IR 05000461/1997002
| ML20147F017 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 03/13/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20147F009 | List: |
| References | |
| 50-461-97-02, 50-461-97-2, NUDOCS 9703190358 | |
| Download: ML20147F017 (24) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION lil
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Docket No:
50-461 License No:
NPF-62 Report No:
50 461/97002(DRS)
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Licensee:
lilinois Power Company Facility:
Clinton Power Station Location:
Route 54 West Clinton, IL 61727 Dates:
January 6-10 and 31,1997 Inspectors:
James E. Foster, Senior EP Analyst i
Robert Jickling, EP Analyst i
Approved by:
James R. Creed, Chief, Plant Support Branch 1 Division of Reactor Safety l
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9703190358 970313 PDR ADOCK 05000161 G
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EXECUTIVE SUMMARY
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l Clinton Power Station
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NRC Inspection Report 50-461/97002 l
This inspection included a review of the Emergency Preparedness (EP) program, an aspect of Plant Support. This was an announced inspection conducted by regional emergency preparedness analysts.
The overall effectiveness of the EP program was good. However, performance was o
weak with respect to use of Emergency Plan elements during and after events.
The EP staff was less than aggressive in assessing Emergency Plan activations.
The inspectors were concerned that operators did not understand that a classifiable event in the Fmergency Plan is not the equivalent of a plant emergency in an operations procedure. Also, operator training on the Reactor Coolant System leak emergency action level was considered inadequate.
i Review of responses to events indicated that they had generally been classified in a e
timely and conservative manner. (Section P1.1)
Review of the September 5,1996, reactor recirculation pump seal failure event e
indicated that the initial Unusual Event classification was properly made, and notifications were made in a timely manner. (Section P1.3.b.2)
The notifications for the Unusual Event were made by the Shift Technical Advisor, e
which distracted him from his primary function of providing an overview of reactor safety. (Section P1.3.b.2)
Actual reactor coolant leak rate during the September 5,1996, event was e
determined to be in the range of 35 37 gallons per minute. However, this was not apparent to the operating shift at the time of the event, and was based on review following the event. Declaration of an Alert was not required by licensee procedures. (Section P1.3.b.4)
The Shift Supervisor during the September 5,1996, event had two opportunities to e
classify an Alert as a conservative decision. Neither of these options was exercised. (Section P1.3.b.5)
Emergency response facilities were in very good material condition with no e
significant problems or concerns identified. Several minor facility enhancements were noted. (Section P2.1.b)
A Violation was issued for failure to maintain the backup meteorological tower e
required by Section 3.2.3 of the Emergency Plan. Corrective action was lacking on this issue, which was over two years old. (Section P2.1.b)
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Audits and self-assassments of the emergency preparedness program were good.
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(Section P7.1)
The Updated Safety Analysis Report contained some outdated information and
would be enhanced by inclusion of some general overview information describing i
the Emergency Preparedness program. (Section P10)
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Report Detalla l
P1 Conduct of Emergency Preparednesa (EP) Activities
P1.1 Actual Emeroency Plan Activations (Generah
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a.
insoection Scoos (82701)
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The inspectors reviewed documentation packages regarding plant emergency response for emergency plan activations that occurred during 1996 and the procedure for documenting and reviewing actual plan activations. The general review is included below and a review of specific emergency plan activations involving a previous seal failure event and the September 5,1996, activation are
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covered in Sections P1.2 and P1.3.
b.
Findinas and Observations Actual activations of the Emergency Plan which took place in 1996 were:
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DATE CLASSIFICATION EVENT 4/9/96 Unusual Event Scram and plant systems do not function properly.
4/10/96 Unusual Event Heat up rates exceed requirements.
8/19/96 Alert Fire on safety system insulation.
9/5/96 Unusual Event Reactor coolant leakage in excess of limits.
- An Unusual Event was declared at 1:40 p.m. on April 9,1996, because operators did not have indication that all control rods had fully inserted following a reactor trip due to a failure of the Reserve Auxiliary Transformer. Alternate rod insertion was initiated as required. The operators were able to determine that all rods were in by verifying that the "all rods in" light for one division of Rod Control and Information System (RC&lS) was illuminated on the back panels. The Unusual Event was terminated at 8:05 p.m. when the licensee determined all control rods fully inserted on the reacter trip.
An Unusual Event was declared at 3:00 a.m. on April 10,1996, when the temperature of the reactor bottom head increased over 100(131 degrees) degrees per hour for a ten minute period. This caused entry into Technical Specification 3.4.11 and the decleration of an Unusual Event. The event was terminated at 7:45 p.m.
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An Alert was declared and terminated at 5:30 p.m. on August 19,1996, per l
Emergency Action Level (EAL) 10.1, " fire in the plant potentially affecting safety
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system," due to a small fire on the oil soaked insidation of the Reactor Core
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lsolation Cooling (RCIC) turbine. The fire was e:<tinguished within 15 minutes of ignition and the only observed damage was to the insulation. The licensee determined that the oil-soaked insulation was due to a loose cap on the outboard turbine bearing housing, which allowed the oil to leak onto the insulation. At 6:07 p.m. an inspection of the RCIC turbine revealed no visual damage.
