IR 05000285/1999003
| ML20206U638 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/18/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20206U637 | List: |
| References | |
| 50-285-99-03, 50-285-99-3, NUDOCS 9905250281 | |
| Download: ML20206U638 (21) | |
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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION j
REGION IV
Docket No.:
50-285 License No.:
DPR-40 l
Report No.:
50-285/99-03 Licensee:
Omaha Public Power District Facility:
Fort Calhoun station l
Location:
Fort Calhoun Station FC-2-4 Adm.,
i P.O. Box 399, Hwy. 75 - North of Fort Calhoun
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Fort Calhoun, Nebraska j
Dates:
March 14 through April 24,1999 Inspectors:
W. C. Walker, Senior Resident inspector V. G. Gaddy, Resident inspector G. A. Pick, Senior Project Engineer P. J. Elkmann, Emergency Preparedness Analyst Approved By:
D. N. Graves, Chief, Project Branch B ATTACHMENT:
SupplementalInformation
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9905250281 990518 PDR ADOCK 05000285
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EXECUTIVE SUMMARY Fort Calhoun Station NRC Inspection Report No. 50-285/99-03 Ooerations Operations personnel demonstrated excellent command and control of activities during
a tornado warning and repairs to 345 kV shield wire. The inspectors noted that correct
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procedures were in use and being referenced during a tornado warning and 345 kV repair activities. However, the licensee's failure to notify the electrical engineer prior to removing Circuit 3422 from service precluded an opportunity to identify the potential for a fast transfer and take appropriate contingency actions (Section 01.2).
In July 1996, the licensee noted leakage past one or both steam generator sample
isolation valves. Leakage past these valves would result in inaccurate information for dose rate determinations if a steam generator tube rupture occurred. Although the valves leaked by, the licensee did not declare the postaccident main steam line monitor inoperable. Failure to declare the postaccident main steam line monitor inoperable resulted in a violation of Technical Specification 2.21. The root cause for failing to declare the monitor inoperable was a lack of depth in the operability evaluation. This licensee identified Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the Enforcement Policy. This deficiency was included in the corrective action program as Condition Report 199701036 (Section 08.2).
Maintenance A parts database error resulted in a maintenance planner failing to identify the correct
part required to replace the casing vent valve for containment spray Pump SI-3A.
Additionally, during the required prejob walkdown, the maintenance craft personnel questioned whether the part was correct but did not raise their concern to maintenance supervision and proceeded to disassemble the valve. The parts database error and the maintenance personnel failure to communicate their concerns contributed to the containment spray pump being unavailable for an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Section M1.3).
During testing to satisfy an inservice testing requirement, the volume control tank
makeup inlet check valve failed its acceptance criteria. The inspectors identified that engineering personnel used a sledge hammer to mechanically agitate the check valve contrary to management expectations. Following mechanical agitation, the valve was retested and declared operable. When asked if the valve could be relied on to perform the function that caused it to be within the scope of the inservice testing program, the licensee indicated there was not a good basis for the check valve to be in the program.
This issue will remain unresolved pending the completion of an evaluation to determine if the check valve should remain within the inservice testing program (Section M2.1).
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The inspectors identified that replacement parts (capacitors and rectifiers) installed in
the control rod drive motor circuitry were not like-for-like replacements. Form, fit, and
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l l-2-function of the replacement parts were different. The licensee only evaluated the change in function. Failing to evaluate the change in fit and form was a violation of Technical Specification 5.8.1. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Policy. This deficiency was included in the corrective action program as Condition Report 199900615 (Section E1.1).
As a result of a Updated Final Safety Analysis review program, a noncited violation of
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10 CFR Part 50, Appendix B, Criterion 111, was identified by the licensee, in that design information was not properly translated into operatiens procedures. The licensee did not have a calculation to demonstrate the capability of the station batteries to operate for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a design basis accident without the battery chargers, as specified in the safety analysis. Upon completing the calculation, the licensee had to modify several
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emergency operating procedures to ensure operators would minimize direct current loads in order for the batteries to last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (Section E8.3).
