciator response procedures. The inspectors were concerned that not including procedure changes into procedure bodies could result in not performing procedures the way they are intended. In response, operations management stated that the current process could be a precursor to a human error event and generated CR 1999-1506 to evaluate the process.
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Conclusions
.The restricted change process did not require that the body of a procedure be changed which could result in procedures not being performed as intended.
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Miscellaneous Operations issues 08.1 Closeout of Old Violations in Accordance with New Enforcement Poliev Guidance The Severity Level IV violations listed below were issued in Notices of Violation prior to
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the March 11,1999, implementation of the NRC's new policy for treatment of Severity Level IV violations (Appendix C of the Enforcement Policy). Because these violations would have been treated as Non-Cited Violations in accordance with Appendix C, they I
are being closed out in this report.
Violation number 50-346/98009-03. This violation is in the licensee's corrective
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action program as TERMS A19355.
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Violation number 50-346/98005-02. This violation is in the licensee's corrective
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action program as TERMS A19325.
Violation numbers 50-346/98002-01a, and 50-346/98002-01b. These violations
are in the licensee's corrective action program as TERMS A19203, and A19205.
Violation number 50-346/98002-03. This violation is in the licensee's corrective
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action program as TERMS A19204.
II. Maintenance M1 Conduct of Maintenance M1.1 General Comments Maintenance and surveillance activities were planned and performed taking into account risk insights developed by risk assessment personnel. The heightened risk sensitivity of station personnel resulted in the request and approval of a license amendment request to extend the frequency of relatively high risk control rod drive breaker testing from a monthly to a quarterly periodicity. The extended time period required to perform this test will result in the reduction of overall plant risk.
M1.2 Maintenance and Surveillance Activities (61726. 62707)
The follow'm maintenance and surveillance testing activities were observed / reviewed during the inspection period:
Incore Instrumentation Channel Check, DB-NE-03233
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EDG 2 Monthly, DB-SC-03071
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D1 Bus Undervoltage Units Monthly Functional Test, DB-ME-03046
SFAS 18-Month Interchannel Logic Test, DB-SC-03115
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Replace Defective LEDs on RIC 45978B, MWO-99-001340
Containment Personnel Hatch Local Leck Rate Test, DB-PF-03291
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Foxboro 1/1 Current Repeater Calibration, DB-MI-05233 HV 5443A,B,&C PMs, MWO-99-1574-0000
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Testing results passed acceptance criteria. Maintenance was accomplished in accordance with maintenance procedures. Post maintenance testing demonstrated the functionality of equipment before retum to service. Minor procedure discrepancies were properly dispositioned. On occasion, system engineers observed the performance of testing activities. System engineers coordinated and provided technical assistance for system outages.. Operators were pre-briefed on the impact of maintenance or testing activities.,When applicable, an operating experience report would be used that pertained to the test or maintenance activity to heighten the importance to station personnel of performing the activity properly. Repairs to a failed auxiliary contact in a -
breaker associated with the emergency ventilation system was conducted in an l
expeditious manner,
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M1.3 Scaffoldina a.
Insoection Scope (71707)
The inspectors reviewed the licensee's control of scaffolding.
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Observations and Findinas The inspectors noted that some scaffolding remained in the plant for extended periods of time because of frequent usage (e.g., scaffolding constructed to perform monthly emergency battery light testing). The scaffolding that was installed on a long-term basis was reviewed yearly by maintenance services personnel to determine its acceptability.
The inspectors questioned licensee management about the rigor used in reviewing the long-term installation of scaffolding. For example, while scaffolding could impact the ability of equipment operators, fire protection personnel, or plant personnel to respond to an event or a fire, it was not clear that these issues were considered during the review.
In response, the maintenance manager indicated that he would review the long-term scaffolding review process.
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The inspectors also noted that although the scaffolding procedure restricted the use of metal for scaffolding used in the battery rooms, and required that metal tools over 6 inches long be insulated in order to minimize the potential for high energy battery
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The inspectors noted that the procedure did not require these metal rungs or hardware be insulated, and an inspection of the battery rooms determined that a fiberglass ladder was present with non-insulated rungs and hardware.
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l M1.4 Conclusions on the Conduct of Maintenance Activities were planned and performed in a risk-informed manner. Pre-evolution briefs heightened personnel awareness of the potentialimpact of work activities. Engineering personnel provided support to maintenance activities and coordinated the more complex activities. However, the impact of long-term scaffolding was not being rigorously reviewed, and fiberglass ladders in battery rooms were not fully insulated to maximize protection to personnel and equipment.
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. Miscellaneous Maintenance lasts (92902)
M8.1 -(Closed) LER 50-346/1998-011-01: manual reactor trip due to component cooling water system leak. This LER update provided clarifying information on the operability and functionality of the auxiliary feedwater system during the event. Additionally, updated corrective actions were provided. The original LER was closed out in inspection Report 50-346/1999-009.
