IR 05000457/1990012
| ML20034B731 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 04/18/1990 |
| From: | Sands S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20034B730 | List: |
| References | |
| 50-457-90-12, NUDOCS 9004300264 | |
| Download: ML20034B731 (22) | |
Text
{{#Wiki_filter:' ,. ' ' . . ' . , U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-457/90012(DRP) k Docket No. 50-457 License No. NPF-77 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Braidwood Station, Unit 2 1b Inspection At: Braidwood Site, Braidwood, Illinois Inspection Conducted: March 20 through April 5, 1990 Inspectors: J. A. Hopkins T. M. Tongue T. E. Taylor ,
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'tM Approved By: S. P. Sands, eting hief 4-/h-hO Reactor Projects Section IA Date Inspection Summary inspection from March 20 through April 5, 1990 (Report No. 50-457/90012(DRP)) Areas Inspected: Special inspection conducted in response to the Unit 2 inadvertent depressurization of the reactor coolant system and the rapid pressurizer cooldown event while in cold shutdown on March 18' 1990.
The review included the validation of the sequence of events, determination of , the root cause, assessment of operator performance and procedure adequacy, evaluation of the licensee's short and long' term corrective action, and determination of the radiological consequences of the event.
Results: No violations or deviations were identified.
The-cause of the event was determined to have been poor communication practices between supervisors and operators.
In addition, procedure deficiencies also contributed to the event. The licensee has instituted a program to improve control room communications by identifying non-routine evolutions which require increased discussion between the operators. Two Open items were identified.
The first item is the licensee's evaluation of the pressurizer cooldown which was in excess of the Technical Specifications limits. The second item is the licensee's evaluation of the damage to the C and D reactor coolant pump shaft seals.
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ATTACHMENT NO.
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SPECIAL OPERATING ORDER.N0s.S0-ST 0039,."HEIGHTENE0; LEVEL'CelAWARENESS.
~ -(HLA) 0F CONTROL ROOM. ACTIVITIES '
2.. RESIDUAL-HEAT REMOVAL: SYSTEM: , , s - -3.. EMERGENCY CORE COOLING SYSTEM-d ' - t i <
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. - a ,, e . ! . TABLE OF CONTENTS PAGE I.
INTRODUCTION............................................
A. SYN 0PSIS OF EVENT....................................-
B. INSPECTION PLAN......................................-
C. PERSONS CONTACTED....................................
II.- INVESTIGATIVE EFFORTS...................................
III. DESCRIPTION - BRAIDWOOD UNIT 2 INADVERTENT RCS DEPRESSURIZATION AND PRESSURIZER C00LDOWN WHILE IN COLD SHUTD0WN...................
A. NARRATIVE DESCRIPTION................................
B. CHRONOLOGY OF1THE EVENT..............................-
C. CAUSE OF THE EVENT...................................
IV. OPERATOR PERFORMANCE / PROCEDURES......................... '7-A.
OPERATOR PERFORMANCE...............................
'3 B.
PROCEDURES.........................................
' V.
RCS INVENTORY BALANCE / RADIOLOGICAL CONSEQUENCES.........
VI. PROPOSED LICENSEE CORRECTIVE ACTIONS.................... 9-A.
SHORT TERM............................-.............
B.
LONG~ TERM...........................................
VII. LICENSEE'S EVALUATION OF PRESSURIZER C00LD0WN'AND REACTOR COOLANT PUMP SEALS..............................
A. - PRESSURIZER C00LD0WN...............................
B.
REACTOR COOLANT PUMP SEALS.........................
VIII. SAFETY SIGNIFICANCE.....................................
A.
IMMEDIATE...........................................
B.
OTHER CONDITIONS...................................
IX. OTHER CONCERNS..........................................
X.
CONCLUSIONS.............................................
-j XI. OPEN ITEMS..............................................
XII. EXIT INTERVIEW..........................................
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Introduction ' A.
Synopsis of Event ! On March 18, 1990, Braidwood Unit 2 was in Cold Shutdown (Mode 5)
and steps were being taken in preparation-for the first refueling outage.
Reactor Coolant System (RCS) temperature was approximatel 135'F and pressure was 350 psig with Reactor Coolant Pumps (RCPs) y C and D in operation. The28ResidualHeatRemoval(RHRf train was operating in the shutdown cooling mode.
At about 8:31 a.m., the 2A RHR Pump was placed on recirculation' to the Refueling Water Storage Tank (RWST) to obtain a sample for boron analysis.
At 10:20 a.m., the 2A RHR pump was shut down.
The Shift Control Room Engineer (SCRE) directed the Unit 2 Nuclear Station Operator (NS0) to leave the RHR to RWST recirculation isolation valve open. The Unit 2 NSO shut the 2A RHR pump RWST suction valve and opened the 2A RHR pump RCS loop suction isolation valve. The actual control board manipulations were performed by a trainee under the direct supervision and prior concurrence of the Unit 2 NSO.
This condition created a flow path from the RCS to the RWST through the RHR recirculation line.
The Pressurizer (PZR) drained rapidly and the RCS depressurized to approximately 31 psig.