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Records reviewed indicated that the above Alert classification was conservative and
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notifications had been made in a timely manner. This was considered a " pass-through event"; the alert was declared based on the guidance in Procedure EC-02,
" Emergency Classifications," Section 4.0, " Instructions," 4.1, " Station Emergency Director," 4.1.5. This indicated that if situations exist in which the plant is in an emergency condition only for a short period of time, and the situation clears before the appropriate emergency classification can be declared, then the appropriate emergency classification shall still be made in a timely manner. The documentation package for this event was adequate. A sequence of events and post fire critique were provided to the inspectors after the exit meeting.
Emergency Planning Standing Orders (EPSOs) contained Emergency Planning Guidelines (EPGs). EPG-13 provided guidance in the review of actual events and documentation of the reviews. None of the event packages reviewed met the guidance of EPG-13 in terms of documentation required or checklist completion.
Discussion with the EP staff indicated that since EPG-13 was guidance, not a
procedure. Strict adherence to detail was not required. The inspectors noted that in October 1996 the Root Cause Investigation Team expressed concern in their
analysis of Condition Report 1-96-09-72 that the large number of standing orders and questioned whether these could be undermining expectations to follow procedures.
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Real events provide opportunities to evaluate the EP program and correct any identified weaknesses. A procedure did not exist to perform EP reviews in a timely manner af ter actual events, nor to provide a summary report from the EP group to plant management describing the results of the review. Opportunities to evaluate the EP program had therefore been missed. Telephone discussion with the licensee EP staff on January 31,1997 indicated that, subsequent to the inspection, the EP staff had performed evaluations for events which had occurred in 1996. and provided a summary report to Clinton management.
The inspectors noted that a licensee team was tasked with a detailed review of the September 5,1996 event, without representation by the emergency planning group, although there were emergency planning aspects to the event.
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Conclusions Discussion and a review of documentation indicated that event classifications and notifications had been made properly and in a timely manner. Classification for the August 19,1996, Alert was particularly conservative. None of the event documentation and evaluation packages met the guidance of standing order EPG-13
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in terms of documentation required or checklist completion. Procedural weaknesses j
resulted in missed opportunities to evaluate the EP program.
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P1.2 Moview of the June 1.1989 Recirculation Pumo Seal Failure Event
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Inanaction Scone (82701)
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The inspector reviewed the records pertaining to the previous recirculation pump seal failure which occurred on June 1,1989. This review was performed to gain
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data and determine the relevance of the previous event to the September 5,1996 event.
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b.
Findinos and Observations Records indicated that on June 1,1989, with the reactor at approximately 40%
i power, the seals on the "B" recirculation pump completely failed. This event i
rapidly led to some 63-64 gallons per minute (gpm) leakage (believed to be
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unidentified leakage, as the documentation did not specify) and the declaration of
an Alert.
IHDet Event
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0045 15 gpm indicated on the Drywell Floor Drains Sump Flow Recorder.
- 0055 Unusual Event declared.
0056 64 gpm indicated on the Drywell Floor Drains Sump Flow Recorder 0100 Alert declared.
0110 Isolated Control Rod Drive water flow to recirculation pump "B".
0140 Downgrade classification to Unusual Event, leakage less than 5 gpm.
A critique of the event was held in accordance with Corporate Nuclear Procedure CNP 1.12, " investigations and Critiques." Critique OP-89-0053 documented the results. Part of the critique indicated that the classification procedure " requires that an alert be declared at the occurrence of an unidentified leak rate greater than 50 gpm." Event documentation focused on unidentified reactor coolant leakage rather than total reactor coolant leakage. No reference to identified leakage was found in any documents in the event package. The critique noted that logkeeping for the event was considered inadequate.
As noted in NRC Inspection Report 50 461/96010, a portion of the Clinton Power Station Updated Safety Analysis Report (CPS-USAR), Appendix D, the NRC Action o
Plan (NUREG-0660 as clarified by NUREG-0737), item II.K.3.25, "Effect of Loss of Alternating-Current Power on Pump Seals," notes that "Even in the case of both seal cooling systems failing, followed by extreme degradation of the pump seals, the primary coolant loss is analyzed to be less than 70 gallons per minute." While discussion indicated that operators received training on the above seal failure event, it was not clear that operators were also aware of this section of the USAR.
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While both this and the September 5,1996 events involved the seals on
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recirculation pump "B", there were significant differences in the two events, one being a rapid catastrophic failure while the loop was not isolated, and the other i
being a developing partial failure while the loop maintenance valves were closed
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c.
Conclusiong Records indicated that the 1989 pump seal failure event was promptly and properly classified as an Alert. Event documentation focused on unidentified leakage; identified leakage was not mentioned. Operators received training on this event.
The USAR indicated that seal failure can result in under 70 gpm coolant loss, but it was undone whether operators were aware of this information. While there were similarities to the September 5,1996 pump seal failure and this event (the same pump's seal was involved), there were differences between the events.
P.1.3 Review of Sootember 5,1996 Recirculation Pumo Seal Failure Event a.
Insoection Scone (82701)
A detailed review of the September 5,1996 event from an emergency preparedness perspective was performed. This review included document and procedure reviews, review of logs, event critiques, assessment reports, inspection of equipment in the control room, interview of the i.eak Detection System Engineer, discussion with NRC personnel present during and after the event, and review of NRC reports. A detailed review of the operational aspects of this event was contained in NRC Inspection Report No. 50-461/96010.
b.