M8.2 (Closed) LER 50-346/1998-009-01: reactor coolant system spray valve degraded with two of eight body to bonnet nuts missing. This LER update provided the results of a finite-element-analysis for various configurations of missing nuts on pressurizer spray valve RC-2 and provided the conclusion that RC-2 would have performed its design function under design basis accident conditions for all nut configurations. Additionally it presented the results of extent of condition inspections done during a mid-cycle outage, the results of inspections done on valve RC-2, and other corrective action efforts. This event was discussed in Inspection Report 50-346/98021 and was dispositioned as a Severity Levelill violation.
Ill. Enaineerina E2 Engineering Support of Facilities and Equipment E2.1 Containment Atmosohere Cleanuo Efforts (37551)
The inspectors reviewed the licensee's continuing efforts to address suspended corrosion product particulates in the containment atmosphere that periodically affected the operation of the reactor coolant system leak detection system (RCSLDS). The RCSLDS was periodically affected when corrosion product particulates from the containment atmosphere plugged RCSLDS filters to the point that the air flow was less than required to obtain representative samples of the containment atmosphere for reactor coolant system (RCS) leak detection purposes.' To address the situation, portable filtration units were placed into containment to clean up the air, which resulted
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In decreasing the frequency of RCSLDS degradations. The licensee concluded in its
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safety evaluation that the portable filters did not adversely affect the capability of the RCSLDS to perform its leakage detection function. However, the source of the corrosion product particulates was still unknown. The licensee planned to perform thorough inspections of the containment during the next refueling outage to detect the source.
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E8.
Miscellaneous Engineering issues (92903)
E8.1 - (Closed) LER 50-346/1999-003-00: failure to perform engineering evaluation for pressurizer cooldown rate exceeding Technical Specification (TS) limit. On April 25, 1999, during the reactor shutdown for the mid-cycle outage, operators noted that the pressurizer cooldown rate based on pressurizer temperature instrument TE RC15 indicated a 160 degree drop in a one-hour period. This occurred while filling the pressurizer from about 50 inches to 280 inches with the RCS at about 160 degrees j
(decay heat removal was providing core cooling) in accordance with procedure.
Relatively cold water entered the pressurizer through the surge line at the bottom of the
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became immersed, resulting in an almost step change of indicated temperature.
Because the cold slug of water did not mix well with the hotter water in the higher
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portions of the pressurizer, saturation conditions at the steam / water interface was j
essentially unaffected and consequently, little change of RCS pressure was observed.
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Therefore, operators determined that the indicated temperature from TE RC15 was not valid because the correct pressurizer temperature should be based on the saturation temperature of the RCS pressure. The cooldown rate based on this method did not exceed the TS LCO 3.4.9.2. limit of 100 degrees in a one-hour period. Condition l
Report 1999-0656 was generated which requested that the situation be reviewed to determine the validity of using TE RC15 for pressurizer cooldown indication.
On July 26,1999, engineering personnel reviewed CR 1999-0656 and determined that i
the initial resolution of this event was incorrect. The review determined that stratification of the pressurizer fluid existed and that TE RC15 accurately showed cooldown of the pressurizer lower shell and was therefore a valid indication. The pressurizer cooldown rate of 160 degrees per hour exceeded the 100 degrees per hour limit. The TS actions required that an engineering evaluation be performed to determine the effects of the condition on the fracture toughness of the pressurizer. An evaluation was completed on
' July 27 during which it was determined that the pressurizer was operable with a safety margin of 4.5 (greater than 1 was acceptable) and that the cooldown rate experienced had no effect on the pressurizer fatigue life and fracture toughness.
The licensee's apparent root cause determination stated that the guidance provided in the plant shutdown and cooldown operating procedure (DP-OP-06903) did not include sufficient information to provide operators with the information needed to monitor pressurizer cooldown limits. As initial actions to prevent recurrence, operators were to read the evaluation of the event and were given instructions to use the TE RC15 indication for monitoring pressurizer cooldown. A revision to the plant shutdown and cooldown procedure was initiated to provide information to prevent recurrence.
The inspectors determined that other factors contributed to the failure to perform the engineering evaluation within the required period. The licensee assumed that the initial determination that the pressurizer cooldown rate was not exceeded was correct and did not assign an aggressive review date to ensure that the situation was thoroughly
. evaluated before plant startup. Condition Report 1999-0656 was evaluated by system engineering personnel but was not given to design basis engineering personnel for
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review until the week of July 19. It was not until July 26, about 3 months after the event occurred, that design basis engineering personnel determined that the original disposition was incorrect. Additionally, regulatory affairs personnel discovered that an almost identical event happened on April 12,1998, (ref. PCAQR 1998-0547, PCAQR 1998-1172, and CR 1999-1110) and generated CR 1999-1339, to document and review the reasons for the missed opportunity to make changes to prevent recurrence.