The situation was promptly recognized by operations personnel who responded by concurrently providing addi-tional charging water, secured the RCPs, shut the 2A RHR-pump _RCS l loop suction isolation valve, and restored RCS inventory with the __ ' 2A RHR pump.
Inventory in the PZR was restored at 10:28 a.m. with RCS pressure approximately 55 psig.
The 2B_RHR train continued to provide core cooling throughout the event. The licensee entered the Technical Specification Action Statement for excessive PZR - cooldown and began to restore RCS pressure. At 4:30 p.m., the RCS-I was 350 psig and PZR hot calibration level was about 50%. B.
Inspection Plan On March 20 1990, Region III dispatched an inspector to investigate theeventWIththeresident' inspectors. The general areas to be inspected were: Develop a chronology of the event.
. Determine the cause of the event.
. Assess operator performance and procedure adequacy.
. Evaluate radiological consequences and RCS inventory balance.
. Evaluate licensee's proposed corrective action.
. Assess licensee's evaluation of PZR cooldown and RCP seals.
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C.
Persons Contacted Commonwealth Edison Company (CECO)
- R. E. Querio, Station Manager
- D. E. O'Brien, Technical Superintendent
- K. i.. Kofron, Production Superintendent-
- M. E. Lohman, Assistant Superintendent - Maintenance
- R. J. Legner, Services Director
- P. G. Holland, Regulatory Assurance
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- D. F. Ambler, Health Physics Supervisor
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- G. E. Groth,-Braidwood Project Manager, PWR Projects Department
- L. W. Raney, Nuclear Safety Supervisor W. B. McCue, Operating Engineer, Unit 0 D. E. Cooper, Technical Staff Supervisor G. R. Masters, Assistant = Superintendent - Operations
. ! E. W. Carroll, Regulatory Assurance J. D. Wagner, Regulatory Assurance
- Denotes those attending the Exit Interview conducted on April 5, 1990.
In addition, other members of the Braidwood staff were contacted by j the inspectors.
II.
Investigative Efforts
On March 20, 1990, the region-based inspector arrived onsite to begin investigating the event. At 11:00 a.m., the team (J. A. Hopkins, T. M. Tongue and T. E. Taylor) met with the-licensee for a brief review of the purpose and agenda of the inspection. The licensee
provided pertinent documentation they had compiled at~ that time.
! The inspectors reconstructed and validated the: sequence of events associated with the Inadvertent RCS Depressurization and Pressurizer-Cooldown while in Cold Shutdown.
The inspectors used various' logs, i the licensee's Potentially Significant Event Report (PSE), computer data printouts and graphs of various parameters, interviews with various j personnel, observations of installed instrumentation and equipment, ~ comparison of findings with other inspectors, and corporate knowledge ' of the plant to adequately assess the event.
The licensee provided all l of the-information the inspectors needed to investigate the event in a i timely manner. The inspectors verified agreement between their
perception of the event and that of the licensee.
i As part of the inspection effort, six licensed operators who were assigned to the Unit 2 control room were interviewed. The operators i were interviewed to help validate the sequence of events, to determine the root cause of the event, to assess the adequacy of their performance ! during the event and to determine the adequacy of the procedures used.
' In addition, the inspectors attempted to determine what influence, if any, the presence of the 0JT trainee had on the event.
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The operators were -interviewed singly by the inspectors for approximately 45 minutes.
' The operators were fully cooperative and appeared to respond with candor to'the questions.
(It should be noted that the operators had been interviewed by the Braidwood Station Management prior to the NRC-interviews.) The inspectors felt confident that valid conclusions could be drawn based on the interviews.- The licensee performed an analysis of the event on the Braidwood simulator.. In order to assess a broad range of possible safety concerns, the licensee reconstructed the event several times with various initial conditions. This was witnessed by the inspector and the results were consistent with those expected.
This is discussed further in Section VIII.B.
, Several issues and concerns were' presented to the licensee at the exit meeting held on April 5, 1990, at the conclusion of the inspection.
III. Description - Braidwood Unit 2 Inadvertent RCS Depressurization and Pressurizer Cooldown While in Cold Shutdown (MODE 5) A.
Narrative Description On March 18, 1990, Braidwood Unit 2 experienced a rapid depressurization of the RCS and cooldown of the PZR. This resulted when a trainee under the direct supervision and prior concurrence of the Unit 2 NSO, aligned the 2A RHR train for shut'down cooling with the RWST recirculation valve open.
At the time, the unit was - in Cold Shutdown'(Mode 5) in preparation for its first refueling outage.
The incorrect alignment of the 2A RHR train resulted in approximately 9800 gallons of reactor coolant and makeup water flowing to the RWST.
There is no evidence of an unplanned release to the environment.. In addition, RCS cooling was maintained and the Reactor Vessel Level Indicating System (RVLIS) indicated 100% throughout the event.
B.
Chronology of the Event On March 18, 1990, Braidwood Unit 2 was in Cold Shutdown (Mode 5) and steps were being taken to prepare the unit for its first refueling outage. At the beginning of the second shift (0700), RCS temperature was approximately 135'F. and pressure was 350 psig.
PZR hot calibration level was-approximately 62% with the PZR heaters and spray in manual control.