Findinas and Observations b.1 Overview of Event On September 5,1996, the licensee began to place the reactor in single loop operations by isolating the "B" reactor recirculation (RR) loop. NRC resident inspectors were present in the control room to observe this evolution. It was believed that sealleakage from the "B" RR pump was causing unidentified reactor coolant leakage to increase; isolating the pump might reduce the leakage. Power was reduced in preparation for securing the "B" RR pump. At 8:09 p.m., the "B" RR pump was secured and the pump discharge valve closed. Approximately forty-five minutes later, unidentified drywell leakage exceeded the technical specification (TS) limit. Operations personnel declared an Unusual Event for unidentified leakage greater than 5 gpm.
Operations personnel then shut the pump suction valve. Since it was believed that sealinjection flow to the RR seals was contributing to the amount of unidentified leakage, a decision was made to isolate sealinjection flow. Twenty-three minutes later the pump seal failed. The drywell floor drain leak rate for unidentified leakage was later calculated by the licensee to have peaked at 38 gpm. Following the loop
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depressurization, unidentified leakage dec eased, and stabilized around 10 opm.
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The Unusual Event was terminated at 9:50 p.m., on September 6,1996, when unidentified leakage consistently remained below the TS limit of 5 gpm.
An event time line from an emergency preparedness perspective was developed, and is attached as Attachment A.
b.2 Initial Unusual Event Classification At approximately 9:55 p.m., on September 5,1996, the plant exceeded Technical Specification (TS) 3.4.5 limit for unidentified leakage.2. 5.0 gpm. Unidentified leakage displayed by the Leak Detection (LD-27) system was 5.52 gpm. At 9:10 p.m., the Shift Supervisor properly declared an Unusual Event per the guidance in Procedure EC-02,4.1. The resident inspectors observed that proper notifications of the Unusual Event were made on site and off site to state, local and NRC officials.
Documentation also indicated that the required notifications were made in a timely manner.
The Shift Technical Advisor (STA) agreed to assist the crew by performing the notifications required for the Unusual Event declaration. Utilization of the STA to make emergency notifications was inappropriate because it took him away from his assigned safety responsibilities. CPS 1401.01, " Conduct of Operations " Step 8.1.6.2a, states that "during off normal conditions one of the primary duties of the STA is to assist the shift supervisor in the identification of the proper emergency action level classification." Per Procedure EC-01, " CPS emergency response organization and staffing," notifications were to be performed by the Assistant Shift i
Supervisor, with the STA providing technical support to reactor operations. The failure to properly evaluate conditions both during and following the event, specifically within the first 30 minutes following the seal failure, in order to monitor possible entry into an emergency classification condition was an apparent violation of 10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures and Drawings,"
previously cited in NRC inspection Report 50-461/96010.
At 9:30 p.m., operations personnel entered Step 8.2.4.3 of CPS 3302.01 which delineated the actions for isolating an idle Reactor Recirculation (RR) loop under
" emergency" conditions. This was an improper use of the earlier emergency classification of the Unusual Event. Emergency classifications are intended to alert onsite and offsite personnel to ongoing conditions, and activate onsite and offsite emergency organizations as necessary for the event classification. Operational
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procedures may contain prompts to review the classification scheme to determine if an emergency action level has been exceeded, but event classifications are not
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intended to drive operational procedure actions. This is especially true at the
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Unusual Event level, where events only indicate a notential for degradation of the level of safety of the plant.
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b.3 Moview of Subsenuant Events
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The inspectors noted that the reactor operator log indicated a catastrophic seal failure and the Shift Supervisor log indicated a gross seal failure at 10:22 p.m. A larger leak rate increase than that displayed via LD 27 should have been anticipated. As the unidentified leak rate exceeded 7.99 gpm, the drywell floor drain sump flow rate computer point went to " white data," meaning that it was no
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longer a dependable reading.
Discussion, review of records, and intervlaws allindicate it was not well known to the operations staff that the 50 gpm leak rate specified as the trigger point for the Alert emergency classification was a combination of identified and unidentified leakage.
The sequence of events indicates that for a period of time shift personnel were either unaware of the actual total leak rate due to lack of familiarity with the LD-27 system limitations (the system was clamped at 8 gpm) or assumed identified leak rate data, based on historical data, without verifying same. There was no indication of tracking of identified leak rates during the event. Without cognizance of the total leak rate, there was a distinct potential that the EAL trigger value could be exceeded without recognition.
It was not until a relief STA arrived that leakage rates were properly calculated, b.4 Leak rate Determination The inspectors interviewed the system engineer for the leak detection system. He indicated that leakage from the recirculation pump seals was well known to operations personnel and plant management prior to the event, and a computer spreadsheet had been developed to calculate and track leakage. Pump run times and other data could be obtained (pump run times from the alarm typer) and input into the spreadsheet by the STA. The spreadsheet then provided two methods of calculating leakage in gpm using fill time and pump run time. The system engineer said that he had recommended the use of gpm using pump run time because the result has fewer variables and is more likely to be near the correct value, t
The system engineer stated that the on-shift STA called him at home to confirm i
that the reading from LD-27 was clamped at 8 gpm. He volunteered to come to the plant and did so even though his presence in the control room was not requested by the STA. He stated that he was consulted, watched calculations, and would have advised the crew if he thought they were near 50 gpm totalleak rate. He stated that they were tracking unidentified leak rate, and did not observe any indications that the identified leak rate was being tracked. The NRC resident i
inspectors later verified that identified leakage calculations were not available.