Plant Support A violation of 10 CFR 50.54(q) occurred when the licensee made changes to
10 emergency action levels in Emergency Plan Implementing Procedure EPiP-OSC-1,
" Emergency Classification," Revision 27, that decreased the effectiveness of the emergency plan without prior NRC approval. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the NRC Enforcement Policy. This violation has been entered into the licensee's corrective action program as Condition Report 199801745. This closes Unresolved item 50-285/98-18-01 (Section P8.1).
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Report Details Summarv of Plant Status The Fort Calhoun Station began this inspection report period at 100 percent power and maintained that level throughout the inspection period.
1. Operations
Corduct of Operations 01.1 General Comments (71707)
The inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations was professional and safety conscious. Plant status, operating problems, and work plans were appropriately addressed during daily turnover and plan-of-the-day meetings. Plant testing and maintenance requiring control room coordination were properly controlled. The inspectors observed several shift turnovers and noted no problems.
01.2 Operator Response to Tornado Warnino and Partial Loss of Fast Transfer Capability a.
Insoection Scope (71707)
The inspectors observed the response of operations personnel to a tornado warning and the partialloss of fast transfer capability.
b.
Observations and Findinos On April 8,1999, at approximately 1 p.m., the plant entered a tornado warning. The tornado warning lasted until 2:10 p.m. The inspectors responded to the control room during the tornado warning and noted that activities were controlled. Abnormal Operating Procedure AOP-01, " Acts of Nature," was open and in use by the control room operators.
I During discussions with operations personnel, the inspectors learned that the capability for fast transfer from 161 kV power to 22 kV had been lost at 11:35 a.m. due to ongoing repairs on one of the 345 kV power lines. The repairs on the 345 kV line were necessary due to a lightning strike on one of the 345 kV towers which caused a shield wire to become disconnected. The shield wire was swinging in the wind and had the potential to short out additional 345 kV power lines. The repair of the disconnected shield wire was considered to be an emergency work activity as defined in Quality Procedure NOD-OP-36, " Control of Switchyard at FCS," Revision 9.
The inspectors questioned operations personnel concerning whether loss of the fast transfer capability from 161 kV power to 22 kV had been anticipated or discussed with engineering personnel prior to Omaha Public Power District electrical operations personnel removing power from Circuit 3422, which powered the 345 kV line that was being repaired. The operations supervisor stated that the loss of fast transfer capability had not been anticipated or discussed with engineering prior to disconnecting
Circuit 3422. However, when the annunciator "4160 volt bus trans to 22 kV blocked fast transfer" came in, operations personnel immediately referenced the annunciator response procedure and verified with the electrical engineer that the fast transfer capability from 161 kV power to 22 kV power had been lost. The fast transfer capability was lost due to the phase difference between 22 kV and the 161 kV electrical system exceeding a predetermined setpoint when Circuit 3422 was de-energized.
The inspectors and operations personnel discussed whether the electrical engineer or the probabilistic risk assessment group should have been contacted prior to removing Circuit 3422 from service. Operations personnel stated that this had not been done due to focusing on the shictd wire and the potential for the shield wire to short across portions of the 345 kV line. During discussions with operations personnel, the inspectors determined that at approximately 10 a.m. the identification of the dangling shield wire was made and a decision was made by system dispatch that Circuit 3422 would have to be de-energized. The actual circuit was not de-energized until 11:35 a.m. The inspectors determined additional contingency planning should have been done prior to deenergizing Circuit 3422. The inspectors discussed this observation with the operations supervisor and the assistant plant manager. They both agreed that the electrical engineer should have been contacted prior to deenergizing Circuit 3422.