Reactor Coolant Pumps (RCPs) C and D were operating, and the 2B RHR train was in the shutdown cooling moda with RHR letdown in service. The 2A-RHR train was idle and hvailable for shutdown cooling.
The 28 Charging (CV) pump was in o Control Tank (VCT)peration and taking its suction from the Volume The PZR Power Operated Relief Valves (PORVs) . were in the " Cold Overpressure Protection" mode and both Safety Injection (SI) pumps were tagged out of service (00S) with their power supplies removed, as required by Technical Specifications
-. .. -
, e . . , - . , . (TS) and appropriate procedures.
The RHR pump RCS loop suction isolation valves, 2RH 87018 (train A) and 2RH 8702A (train B) . were also tagged 00S for protection against a RCS wide range loop pressure detector failure and to provide additional cold over-l pressure protection. The operators were at step 45, (Shutdown Main . L Steam System) of procedure 2BwGP 100-5, " Plant Shutdown and Cooldown."
' Most of the primary systems were operating on the B train to support _- . a SI test performed on the A train the previous day.
Preparations l _ were being made to switch the applicable systems to the A train to conduct the same test later that day on train B.
An extra Shift Foreman (SF) was coordinating the activities outside of the control ' room to switch the applicable systems to the A train for the SI test.. An extra NSO was stationed to specifically monitor PZR level and RCS_ temperature to allow the unit NSO to maintain an overview of the entire plant. status while switching components to the A train.
At about 8:31 a.m., the 2A RHR pump was placed on re' circulation to - i the Refueling Water Storage. Tank (RWST) to obtain a boron sample.
The sample was needed to verify that RHR boron concentration was greater than RCS boron concentration prior to placing the 2A RHR train in shutdown cooling. This required opening the RHR system discharge to RWST isolation valve,.2RH 8735.
(2RH8735.isa normally locked shut manual isolation valve.) This was done by.
approved procedure 2Bw0P RH-5, "RH System Startup for Recirculation."
The actual control room manipulations were performed by(0JT) program a trainee
enrolled in the Initial Licensing On-The-Job Training _ I All actions were performed under the direct supervision and prior concurrence of the Unit 2 NSO.
At about 10:20 a.m., chemistry personnel reported that the-sample had been drawn, and the 2A RHR pump was shut down.
The Unit 2
NS0 asked the Shift Control Room Engineer (SCRE) what position he wanted valve RH 8735 left in. The SCRE instructed the unit NS0 to leave the valve open.
l At about 10:21 a.m., the unit NSO and trainee shut the 2A RHR pump RWST suction valve, 2RH 8813A and opened the 2A-RHR-pump RCS loop suction isolation valve, 2RH 8701A.
(2RH 8701A has a stroke time of approximately 80 seconds.) This condition created a flow path from the RCS to the RWST through the RHR recirculation line. Computer plots indicated RHR flows in excess of 5000 gpm through the 2A RHR train.
The unit NSO went to record his actions in the NSO Log.
At about 10:22 a.m., the PZR low level alarm-(17%) sounded and alerted both the unit and extra NSO.
After a brief discussion with the extra NS0, the unit NS0 diagnosed the problem. RCS pressure was approximately 320 psig.
At about 10:23 a.m., the hot calibration PZR level channels were below the indicating range and RCS pressure was approximately 80 psig. The extra NSO increased charging to maximum, switched-CV suction to the RWST, and stopped-the RCPs.
The extra NS0 stated that RCS C and D loop flows were less than 100% and RCPs number 1
' . - -. , > .. ,, . seal differential pressure was 25 - 50 psid when he tripped the pumps. The unit NSO attempted to1close valve 2RH 8701A.
(2RH 8701A-will not_ reposition until it has fully stroked.).The unit NSO monitored the position of valve 2RH 8701A in order to close it as soon as possible.
At about 10:24 a.m., RCS pressure was at its lowest value of 31 psig.
The Center Desk NSO was alerted by the Unit 1-NS0 that the RCPs were . off and there might be a problem on Unit-2. The Center Desk NSO.
' went to the Unit 2 control room, was briefed by the NS0s and directed the unit NSO to align the 2A RHR train for injection from ~ the RWST.
At about 10:25 a.m., valve 2RH.8701A was fully open and the unit NSO shut the valve.. The NS0s were monitoring the RVLIS 2B RHR train for indications of cavitation.
The SCRE became aware of a possible concern when he observed a group - of NS0s and the extra SF in the Unit 2 control room.- (Thearrival of the extra SF. in the Unit 2 control-room was purely coincidental.)
When the SCRE first saw the activity, he assumed the extra SF had the situation under control.
The SCRE resumed his duties until an " Equipment Attendant (EA), who had just walked into the control room, asked the SCRE if he knew there was a problem on Unit 2.
At about 10:26 a.m., valve 2RH 8701A'was' fully shut. The unit NSO , opened the 2A RHR pump RWST suction valve, 2RH.8812A and started the 2A RHR pump to restore RCS inventory.- - ' l At about 10:27 a.m., the Center Desk NS0-and the extra NS0 observed that PZR level had not recovered.