The system engineer had retained his leak rate spreadsheet and a printout was provided to the inspectors. The following values were obtained from the
spreadsheet for unidentified leakage:
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TIME GPM (FILL TIME)
GPM (RUN TIME)
2159 6.533 6.532 2222 6.035 5.618
2230 21.015 24.008 t
2237 23.498 38.112
2246 18.918 20.812
2257 15.072 15.730
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2309 12.170 12.530 2322 11.079 14.063 i
The following values were obtained from the spreadsheet for. identified leakage:
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TIME GPM (FILL TIME)
GPM (RUN TIME)
2159 7.306 6.270 2222 7.425 6.127 2230 10.020 8.600 2237 8.093 8.679 The system engineer indicated that the maximum calculated leakrate at 10:37 p.m.,
was in error due to the system not recognizing momentary pump stoppage (this would extend the perceived length of pump run time). Therefore he believed that the correct maximum leakrate would have been 24-27 gpm rather than 38 spm.
He believed that the identified leakrate was in the range of 8-10 gpm, with a possible maximum of 14 gpm. This would make maximum total reactor coolant leakrate during the event in the range of 32-41 gpm, with a maximum value of 3E 37 nom being most likely. The highest calculated leaktate estimates showed the leakrate to be below the trigger point for declaring an Alert. Therefore, an Alert declaration was not required by Procedure EC-02 which required an Alert be declared when total reactor coolant leakage exceeds 50 gpm. However, the leakage rate of 35-37 gpm was not apparent to the operating shift at the time of the event, and is based on review following the event.
b.5 Classification Ootions At approximately 10:48 p.m., with the information that the leakrate had peaked at approximately 38 gpm, and identified leakage in the (unverified) range of 6-10 gpm, maximum total reactor coolant leakage was approximately 48 gpm, near the 50 gpm trigger point for declaring an Alert. If it was suspected that the instru-
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mentation uncertainties had caused actual total reactor coolant leakrate to exceed 50 gpm, the Alert could have been conservatively declared under the guidance in Procedure EC-02, Section 4.0, " Instructions," 4.1, " Station Emergency Director,"
4.1.5. This procedure indicated that if situations exist where the plant is in an emergency condition only for a short period of time, and the situation clears before the appropriate emergency classification can be declared, the appropriate emergency classification shall still be made in a timely manner.
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- The Shift Supervisor also had the option of classifying the event as an Alert on bis
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own judgement, under the guidance in Procedure EC-02, Category 13, "Other Hazardous Conditions," Item 13.6, " Judgement of the Individual with Command Authority." The procedure provides the option to classify an Alert if "other plant conditions exist that warrant precautionary activation of the Technical Support Center (TSC) and placing the Emergency Operations Facility (EOF and other key emergency response personnel) on standby." Considering the leak rate instrumentation difficulties being experienced, this would have been a conservative, i
supportable decision, b.6 EP Staff Review of Event
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The EP staff, by memorandum dated December 12,1996, more than three months after the event, documented an evaluation of the requirements and actions during this event. EP concluded that an Alert was not required during the September 5, 1996 seal leak event, and "although primary indication for the leak was lost the shift supervisor was exercising his best engineering judgement of the sequence of events" in this determination. The memorandum did not provide a sequence of events or supporting documentation, and did not discuss the option to make the declaration based upon a more conservative judgement of overall plant conditions.
Evaluation of the adequacy of the classification and related notifications should have been made promptly after the event and should have included supporting information.
Discussion with the emergency planning staff indicated that they had informally reviewed this event in a timely fashion, but had not formally documented the results of their review nor had they been involved in the assessment team activities.
b.7 Subseouent Licensee Actions Revision 8 to CPS No. 4001.01, " Reactor Coolant System Leakage," dated September 11,1996 " clarified EC-02 expectations for the use of identified and unidentified leakage in determining EC-02 ALERT classification." This clarification i
added a note to step 4.0, " Subsequent Actions," 4.1, " Classify the event per EC-
02, Emergency Classifications."
The Emergency Action Level (EAL) provided in EC-02 added an asterisk and note to clarify the terms " leak rate" or " total leak rate."
A Condition Report was written regarding the failure to track total drywell leakage (CR 1-96-09-137, dated September 20,1996).
Telephone discussion with licensee emergency planning personnel on January 31, 1997, indicated that emergency classification training had been revised to clarify that if a trigger point of an Emergency Action Levelis being approached, or is thought to have been approached due to instrument difficulties, then a supportable and conservative course is to make the classification at the level being approached.
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c.
Conclusions The Unusual Event was properly classified at the appropriate time and notifications were made in a timely manner. The resident inspectors observed that proper notifications of the Unusual Event were made on site and off site to state, local and NRC offidels. Inspector review of applicable emergency response procedures determined that the declaration of an Unusual Event was in accordance with the procedures.
Operators knew the seal had failed, were aware that leakage from a failed seal could exceed 50 gpm, and did not know the actualleak rate. Operations personnel did not fully understand what leakage conditions constituted entry into an Alert classification.
The licent.se properly exercised the emergency response procedures to declare an Unusual Event when unidentified reactor coolant leakage exceeded five gpm.
Subsequently, given the known seal condition, the lack of accurate leak rate data, and the potential leakrate from a failed seal, declaration of an Alert would have been a conservative, supportable decision.
l Noncompliance items related to this event were not associated with this report.
Potentialitems of noncompliance were issued in NRC Inspection Report 50-l 461/96010.
P2 Status of EP Facilities, Equipment, and Resources P2.1 Material Condition of EP Facilities a.