Technical Specification 3.4.9.2.a. states, in part, that the pressurizer temperature shall be limited to a cooldown of 100 degrees in any one-hour period. The action statement for TS 3.4.9.2.a. states, in part, that with the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. On April 25,1999, the pressurizer cooldown rate was 160 degrees in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. On May 8, 1999, the licensee raised coolant pressure above 500 psig. The inspectors concluded a violation of TS 3.4.9.2.a occurred when the licensee failed to perform an engineering evaluation of the pressurizer prior to exceeding 500 psig reactor coolant system pressure after the cooldown of 160 degrees in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> occurred. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as LER 50-346/1999-003-00 (NCV 50-346/1999010-01 (DRP)).
E8.2 (Closed) IFl 50-346/97011-04(DRP): control of design calculations. This item was open pending inspection of the licensee's program for controlling design calculations. The
. Inspectors completed this review with no regulatory or safety issues noted.
E8.3 (Closed) IFl 50-346/97011-05(DRP): control of instrument information sheets. This item j
was open pending evaluation of the interrelationships between surveillance procedures and instrument information sheets. The inspectors completed this review with no regulatory or safety issues noted.
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IV. Plant Suonort
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R1 Radiological Protection and Chemistry (RP&C) Controls
. R1.1 As Low As Reasonably Achievable (ALARA) Brief (71750)
The inspectors attended an ALARA brief for a containment entry to inspect portable filtration units. The brief was thorough, and personnel who participated in the brief provided good interaction and input.
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V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on September 13,1999. The licensee acknowledged the findings presented. The inspectors 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
' Licensee
' D. H. Lockwood, Supervisor, Compliance
- D. L.- Miller, Senior Engineer, Licensing D. L. Eshelman, Manager, Operations
. F. L. Swanger, Manager, Design Basis Engineering -
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G. A. Skeel,- Manager, Security _
, G. G. Campbell, Vice President Nuclear -
G.~ M. Wolf, Engineer, Regulatory Affairs
.'J. W. Rogers, Manager, Plant Engineering J. H. Lash, General Manager, Plant Operations
~ J. O'Neill,- Supervisor, Quality improvement Process.
J. E. Reddington, Superintendent, Mechanical Services J. L. Freels, Manager, Regulatory Affairs
. L. W. Worley, Director, Nuclear Assurance M. C. Beler, Manager, Quality Assessment '
P., R. Hess, Manager, Supply _
R. B. Coad, Jr., Superintendent, Radiation Protection -
S, A. Coakley, Manager, Wcrk Managsment S. Garchow, Training Manager, NSS
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- S. P. Moffitt, Director, Nuclear Support Services-T. J.; Chambers, Supervisor, Quality Assurance NRC K. S. Zellers, Resident inspector, Davis-Besse
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, INSPECTION PROCEDURES USED
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_. IP 37551:
~ Onsite Engineering (
IP 61726:
1 Surveillance Observations
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- IP 62707i Maintenance Observation -
IP 71707:
- Plant Operations ~
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!P 71750:
Plant Support Activities
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- lP.92902:
Followup'- Maintenance:
L IP 92903:-
Followup - Engineering --
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ITEMS OPENED, CLOSED, AND DISCUSSED
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Ooened
~ 50-346/99010-01(DRP)
- NCV
- Technical Specification violation for not performing an engineering evaluation for the pressurizer following a pressurizer cooldown in excess of TS limits.
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- Gp.ged
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50-346/99010-01(DRP) :
NCV Technical Specification violation for not performing an
. engineering evaluation for the pressurizer following a
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pressurizer cooldown in excess of TS limits.
. 50-346/98009-03(DRP)'
. failure to translate emergency sump design specs into USAR
. 50-346/98005-02(DRP) -
VIO - inadequate control temperature service manifold isolation
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valves for supply hose
' 50-346/98002-01a(DRP)
VIO failure to follow water balance inventory test procedure 50-346/98002-01b(DRP)_
VIO - failure to follow procedure use and adherenco procedure 50-346/98002-03(DRP)
VIO failure to meet 10 CFR 50.72 one-hour reporting requirements 50-346/1998-011-01 LER reanual reactor trip due to component cooling water system i
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'50-346/1998-009-01
. LER-reactor coolant system spray valve degraded with two of eight body to bonnet nuts missing 50-346/1999-003-00 LER: failure to perform engineering evaluation for pressurizer cooldown rate exceeding TS limit 7"
50-346/97011-04(DRP)
IFl control of design calculations
. 50-348/97011-05(DRP) -
IFl control of instrument information sheets
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LIST OF ACRONYMS AND INITIALISMS USED ALARA
' As Low As Reasonably Achievable BAAT Boric Acid Addition Tank CCW Component Cooling Water
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CFR
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l CRSRO Control Room Senior Reactor Operator i
ESF Engineered Safety Feature IFl Inspection Followup Item
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lnspection Report LCO Limiting Condition for Operation i
MWO Maintenance Work Order.
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- Non-Cited Violation NRC Nuclear Regulatory Commission PCAQR-Potential Condition Adverse to Quality Report PDR Public Document Room RCS Reactor Coolant System i
RCSLDS Reactor Coolant System Leak Detection System RO-Reactor Operator RTD Resistance Temperature Detector TM Temporary Modification TS Technical Specification USAR-Updated Safety Analysis Report l
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