The NS0s realized that valve i I 2RH 8735 was still open and the Center Desk NS0 directed the unit I NSO to shut the RHR pump discharge header cross-tie valve, 2RH 8716A.
At about 10:28 a.m., valve 2RH 8716A was shut..This-isolated the flow path from the 2A RHR train to the RWST.
PZR inventory was restored as indicated by a rapid drop ~in PZR water space temperature.
At about 10:29 a.m., PZR hot calibration level was ap3roximately 24% and RCS pressure was 55 psig. The unit NSO stopped t1e 2A RHR pump to avoid overfilling the PZR and possibly overpressurizing the RCS, and to prevent additional cooling of the PZR.
The SCRE.went to the unit and was briefed by the Center Desk NS0 and the unit.NS0s.
Concurrently, the Shif t Engineer (SE) was informed that Unit 2 was losing PZR level by the Unit 1 SF.
At about 10:30 a.m., charging flow was reduced with PZR level at 40%. RCS pressure was approximately 55 psig.
During the reflood of
the PZR, the Surge Line and Water Space temperatures indicated a cooldown of approximately 250 F in an eight minute period. This exceeded TS 3.4.9.2.b limit of 200*F per hour.
The appropriate Action Statement was entered and complied with,
' . . . , ~ .. ' , L The SE, SCRE, and the Center Desk NSO evaluated the plant conditions ! and decided to restore PZR level and RCS pressure to their original , values before the event. The PZR temperature heatup surveillance, 2Bw0S 4.9.2-1, was entered and a controlled repressurization commenced.
i , At about 10:36 a.m., 28 RHR flow was reduced to stop the RCS cooldown.
RCS pressure was 60 psig and PZR level was 50%. Shortly afterwards, the NS0s noticed that the RCPs number 1 seal leakoffs were at a lower rate than before the event.
The SCRE directed the , L NS0s to shut the appropriate seal leakoff isolation valves.
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l At 4:30 p.m., RCS pressure was stabilized at 350 psig with PZR hot' calibration level at 50%. C.
Cause of the Event The root cause of the event was personnel error due to a lack of L communications between the SCRE and.the Unit 2 NSO.
Prior to the event, the 2A RHR train was recirculating to the RWST per 2Bw05 RH-5 to obtain a boron sample. When.the Chemistry department reported that the sample was taken, the unit NSO secured , the 2A RHR pump. The unit NSO asked the SCRE what position he' ) wanted valve RH 8735 left in.
Routinely, a unit NS0 does not ask l the SCRE about individual valve positions while completing a I procedure.
The unit NS0 would simply complete.the procedure in.
I effect.
However, a Note in the procedure gives the SE-or,his Designated Assistant the option to align valves in positions other than the positions specified in the procedure.for Modes 5 and 6.
In the interest of keeping radiation exposure As Low As ~ Reasonably Achievable (ALARA) to the field operator, the SCRE told the unit NS0 . i to leave valve RH 8735 open. The SCRE explained that this was to avoid having the field operator enter a' radiologically controlled area a second time if another boron sample was required. 'During interviews, the SCRE stated that his intent was to " hold" in the recirculation. procedure, leave the 2A RHR train alone, and wait for the sample results.
If the sample was satisfactory, the local operator would only make one trip into the radiation area to close valve RH 8735 and align the 2A RHR-train for-shutdown cooling. -The SCRE did not clearly direct the unit NS0 how to align the 2A RHR .' train.nor did the unit NS0 question the SCRE about the lineup.
During interviews, the unit NS0 stated that his perception of this discussion was to complete 2Bw05 RH-5 with the exception of shutting valve RH 8735.
The unit NS0 stated he understood the SCRE's desire to limit the field operator's trips into the radiation area.
Additionally, the unit NS0 stated that he planned to restore the 2A RHR train to the shutdown cooling mode after the sample was taken in order to provide RCS Cold Overpressure Protection via the 2A RHR suction relief valve. The unit NS0 did not inform the SCRE that l 2Bw0S RH-5 was complete, nor did he discuss his intention to return the 2A RHR train to the shutdown cooling mode.
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. , + .. . The unit NSO took the initiative to place the 2A RHR train in the shutdown cooling lineup.
This created a flow path for high pressure (350 psig) RCS water to travel through the 2A RHR pump, heat exchanger and, recirculation line to the RWST (atmospheric pressure).
The unit NSO did not inform the SCRE when this lineup was completed.
l There. appear to be several factors which contributed to the event.
One is a communications culture in the control room that permits , the NS0s to operate systems without informing the CCRE. This - , method.of communications.is utilized during normal power operations because it gives the NS03 'the flexibility to manipulate systems . - without distracting the SCRE by requesting permission for routine evolutions.
However, in Mode 5, when many systems are in abnormal configurations with numerous evolutions in progress, the SCRE needs . to be kept informed of system lineup changes. One of the licensee's immediate corrective actions was to issue.Special Operating Order 50-ST-0039, " Heightened Level of Awareness (HLA) of Control Room Activities."