Insoection Scone (82701)
The inspectors toured the Control Room, Technical Support Center (TSC),
Operational Support Center (OSC), and Emergency Operations Facility (EOF) and Backup Emergency Operations Facility (BEOF) to assess their material condition, in each facility, associated equipment, procedures, telephones, and instrumentation were inspected.
b.
Findinas and Observations Each facility was very well maintained in a state of operational readiness. Minor enhancements, such as new or enhanced status boards, were noted in several facilities.
A raised floor modification had been started in the TSC, correcting a long-standing problem of a cable tripping hazard. Conduits involved with a partially completed modification made the previous tripping hazard worse. The EP staff was unable to determine if the modification would be completed. Modifications had been made to the Area Radiation and Process Radiation (AR/PR) monitoring system which was intended to improve the reliability of the system in both the main control room and
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the TSC, where the system is utilized for accident analysis and dose projection.
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Sound-deadening material had been added to the facility since the last rout,ne i
inspection.
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The EOF ten-mile emergency planning zone map had been replaced with a current l
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Safety (IDNS). Excellent position-specific nuclear emergency response manuals.
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j were present. A digital level indication system had been installed on the EOF i
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decontamination shower receiving tank, correcting a long standing reliability problem. Licensee personnel advised that speakerphones in the emergency l
manager conference room now allow Joint Public Information Center personnel to i
hear conference room meeting discussions.
The OSC and BEOF were in excellent material condition. A very minor problem was
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noted with an update to procedures maintained in the BEOF, and this was quickly corrected by the EP staff. Revisions to the BEOF included new furniture, and
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Each facility was well-maintained and in a very good operational state of operational readiness. The inspector requested testing of numerous pieces of equipment (survey equipment, computers); no significant problems were evident. Copies of the Emergency Plan, Emergency Plan Implementing Procedures and appropriate forms were present in each facility, as required. Minor enhancements intended to improve performance were noted in various facilities.
The possibility of eliminating the requirement for a backup meteorological tower was discussed, and data for the towers reliability was provided for review.
Portions of the 1996 annual audit (and two previous audits) addressed the material condition of the Meteorological Monitoring System, noting that it had been meeting the 90% information reliability criteria, but had numerous failures caused by icing and lightning, due to outdated equipment. Audit 038-96-08 noted that "new equipment, which would aid in preventing these failures, has been on site for two years. A modification to replace the outdated equipment has been approved for j
1997." Discussion with EP personnel indicated that the primary meteorological
_ tower upgrade was still pending, but the tower had approximately 95% data
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recovery. EP staff indicated that backup meteorological information is available from the Weather Service located in Lincoln, Illinois, approximately 25 miles from the site.
Discussion with NRC personnel on January 30,1997 indicated that a Condition Report (1 94-10-002) dated September 30,1994, Issued October 3,1994, indicated that the backup meteorological tower wind speed and wind direction sensors were inoperable. The Clinton Power Station Emergency Plan, Revision 10, Section 3.0, " Emergency Response Facilities and Equipment," Paragraph 3.2.3,
" Meteorological Monitoring System," states that "A backup meteorological tower is located at the CPS site and is instrumented at the 33 foot (10 meter) level. The meteorological parameters measured are wind speed and wind direction. Sigma Theta is calculated from the changes in wind direction. All three of these variables i
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are available in the Technical Support Center and are read from a line printer."
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' Contrary to the above, the wind speed and wind direction sensors have been unavailable since October 1994. This is a violation of NRC requirements (50-461/97002-01).
c.
Conclusions Overall, emergency response facilities were in very good material condition with few problems or concerns identified. Several very minor problems were noted by the inspectors and immediately corrected. Several minor facility enhancements were noted. A violation for failure to maintain the backup meteorological tower was identified.
P3 EP Procedures and Documentation a.
insnection Scoon (82701)
The inspector reviewed a small selection of licensee emergency procedures and emergency plan implementing procedures (EPIPs),
b.
Findinns and Observations Discussion with licensee personnel indicated that the classification scheme developed by Nuclear Management and Resources Council (NUMARC) would not be adopted. The Protective Action Recommendation generation flowchart in procedure RA-02 had been revised to match that in current NRC guidance, c.
Conclusions No problems were identified in the procedures and documentation reviewed.
P5 Staff Training and Qualification in EP a.
Insnection Scone (82701)
The inspectors discussed and reviewed available documentation relative to training regarding classifying events according to reactor coolant leak rates. An inspector had previously observed pertinent training.
b.
Findinos and Observations Licensee training staff indicated that initial training covered all of the Emergency Action Levels (EALs), and operator requalification covered select EALs. Discussion of whether EC-02 EAL 4.1 included both identified and unidentified reactor coolant system (RCS) leakage for the Alert classification indicated that it was unknown
- whether the question had come up during training sessions. It was indicated that a significant amount of training had been given to operatore relative the June 1,1989 recirculation pump seal failure event.
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As documented in Report 50 461/96010, on September 6,1996 an NRC inspector
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noted a lack of knowledge by operations personnel concerning what constituted RCS leakage greater than 50 gpm as listed in the EALs for an Alert status. At the request of the inspector, the training department reviewed the lesson plans.
L Neither initial operator training nor requalification training taught the operator that l~
identified leakage and unidentified leakage must be added together to determine if an Alert condition existed.
- Discussion with training personnel indicated that exercises and drills have utilized recirculation pump seal failure scenarios; however, these have utilized rapid, catastrophic failure quickly resulting in over 50 gpm unidentified leakage.
l Discussion indicated that training following the September 5,1996 event had
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clarified the definition of total RCS leakage. As noted in Section P1.3.b.7, EAL 4.1 j
had been clarified by note as to a definition of total RCS leakage.
c.