The inspectors observed that frequent briefings were held during the event and that the licensed senior operator and the shift technical advisor were reviewing Emergency Operating Procedures EOP-00," Standard Post Trip Actions," EOP-20," Functional Recovery Procedure," and EOP 02, " Loss of Off-Site Power / Loss of Forced Circulation,"
in the event that an inadvertent loss of 161 kV power occurred without fast transfer capability available.
The inspectors noted that during the tornado warning and 345 kV repair activities the operations personnel discussed the importance of restoring fast transfer capability as soon as possible, and plant-wide announcements were made to restrict access to the diesel generator rooms and the switchgear rooms.
c.
Conclusiorm Operations personnel demonstrated excellent command and control of activities during a tornado warning and repairs to a 345 kV shield wire. The inspectors noted that correct procedures were in use and being referenced during i tornado warning and 345 kV repair activities. However, the licensee's failure to notify the electrical engineer prior to removing Circuit 3422 from service precluded an opportunity to identify the potential for a fast transfer and take appropriate contingency action.
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Operational Status of Facilities and Equipment O2.1 Review of Eauioment Taaouts (71707)
The inspectors walked down the following tagout:
Danger Tagout 99-0213 Raw Water / Component Cooling Watar Heat Exchanger AC-1B The inspectors did not identify any discrepancies. The tagout was properly prepared and authorized. All tags were on the correct devices and the devices were in the position prescribed by the tags. The inspector also performed a walkdown after the tagouts were cleared. All components were in the proper position for the required system lineup.
Operations Organization and Administration 06.1 Review of Evaluation and Accreditation Reoorts by the Institute of Nuclear Plant Operations (71707)
On April 1,1999, the inspectors reviewed the Institute of Nuclear Plant Operations'
Report for the Fort Calhoun Station. The report covered the previous 12 months of activity at the Fert Calhoun Station. The inspectors also reviewed and discuseed the institute of Nudmr Plant Operations' Training Accreditation Report with the licensee.
Miscellaneous Operations issues (92901)
O8.1 (Closed) Inspection Followuo item 50-285/98028-01: undervoltage on 161 kV line resulting in a fast transfer, due to severing a line in the switchyard during trenching activities.
On January 21,1999, the plant experienced an unplanned fast transfer from 161 kV power to 22 kV power due to sensed low voltage on the 161 kV line. Although the 161 kV line voltage was never disturbed, the sensed low voltage was due to a line from the low voltage potential transformer being severed during trenching in the Fort Calhoun Station switchyard.
The licensee has revised Quality Procedure NOD-OP-36, " Control of Switchyard at FCS." The inspectors verified the corrective actions described in the licensee's root cause analysis and considered the actions to be acceptable.
08.2 { Closed) Licensee Event Report 50-285/97011-00: violation of Technical Specifications resulting from inoperability of a radiation monitor.
l On August 5, i997, the licensee declared Radiation Monitor 64, the postaccident main steam line radiation monitor, inoperable because of leakage past Valve HCV-922, Steam Generator RC-28 sample isolation. The operators concluded that leakage past l
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i Valve HCV-922 would mix with the steam sample flow from Valve HCV ??1. Steam Generator RC-2A sample isolat,on, which would provide inaccurate information for determining dose rates if a tube rupture occurred in Steam Generator RC-2A. These sample points aid in determining the type of emergency to declare based on specific action criteria.
On August 19,1997, the licensee determined that, in July 1996, following similar valve leakage 3, operators did not declare Radiation Monitor 64 inoperable for an extended period. This resulted in a violation of Technical Specification 2.21, Table 2-10," Post-Accident Monitoring Instrumentation Operating Limits," Item 3, because the single channel was inoperablo. This Technical Specification requires restoration to the minimum channels be operable within 7 days or to prepare and submit a special report within 14 days detailing actions taken, the root cause, and plans for restoring the
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channel to operable. The licensee issued Condition Report 199701036 to evaluate this event and implement corrective actions. The licensee attributed the root cause to a lack l
of depth in the operability evaluation. A contributing cause was identified as allowing the radiation monitor to remain in service for long periods without considering sample dilution and cross contamination.