(Attachment 1) (SeeCorrectiveActionsbelow.). It is perceived that this culture has developed into a generally accepted system of communications between the SCREs and the unit NS0s. The NS0s feel that when they understand what the SCRE wants, there is no' need for additional guidance and starts to complete the task. The SCREs feel that since the procedures are so well developed, when the NS0 enters a procedure, he has the knowledge to complete the' task and/or ask for help if necessary. This culture aspears to have encouraged the lack of clear direction for aligning = tie 2A RHR train from the SCRE and the lack of followup report of - system lineup changes from the unit NSO.
A second contributing factor to this event was procedural deficiencies in 2Bw0S RH-5, "RH System Startup for Recirculation," and RH-6, " Placing the RH System in Shutdown Cooling." These procedures give the SE or his designee (SCRE) the flexibility to configure the RHR system for this evolution while staying within the boundaries of approved procedures.
The licensee has revised these procedures to prevent this' specific event from recurring.
(SeeCorrectiveActionsbelow.)
l IV. Operator Performance and Procedure Adequacy to Mitigate the Event A.
Operator Performance When the PZR low level annunciator alarmed, the NS0s took immediate steps to restore RCS inventory by increasing charging flow and switching CV suction to the RWST.
The NS0s tripped the RCPs when the pumps' Net Positive Suction Heads (NPSH) and Number 1 Seal Differential Pressures were below the required values.
In addition, l the NS0s monitored the 2B RHR Train for indications of cavitation and RVLIS indication throughout the event.
The NS0s aligned the 2A RHR train for injection from the RWST as soon as vlave 2RH 8701A had fully stroked.
The NS0s recognized that PZR level was not being rapidly restored and shut valve 2RH 8716A to
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. + .. , L - .-, u , isolate the flow path. The operaters were continually aware of the PZR cooldown limits and took delibe* ate action to restore PZR level and RCS pressure in a controlled ano systematic manner.
l l The operators appeared to have quickly diagnosed the event and took prompt, effective action to mitigate the event.
B.
Procedures The inspector reviewed procedures, control. room logs, various i records and conducted personnel interviews to assess the adequacy .i , of the procedures used to mitigate the event.
Prior to.the event,
' l-the operators were at step 45, (Shutdown Main Steam System) of - procedure 2BwGP 100-5, " Plant Shutdown and Cooldown." The operators
used 2Bw0P RH-5, "RH System Startup for Recirculation," to recirculate ' the 2A RHR train.
There are two abnormal procedures PRI-1, ". Excessive Primary Plant Leakage," and PRI-10, " Loss of RH Cooling," which could-be used to ! mitigate a Mode 5 LOCA. The procedures have the initial actions for ! maintaining RCS inventory and leak isolation. However, these ! procedures do not address a Mode 5 inter-system LOCA.
The operators ! did not use these procedures to mitigate the event. The inspector ! determined that the duration of the event (approximately 6 minutes ! for the entire event) prevented the operators from referring to-
these procedures.
The operators used main control board (MCB) ! indications, system knowledge and good engineering practices to ' mitigate the event.
L V.
RCS Water Inventory Balance / Radiological Consequences - A review of the data ~after the event indicated that the volume in the-RWST increased by approximately 9800 gallons.
Calculations using initial { PZR and Surge Line volumes and temperatures, RCS volume and temperatures, and corrections for density changes indicate the PZR and Surge Line contained approximately 5500 gallons.
The remaining difference in water inventory has been attributed to voiding in the suction lines of the C l; and D RCPs.
The loss of NPSH due to RCS pressure reduction with the pumps in operation is believed to have caused localized voiding in the Steam Generator (SG) tubes and cross over piping. The NS0s' stated that RCS C_and D loop flow indications were less than 100% prior to the pumps tripping.
Personnel in containment near the RCPs stated that a louder, j , higher pitched sound occurred just prior to the pump trip.
Both of these are indicators of cavitation and voiding in the RCP suction piping.
These are " pump" effects and not " thermal" effects.
i Part of the rational for performing a RCS water inventory balance was to look for evidence of an unmonitored release of radioactive material to the environment. The inspector reviewed radiation monitor data from the Unit 2 RHR cubicles, Auxiliary Building Exhaust Stack, containment
piping penetrations and the containment area.
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. The only radiation level increases-were observed at the containment area
monitors at the 426! elevation and several of the pipe senetration area f monitors.
The largest increases were observed in the R.iR piping penetration area of the auxiliary building..These were initially 7 to ~ 15 millirem per hour (MR/HR)-and increased to 20 to 31 MR/HR after the
event. The containment radiation levels increased from 0.15 to 0.6 MR/HR.. The licensee and inspector concluded that the increases were due to corrosion products released by the rapid RCS depressurization.
The RHR cubicles and the Auxiliary Building Exhaust Stack monitors did not increase.
Based on the information above the inspector was able to determine that-there was no evidence of an unplanned release of radioactive material to the environment.
Additionally, there is no L evidence of an airborne release to the affected areas of the Auxiliary Building.
VI.
Proposed Corrective Actions , A.
Short Term Corrective Actions 's As mentioned above procedure 2Bw0P RH-5 had a' Note which gives the SE or his designee (SCRE) the flexibility to configure the RHR system differently than the procedure states.
Normally the RHR train will be aligned for injection from the RWST when RH-5 is completed.