Conclusions Training concerning the Alert entry conditions for RCS leakage had been inadequate. This was rectified by subsequent training and classification procedure clarification.
P6 EP Organization and Administration 182701)
There have been no changes in overall organization of the EP function since the last inspection. There have been several changes in the individuals occupying upper
management positions. Several times over the past years, Clinton has exchanged
drill observers or personnel to participate on " mock NRC" teams.
P7 Quality Assurance in EP Activities P7.1 Audits (82701)
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a.
Inanection Scone (82701)
The inspector reviewed documentation of emergency preparedness audit and self-
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assessment activities which had taken place since the last routine inspection.
b.
Findinas and Observations The inspector reviewed Nuclear Assessment Audit Report 038-96-08, " Emergency Response / Emergency Operating Procedures," dued June 25,1996. This audit was performed by six individuals during May 13-31,1996. The audit concluded that the Emergency Response and Emergency Operating Procedure Programs are effectively implemented. No audit findings were issued, but one Condition Report was issued regarding evaluation of letters of-agreement. The audit also contained two recommendations and five observations. Evaluation of the interface with offsite authorities was effective and consisted of interviews conducted utilizing a
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standardized list of questions. No problems were identified as a result of the interviews. The audit concluded that the established interfaces with State and local govemment agencies were considered effective.
Also reviewed was the EP staff annual review of the program required by Corporate Nuclear Procedure 4.03, dated October 31,1996, which was considerably more comprehensive than the last 1995 review. The NRC EP inspection modules were utilized as a portion of the review.
c.
Conclusions Audits and surveillances of the emergency preparedness program were very good.
The annual audit per 50.54(t) was well-detailed and comprehensive.
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P8 Miscellaneous EP lasues P9 Temporary instruction (TI) 2515/134 On-shift Dose Assessment a,
inanection Scone The inspector discussed riose assessment capability and provisions with licensee personnel, reviewed the Emergency Plan and Emergency Plan implementing Procedures (EPIPs), and inspected equipment in response facilities. Training for dose assessment was also discussed, i
b.
Findinas and Observations The Emergency Plan did not contain specific commitments for on-shift dose assessment. Form EC-01, Step B.1 provides for dose assessment for either an Alert, Site Area Emergency or General Emergency. Emergency Plan Implementing Procedures RA-01 " Manual Radiological Dose Assessment," and Procedure RA-16,
" Computerized Radiological Dose Assessment," provided methodology for the dose calculation. The procedure currently does not include a requirement to utilize the actual meteorological stability class, but utilizes default adverse weather conditions meant to be conservative. Licensee personnel had previously been advised that this was unacceptable, and a commitment was made to revise the procedure following the current refueling outage.
Discussion with licensee personnel indicated that adequate training was provided.
Licensee personnel committed to modifying the Emergency Plan and appropriate sections of procedures addressing dose assessment capability following the current refueling outage. This will be an inspection Followup Item (IFI) (50-461/97002-02).
The inspectors observed the use of a revised dose assessment program MESOREM which contains an upgraded dispersion model for more precise dose assessment.
Excellent cooperation on dose assessment issues existed with the State of lilinois,
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with all Illinois plants using the same dose assessment model. The lilinois
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Department of Nuclear Safety provided the computer programming for the dose i
assessment module.
c.
Conclusions
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The Emergency Plan did not clearly commit to a capability for on-shift dose assessment. However, EPIPS do contain such provisions. Needed equipment and a
i personnel training were provided. The licensee indicated that clarification of their commitments would be added where necessary. Plan / procedures addressing on-
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shift dose assessment capability remain on hold until the end of the current outage.
- With these clarifications, all of the acceptance criteria for the Tl will be met and this
Tl is closed. Documentation as to these findings is attached as Attachment B.
Pio Review of USAR Commitments I
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a.
insoection Scone I
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l A discovery of a licensee operating their facility in a manner contrary to the Updated Safety Analysis Report (USAR) description highlighted the need for a i
. special focused review that compares plant practices, procedures, and parameters j
to the USAR descriptions. While performing the inspections discussed in this j
report, the inspectors reviewed the applicable portions of the USAR that related to Emergency Preparedness.
,
b.
Observations and Findinos
j Section 13.3 on Page 13.3-1 of the CPS-USAR, " Emergency Planning," covered EP.
This section noted that a separate Clinton Station Emergency Plan is maintained,
having been developed following the guidance in NUREG-0654.
,
The remainder of the section addressed both dose assessment and remote system
interrogation, and appeared neither accurate nor current. The section indicated, in
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inpart, that "A digital computer system is being implemented at Clinton... offsite dose calculations will be performed during both normal and accident release conditions. The Clinton dose assessment computer did not, and was not intended to perform offsite dose calculations during both normal and accident release
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i conditions. This was another example where Illinois Power had not corrected differences between the UFSAR and plant conditions. Discussion with licensee personnel indicated that this section would be deleted.
The inspector discussed whether a general overview of the EP program should be j
placed in the USAR in place of the brief reference to the Emergency Plan. The USAR was not meant to contain a full Site Emergency Plan, nor be updated each time an Emergency Plan change was made. Internal NRC guidance was not specific as to the acceptable level of detail for this section of the USAR. Subsequent discussion with the Emergency Planning staff on January 31,1997, indicated that an overview of EP would be placed in this section.