The inspectors verified the corrective actions described in the licensee's closeout of Licensee Event Report 97011-00 and considered the actions to be acceptable.
This licensee identified Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Policy (50-285/99003-01).
This deficiency wc.s included in the corrective action program as Condition Report 199701036.
08.3 Administrative Closure of Violations Based Upon Chanaes in the Enforcement Poliev J
The inspectors performed a review of the following outstanding violations in the operations area:
Violation 50-285/98009-01, reactor coolant Pump RC-3A pump cavitation,
corrective action program reference, Condition Report 199800640; and Violation 50-285/97015-01, inadequate corrective actions in that deficiency
stickers were not removed from control boards and charging pump suction valves were overtorqued, Condition Report 199701021.
These Severity Level IV violations were issued in Notices of Violation prior to the March 11,1999, implementation of the NRC's nes.' policy for treatment of Severity Level IV violations (Appendix C of the Enforcement Policy). Because these violations would now be treated as noncited violations, in accordance with Appendix C, they are being closed in this report. The inspectors verified that the licensee had generated a corrective action reference (condition report) for each of the violations listed. In addition, these violations have docketed response..
i 5-II. Maintenance M1 Conduct of Maintenance M1.1 General Comments - Maintenance a.
Insoection Scooe (62707)
The inspectors observed or reviewed portions of the following work activities:
Cleaning raw water / component cooling water Heat Exchanger AC-1B;
Replacement of casing vent valve for containment spray Pump SI-3A; and
Replacement of capacitors and rectifiers in control rod drive motor circuitry,
b.
Observations and Findinas With the exception of the maintenance described in Sections M1.3 and E1.1, the inspectors identified no substantive concerns. All work observed was performed with the work packages present and in active use. The inspectors frequently observed supervisors and system engineers monitoring job progress, and quality control personnel were present when required.
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Conclusions The maintenance activities observed were conducted in a controlled and professional manner.
M1.2 General Comments - Surveillance.
a.
Insoection Scope (61726)
The inspectors observed or reviewed all or portions of the following test activities:
S'ameillance Procedure OP ST-SI-3008, " Safety injection and Containment
Spray Pump inservice Test and Valve Exercise Test," Revision 24; and Calibration Procedure IC-CP-01-2909," Calibration of Safety injection Loop 1B
Check Valves Leakage Cooler SI-4A Pressure Control, Loop P-2909,"
Revision 3.
b.
Observations and Findinas The surveillance testing was conducted satisfactorily in accordance with the licensee's approved programs and the Technical Specifications.
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Conclusions
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Surveillance activities were generally completed thoroughly and professionally.
M1.3 Incorrect Electronic Parts Database
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Inspection Scooe (62707)
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The inspectors followed up on an error in the electronic parts database, b.
- Q,bservations and Findinas On March 15,1999, containment spray Pump SI-3A was removed from service to clean the mechanical seals and base plate and to replace a leaking casing vent valve. Work on the mechanical seals was scheduled to last 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> and replacement of the casing
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vent valves was scheduled for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. These activities were to be performed concurrently. Cleaning of the mechanical seals was completed that day. However,
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replacement of the casing vent valve could not be completed, as scheduled, due to a parts problem. Work Order Package 15197 identified the casing vent valve as a gate valve and stated to repair it using Procedura PE-RR-VX-0417S," Inspection and Repair of Safety Related Dresser Hancock Type 950W Gate Valve." When the valve was disassembled, maintenance personnel noted that the valve was a globe valve and should be replaced using Procedure PE-RR-VX-0418," Inspection and Repair of
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Dresser Hancock Type 5500W/5520W Globe Valves." Since the wrong parts and
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procedure were specified, maintenance personnel reassembled the casing vent valve.