(This is the required lineup for. Modes 1, 2 and 3.) Specifically, , this Note allows the operator to complete procedure RH-S with I valve RH8735 open and the RHR train not in the injection mode.
The licensee has issued a Temporary Procedure Change (TPC) which . moves the step-to lock valve RH8735 closed before the Note.
l The inspector identified a deficiency in procedure 2Bw0P RH-6 which gave the operator the flexibility to-align the RHR system for shutdown cooling with valve RH8735 open.
The licensee has issued a TPC which requires valve RH8753 to be locked closed and remain locked closed during this procedure.
. In the PSE report, the licensee identified the communications breakdown between the SCRE and the unit NS0 as-the apparent cause of the event.
The licensee issued a Special Operating Order,- " Heightened Level of Awareness (HLA) of Control Room Activities," to ensure that increased levels of discussion occur prior to the performance of infrequent or non-routine evolutions. The purpose of the HLA is to enhance the concepts which exist in BwAP 300-1, " Conduct of Operations," regarding communications and control of activities.
The licensee's immediate corrective actions were reviewed and determined to be acceptable.
B.
Long Term Corrective Actions On March 28, 1990, the licensee conducted the Personnel Error Evaluation Program (PEEP), which is part of their self-evaluation l
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The purpose of the PEEP is to determine the root cause of the event and possible corrective actions.
Based on the PEEP, the licensee he.s instituted a number of long term corrective actions: Better define the scope of the HLA program.
. l Search other procedures for deficiencies similar to the Note . in procedure 2Bw0P RH-5.
Review other system lineups and procedures to identify valves . I which could impact the plant in a similar manner, , Enhance the Main Control Board (MCB) system mimics to include l . manual valves.
After these items were reviewed, it was determined that the-licensee's broad based approach to corrective actions should be ' commended.
Continued effort on the licensee's part to improve . communications, MCB mimics and operating procedures will help prevent this type of event from recurring.
l VII. Licensee's Evaluation of PZR Cooldown and RCP Seals A.
PZR Cooldown , During the event the PZR' exceeded the TS cooldown limit of 200'F.in I one hour. TS 3.4.5.2.b requires that a written engineering evaluation l be preformed to demonstrate that continued plant operation is acceptable.
- l Pending review of the licensee's engineering evaluation, this is an-
OpenItem(50-457/90012-01(DRP)). B.
RCP Seals During the event RCS pressure went below the minimum required NPSH l and Number 1 Seal Differential Pressure values _with the RCPs pump ! operating.
Immediately after the event the operators observed - that the number 1 seal leakoff flow values were lower.than before.
' The operators then closed the number one seal leakoff isolation
valves.
l . As part of the licensee's evaluation of possible number one seal damage, the licensee measured the RCP " break away torque" on the C and 0 RCPs.
The values were higher than previously recorded but still below the maximum permitted by Westinghouse (W),-the pump manufacturer.
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->
. . - . , - <. . Higher breakaway torque values indicate increased resistance on the " control surfaces" of. pump components, particularly the sealing surfaces. This could indicate degradation of the number one seal.
The licensee initially determined that there was no conclusive.
evidence of seal _ damage; only indication of seal degradation..The seal package for the C and D RCPs were scheduled for replacement during the outage.
The licensee has committed to perform a; detailed.
, engineering: analysis af ter' the seals are removed.
Pending review of - the licensee's evaluation, this is'an Open Item (50-457/90012-02(DRP)).- VIII. Safety' Significance . A.
Immediate The unit was in the third day of its first refueling outage. The cure was still capable of producing a significant amount of. decay heat.
Throughout the event, the 2B RHR train was in the shutdown , cooling mode. The operators did not observe any indications of cavitation during the event. Additionally,: RVLIS indicated 100% throughout the event.
Based on the low initial RCS temperature of 135'F, RVLIS indicating 100% throughout the event and no indication of cavitation on 28 RHR-train, the inspector has determined that no loss of core. cooling occurred. Any temperature increases have been attributed to the hot pressurizer water mixing with RCS coolant. The immediate safety significance of the event was minimal.
B.
Other Conditions The worst case scenario is immediately after RHR is first placed - in shutdown cooling (350 F, 350 psig). Under these conditions, the licensee and inspectors anticipated voiding in the loops and reactor vessel.
Both RHR trains may become steam bound and become inoperable.
(While enacting this event on the Braidwood simulator, voiding in the vessel and loops was observed.) Assuming the operators took the same action to mitigate the event, with the exception of starting the 2A RHR pump, RCS inventory could be restored 'using the charging pumps.
With the plant configured per 2BwGP 100-5, Step 37, the operators could restore core cooling using 2Bw0A PRI-10, " Loss'of RH Cooling."
The inspectors have determined.that if the event occurred under worst case conditions, the procedures and equipment available to the operators coupled with their training would mitigate the event and minimize any challenges to reactor safety.
XI. Other Concerns During the course of the inspection two additional concerns were [ identified. The first concern is that the Sequence of Events Recorder (SER) was not printing durin the Main Control Board (MCB)g the event. The SER is. designed to print annunciator alarms as they are received.
This data is used to help determine the cause of an event.
The licensee
- - - - - - .