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Conclusions The USAR would be enhanced by inclusion of general overview information as to the Emergency Preparedness program. Portions of Section 12.3 were being revised to reflect the current requirements.
X1 Exit Meeting Summary The inspector presented the inspection results to members of licensee management at the conclusion of the inepecoon on January 10,1997. The licensee acknowledged the findings presented.
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The inspector 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 J. Cook, Sr. Vice President
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W. Connell, Vice President P. Yocum, Manager - Clinton Power Station D. Thompson, Manager - Nuclear Station Engineering Department R. Phares, Manager - Nuclear Assessment J. Palchak, Manager - Nuclear Training and Support M. Lyon, Assistant Plant / Manager - Operations R. Bedford, Asst. Director - Operations P. Telethurst, Director - Licensing G. Baker, Director - Nuclear Assessment i
D. Walters, Nuclear Assessor W. liiff, Recovery Program Director R. Frantz, Sr. Licensing Engineer t'. Turner, Nuclear Program Controller M. Stickney, Supervisor - Regulatory Interface W. Evans, Supervisor - Emergency Response W. Yaroz, Supery!snr - Emergency Exercises J. Wemlinger, Director - Operations Training & Emergency Response INSPECTION PROCEDURES USED l,
IP 82701 Operational Status of the Emergency Preparedness Program Tl 2515/134 Temporary Instruction, On-shift Dose Assessment ITEMS OPENED 50-461/97002-01 VIO Failure to maintain backup meteorological tower 50-461/97002-02 IFl Changes need to Emergency Plan and EPIPs regarding dose assessment l
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LIST OF ACRONYMS USED l
AR/PR Area Radiation / Process Radiation BEOF Backup Emergency Operations Facility CFR Code of Federal Regulations CPS Clinton Power Station l
CRD Control Rod Drive l
DRP Division of Reactor Projects l
DRS Division of Reactor Safety EAL Emergency Action Level i
EOP Emergency Operating Procedure l
EOF Emergency Operations Facility EPIP Emergency Plan implementing Procedure EPPSO Emergency Planning Standing Order ERDS Emergency Response Data System ERO Emergency Response Organization
FRMAC Federal Radiological Assessment and Monitoring Center GPM Gallons Per Minute IDNS lilinois Department of Nuclear Safety
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IFl Inspector Follow-up item IP lilinois Power IR
Inspection Report
JPIC
Joint Public Information Center
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LCO
Limiting Condition for Operation
Unidentified Drywell Leakage Monitor based on sump run times and
measured pump flow rates
NAD
Nuclear Assessment Department
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Number
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NPF
Nuclear Power Facility
NRC
Nuclear Regulatory Commission
Nuclear Regulatory Commission document
Operational Support Center
Public Document Room
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Reacto,r Coolant System.
HCIC
Reactor Core Isolation Cooling
RC&lS
Rod Control and Information System
Reactor Feed Pump
Reactor Operator
Reactor Recirculation
Ti
Temporary Instruction
TS
Technical Specifications
Updated Safety Analysis Report
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CHRONOLOGY OF CLINTON UNUSUAL EVENT ATTACHMENT A
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APPROX! MATE
IlME
BRIEF EVENT DESCRIPTION
9/5/96
1712 Drywell floor leakage (DWFL) 3.96 gpm (via LD-027)
1800 Single loop operations crew brief completed. Items discussed included actions to
take if sealleakage increased, actions for reactor scram, and actions for feedwater
pump problems.
1805 Commenced reducing power with control rods.
1944 Completed reduction of power to 81%.
2001 Began closing Loop B flow control valve in accordance with CPS 3302.01 " Reactor
Recirculation." Reactor power 80%.
2009 Shutdown the "B" Reactor Recirculation pump. Reactor power 58%.
2010 Loop B discharge valve closed.
2015 RR Loop B suction for reactor water cleanup closed.
2030 Shut RR seal staging flow valve. Attempted to increase the RR loop cooldown by
directing this portion of sealinjection water into the loop. This action was
performed out of sequence from the procedure. It was to be performed only in the
event of an emergency condition (such as a significant system / seal leak). DWFL
4.22 gpm.
2055 Exceeded TS 3.4.5 lirnit for unidentified leakage.2. 5.0 gpm. Began 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> action
statement to reduce leakage (CPS 4001.01 Reactor Coolant Leakage). DWFL 5.52
gpm.
2110 Declared UNUSUAL EVENT per EC-02,4.1
2118 DWFL 5.86 gpm
2125 State IEMA and IDNS notified of Unusual Event declaration.
2130 Closed RR suction valve in accordance with " emergency" isolation section of the
procedure. Completed isolation of RR pump B with the exception of the seal
injection isolation valv. _ -
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APPROXIMATE
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BRIEF EVENT DESCRIPTION
2144 Entered TS 3.4.5 for unidentified leakage increasing > 2 gpm in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
NRC notification complete for Unusual Event. DWFL 5.94 gpm
2159 Shut control rod drive (CRD) sealin}ection path. This completed loop isolation
after determining that the loop was not going to depressurize through the seal with
CRD flow lined up. This decision further jeopardized the seal package. This step
was performed contrary to procedure. The procedure stated that prior to isolating
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CRD injection flow, one of two conditions had to be met. Either the loop was to be
cooled down to less than 250*F, or, the RR seals had already failed, (which was
defirsd as the loop depressurized to approximately drywell pressure.)