_ _ _ _..... _ ~ . - . , ' .. s stated that the SER has a history of. overloading during operation with-numerous MCB alarms (particularly during Modes 5 and 6 and reactor trips).
A larger capacity SER was previously scheduled for installation in Unit 2 and during the next refueling outage for Unit 1.
Byron Unit 1 SER has been replaced and Unit 2 is scheduled for installation during its next
refueling outage.
! The second concern relates to communications culture identified in ! Section III.C above. The control room practice which permits a unit NSO to take routine action without informing the.SCRE apparently encouraged the Unit 2 NS0 and extra NS0 to begin to mitigate the event without , ' informing the Center Desk NS0 or SCRE.
The Center Desk NSO was informed by the Unit 1 NSO after the RCPs were tripped off and the SCRE was informed by an EA when pressurizer level was being restored. The concern is why the NS0s did not inform the: Center Desk NSO/SCRE when a problem was identified.
Additionally, the Center Desk NSO went to investigate an abnormal ., condition on Unit I and did not inform the SCRE.
The licensee intends to address this concern as part of_the reevaluation of the scope of the HLA program.
X.
Conclusions , \\ The inspectors found that the sequence of events-developed by the i licensee using available records and documents including logs, computer.
i printouts, charts, and graphs, coupled with interviews of the control j room operators was a true and valid description of the event, j i The root cause of the event was personnel error due to a lack of ! communications between the SCRE and Unit 2 NSO. Two contributing { factors were the method of communications in the control room and procedural deficiencies in 2Bw0D RH-5 and RH-6.
The presence of.
the DJT trainee did not appear to have any influence on the event.
' Based on interviews, review of procedures, and discussions with-the-licensee, the operators appeared to have quickly diagnosed the event and took prompt, effective action to mitigate the event.
Even though the operators did not refer to procedures,-they used MCB indications, system knowledge, and good engineering practices to mitigate the event.
As discussed in Section V, it was concluded that the RCS water inventory balance closely accounted for the increase in RWST level during the event.
Additionally, there is no evidence of an unplanned release of radioactive ! material into the environment.
j As part of the licensee's immediate correction actions, procedures Bw0P RH-5 and RH-6 were modified to ensure valve RH8735 is locked shut when the RHR' train is not'in recirculation.
The HLA program was implemented to ensure increased levels of discussions occur prior to performing non-routine operations. The licensee's broad based ~ approach to long-term corrective action should be commended.
Continued effort on the licensee's part to improve communications, MCB mimics and operating procedures will help prevent this type of event from recurring.
.
_ _ _, _ _ _ _. _ _ _. _....... _. _.... - .-
, y . . ., . Based on the initial low RCS temperatu're_ and lack of indications of voiding'in the core, the immediate safety-significance of the event was minimal. Under other initial conditions, the procedures and equipment available coupled with the operators' training would mitigate the event and minimize any challenges to the RCS system.
During the inspection of an event in December 1989,_a concern was identified with the disparity between times shown in the written logs and verifiable computer data or records (Inspection Report 50-456/89030 (DRP)). A recommendation was made to the licensee to provide guidance to the operations personnel on the importance of accurate and timely log entries.
The operations personnel written logs that were reviewed-- during this event closely corresponded with computer data, charts, and other records.
Continued effort on the licensee's part will help maintain this improvement in written logs.
Finally, the licensee was encouraged to submit a voluntary Licensee Event Report (LER) as a vehicle for tracking RHR related events.- Additionally, since the entire nuclear industry has a heightened level of awareness of RHR related events, a courtesy ENS phone call would be appropriate for future events.
XI. Open Items i 0)en items are matters which have been discussed with the licensee, ! w11ch will be reviewed by the inspector and which involve some action
on the part of the NRC or licensee or both. Open items disclosed I during the inspection are discussed in Section VII.
p XII. Exit Interview The inspectors met with the licensee representatives denoted in Section I
during the inspection period and at the conclusion of the inspection on April 5, 1990.
The inspectors summarized the scope and results of the inspection and discussed the likely content of this inspection report.
l The licensee acknowledged the information and did not indicate that any t of the information disclosed during the inspection could be considered proprietary in nature.
i ! l !
q
. ,. _- - - - - . -. _. - _. -.. - -. . -- --. - $.:[s % Attachment 1 SuhP 396-3r4.
i < .- not.t.m n . l
l - . ' SPECIAL CPERATING CRDet
. Special Operating Order No. $0.ST-0039 nev 1 Effective Date.
03-22-90 ' Heightened Level of Awareness-(HLA) of l TIM Control Roon Activities
! The purpose of this memo is to enhance the concepts which exist I in BwAP 300-1 Conduct of Operations, regarding commincations-L and control of activities. It describe the minimum acceptable ' level of discussions that will occur prior-to any non-routine or ' . infrequent activity or event performed by Control Room Personnel.
Thit Heightened Level of Awareness will consist of - three parts:
1. As a minimum those evolutions or operations related.to the evolutions identified in Appendix A will have a Heightened Level of Awareness applied. - This list is no't all inclusive and can be modified or added to as necessary by the SE/SCRZ.
' 2. The SE/SCRE has the responsibility to identify events or evolutions which will require a Heightened Level of Awareness , by control room personnel.
This list of qualifying events will be presented at the Shift Briefing for those that are planned to occur during the shift.- This list will also be updated during the Shif t as new evolutions are identified.
The SCRE will maintain this list at his desk.
The SE and SCRE will include a discussion of the list during their ' turnover.
- 3. The NSO's are tasked with implementing a Heightened Level of i
, Awareness, on all activities on the SCRE's list. but are not ' limited to those events only.
The Heightened level of Awareness should be applied to any additional events or evolutions with_which the NSO is unfamiliar, uncomfortable or otherwise decides that this level of awareness is applicable, e . The Heightened Level of Awareness is defined as:
A minimum of two NSO's (unit and another) must discuss.
- the event or evolution in detail.
W=% APPROVED .Ng.i 4386-1- ' (0594Q/0021Q) , lllDW* /n 90192 , ' ,. . _.
. -... . .. .
_. _ _ _. _ _ _ _ . _ _.
.. _ _.
. _.. ___._.. . _ ' w 330 3g3 .q 1,"..%
- s Revision $1-8) - ^ .
- '
l - . ., SPECIAL CPsht!NG CADER
'
NT-29. anl 1 Special Operating Order No.
' 03-22-90 'affeetive Date .; titu:: ' Heightened Level of Awareness (HLA) of _ - .
Control Room Activities , ' There must be's clear understanding'of all the actions o and expected results among_those involved in the discussion.
. The discussion will include all actions that will be -
taken during the event or evolution.
, ) As the event is ongoing, a' continuous review or awareness
must be maintained ensuring expected results are observed.
in response to-actions taken.
I . This Heightened L'evel of Awareness is intended to provide the additional support an individual needs, to assure that infrequent or particularly difficult operations are well understood. prior to performance of the evolutions.
- w - .- /M/O ' ' G R. Masters.
Asst. Supt. Operating , Braidwood Station ' Attachment /
(Final) APPROVED M (0594Q/0021Q) .wan
9019z <
.. -. . - - -....... -. -- , - . . ... -
_. _ _. _. __,__ _ _ _ _ _ _ _ ._ _ _ _ _. I ! . . i *: ', ,.%- M N 373 a hevielea $1-f' , ,, ., ' .. ' SPE3AL CPERATIBIG OItDER
~ Special Operating Order llo,
~ affective Date ,
y_TLE:. APPDIDIX A HLA Activities
Operations involving manipulation of system isolations between - high and low pressure systems.
Operations involving CC heat exchangers especially when swapping - - , from unit to unit.
Any evolutions involving RH train such as recircing, placing in-l - i shutdown cooling, swapping train to train or filling or draining ' l the Reactor Cavity.
Upon recommenting procedures that are presently in progress but . - have required little or no operations over a significant time-period.
. Operations involving reduced RCS inventory.
- . Solid operations including placing the RCS in a solid condition - L and drawing a bubble in the pressurizer,
. I / (Final) APPROVED b.i f.3BB-1- , " (0594Q/0021Q) ll'AkW* 9019z l
. .
-. . - _ _ _ - - - - . _ _ _ _. _ . _.. _ - _ _ _ - - _ _ _. ,
^ BwAP 300-1 . , '
Revision 3 . ,,
- , , . < r p.
Fire Brigade.
' i A brigade of at least,5 members shall be maintained on-site at a11' times. The rire Brigade shall not include s
- those personnel. required in the control room.nor those , personnel.necessary.for the safe shutdown of'the unit.
, Only personnel who have satisfactorily completed the
required fire fightlug training shall be assigned as Fire Brigade' members. The members of the Fire Brigade shall be designated each shift.. Those personne)
- s assigned to the fire brigade shall be clean shaven to the estent necessary to obtain a seal with respiratory protection equipment.
g.
Operating Department equipment keys shall be controlled-in accordance with BwAP 330-5, Operations Equipment Key Control.
r.
Labellng of components or systems and posting of signs should only be performed in accordance with BwAP 330-4, Component Labeling or BwAP 330-8, Temporary Control Board and Annunciator Labeling.
s.
Communications.
i 1) Communications to operating personnel munt be w clear, formal and concise. These directions shall ' be given in such a manner that they are explicit and understandable. The use of slang, ambiguous statements and words that sound alike (i.e.' . Increase / decrease) should be avoided. All orders that involve the operation of plant equipment, ,
comples jobs or invnive safety will be repeated I back (either paraphrased or verbatim) so that both the director and the operator are satisfied that the orders are understood. Orders and repeat backs can be either oral or written. - Upon completion of a directed evolution, the operator will report back to the controlling station the exact action that he has taken. The controlling station will then return an acknowledgement of the report back to the operator. Whenever possible, the Individual ordering an action should verify that it has been carried out correctly by observing expected indication (Indication lights, meters, gauges, i ! etc.) and plant / system reaction.
2) The page announcing system should normally be used by operating personnel to announce emergencies, - unexpected events, to relay information regarding plant status and, where not possible by direct phone communications, to direct actions in the APPROVIQ plant.
APR 1(> 1390 .. woos omensmamse
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