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2200 DWFL 6.13 gpm
2207 Security reported that all ERO notifications were complete.
2208 Plant manager arrives due to Unusual Event.
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2220 Reactor operator informed resident (required by procedure) that seal temperature
had reached 1600F.
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2222 Sudden failure of Reactor Recirculation pump "B" seal. Seal pressure decreased
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rapidly from 950 to 280 psi within a few seconds. Containment was evacuated per
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CPS 4001.01. Drywell pressure started to increase. As leakage exceeded 7.99 the
LD-027 instrument locked at 7.99 gpm and did not go higher. Operators were not
aware that the instrument locked at 7.99 gpm. The operating procedure for leak
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detection and the off normal procedure for abnormal reactor coolant leakage did not
address this fact. The drywell floor drain sump flow rate computer point went to
" white data" (no longer a dependable reading). Remaining leakage data is from
STA calculations. DWFL 5.6 (reconstructed after 10:55 p.m.)
2230 DWFL 24.0 gpm (reconstructed, NRC Calc 19.9). The licensee calculation for
DWFL did not take credit for the time period that the sump pump was running.
NRC leak rate calculations are therefore lower.
2223 On-shift STA calls LD System Engineer at home.
2237 DWFL 38.1 gpm (reconstructed, NRC calc 31.1) Discussion with system engineer
and review of calculations suggests actual unidentified leakrate value at this time
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mey be closer to 27 gpm). LD System Engineer calls, advises LD-27 clamps at 8
gpm.
2246 DWFL 20.8 gpm (reconstructed, NRC calc 17.6)
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2248 Oncoming STA entered the Control Room and discovered that the floor drain flow
rate indication (LD-027) was clamped at 8 gpm.
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APPROXIMATE
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BRIEF EVENT DESCRIPTION
2250 STA advises Shift Supervisor that leakrate is decreasing.
-2255 Entered actions for leak detection instrumentation per technical specifications. On-
shift STA began performing manualleakage calculations.
2257 DWFL 15.7 gpm (NRC Calc 13.7)
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2300 STA reports that spreadsheet calculation shows 15.7 gpm, down from 38.1 gpm.
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Totalleakrate assumed to have been approximately 48 gpm. Identified leakrate not
verified.
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2337 DWFL 16.5 gpm (NRC Calc 14.8)
2345 System Lead Engineer arrived onsite.
2351 DWFL 11.6 gpm
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0006 DWFL 10.8 opm
0055 Exceeded 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LCO for unidentified leakage. Entered 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> reactor shutdown
!
statement. Control room operators directed to review procedures for normal
reactor shutdown.
0118 DWFL 10.6 gpm
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0230 Power decrease commenced.
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0700 Reactor power 23%
1106 Turbine off of the grid.
1206 Manual reactor scram as part of normal shutdown, enter mode 3.
1543 DWFL 5.5 gpm
1645 DWFL 2.2 gpm
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2150 Exited Unusual Event for excessive unidentified leakage.
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FORM FOR DOCUMENTATION OF ON-SHIFT DOSE ASSESSMENT CAPABluTY
ATTACHMENT B
Clinton Power SW/ Unit 1/50-461
lainois Power Comoony
10/11/96-
SITE / UNIT / DOCKET #s
LICENSEE
DATE
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..4.01:..E DOSE ASSESSMENT COMMITNENT IN EMERGENCY PLAN
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Acceptance Critene
Person (s) Contacted
Position Title (s)
Plan Section
Revenson No.
Meets acceptance
I
(Refer to page 1 of this Appendix for
conesining
and Does
critorie?
further detail on the acceptance entena)
commitment
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William Yarosz
EP Supervssor
None - Plan revision
Rev.10,
Not yet
Section 4.01 frem 1
being initiated to add
10/95
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Emergency Plan contains commitment for on-
this comrmtment
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shift dose assessment capability.
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William Yarosz
EP Supervisor
Subsections 3.2.2;
Rev.10,
Yes
Section 4.01 Item 2
4.2.2; 4.3.2.3
10/95
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Emergency Plan contains commitment for
backup dose assessment capability.
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04.02 ON-SHIFT DOSE ASSESSMENT EMERGENCY Pt.AN iMPLEllENTING PROCEDUfE
k
Person (s) contacted
Posstion Title (s)
Procedure / Indication
Revesson No.
Meets acceptance
j
and Dato
criteria?
Wil!iam Yarosz
EP Supervisor
EC-01, Form 1, Step
Rev.2,
No, procedure
Section 4.02 Item 1
B.1 for Alert, SAE,
3/96
does not include
Procedure initiates dose assessment
and GE events
actual stability
class.
William Yarosz
EP Supervisor -
Rev. 4, 3/93;
Yes
Section 4.02 Item 2
Rev. 3, 5/95
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Indications initiate dose assessment
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William Yarosz
EP Supervisor
Rev. 4, 3/93; Yes
Section 4.02 Item 3
Rev. 3, S/95
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Procedure for performing dose assessment
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available.
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' 04.03 i ON-SHfFT DOSE ASSESSMENT TRAINING
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- Person (s) contacted
Posetion Titlo(s)
Personnel Trained
Meets acceptance
(Title /s)
criteria?
William Yarosz
EP Supervisor
RP Shift Supervisors
n/a
Yes
}
Section 4.03 Item 1
& RP Techs; about 30
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On-shift Personnel trained for dose assessment
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Inanector: Thomas Ploski Reaion Ill. DRS. Plant haart Br.1
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