Attempts were made to secure the correct parts, but there were no spare globe valve.
intemals on site. On March 16, parts were obtained, the valve was repaired, and the containment spray pump was returned to service.
j The licensee performed an evaluation and determined that errors in the work package were due to an error in the electronic parts database. The electronic parts database specified the casing vent valve as a gate valve when it should have identified the valve as a globe valve. A maintenance supervisor checked the olant drawing and noted that the valve was identified correctly as a globe valve.
Management expected maintenance personnel to perform field walkdowns and review work documents prior to starting an activity. The inspectors asked why the errors in the maintenance work documents were not identified prior to starting the work.
Mair:tenance personnel stated that the walkdowns were performed but they were unable to determine the valve type. Although maintenance personnel could not identify the valve type, no attempts were made to verify the type prior to disassembling the valve.
The valve was not correctly identified until after it was disassembled.
The inspectors asked the maintenance manager whether all required mair,tenance
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This was not satisfactorily performed because the maintenance planner specified the wrong maintenance procedure. Maintenance personnel stated that the incorrect maintenance procedure was specified because the electronic parts database incorrectly identified the casing vent valve as a gate valve. Maintenance personnel stated that, if the part database had been correct, the correct part and procedure would have been specified.
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Conclusions A parts database error resulted in a maintenance planner failing to identify the correct part required to replace the casing vent valve for containment spray Pump SI-3A.
Additionally, during the required prejob walkdown, the maintenance craft personnel questioned whether the part was correct but did not raiso their concern to maintenance supervision and proceeded to disassemblo the valve. The parts database error and the maintenance personnel failure to communicate their concerns contributed to the containment spray pump being unavailable for an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Mechanical Aaitation of Volume Control Tank Makeuo Inlet Check Valve a.
Inspection Scope (62707)
The inspectors followed up and evaluated the circumstances surrounding the mechanical agitr. tion of the volume control tank makeup inlet check valve, b.
Observations and Findinas On March 16,1999, the licersee performed surveillance Procedure OP-ST-CH-3002,
" Chemical and Volume Control System Category C Valve Exercise Test." The surveillance satisfied the requirements of Technical Specifications 3.3(1)a for check Valves CH 129 (boric acid Pump CH-4A discharge check valve), CH-130 (boric acid i
Pump CH-4B discharge check valve), and CH-151 (volume control tank makeup ir.!et I
check valve). These valves were included in the inservice testing program.
During the surveillance test, Valve CH-151 failed its acceptance criteria of less than 2 gallons backleakage. The surveillance was stopped when the backleakar'e exceeoed 6 gallons. The valve was immediately declared inoperable.
On the morning of March 17, the inspectors observed the system engineer and the shift technical advisor in the vicinity of Valve CH-151 carrying a sledge hammer. Later that morning, the inspectors asked the system engineer if the sledge hammer had been used to mechanically agitate the check valve to ensure it seated properly. The system engineer stated that the sledge hammer had been used to agitate the check valve but that tha force used was minimal. However, this minimal force could not be quantifie The inspectors also learned that, following the mechanical agitation, the surveillance procedure was successfully performed and the check valve was declared operable.
Licensee management stated that mechanically agitating the check valve was an unacceptable practice.
Since the check valve was in the inservice testing program, the inspectors asked if the check valve could be relied upon to perform the function that caused it to be within the scope of the inservice testing program witnout it being mechanically agitated. The i
licensee indicated that there was not a good basis for the check valve being in the inservice testing program. At the end of the inspection period, a contractor was j
reviewing the charging system to identify the basis for including the check valve within
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the scope of the program. If no basis could be determined, the licensee indicated that the check valve would be taken out of the inservice testing program. This issue will J
remain unresolved pending the completion of the licensee's evaluation I
(50-285/99003-02).
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Conclusions During testing to satisfy an inservice testing requirement, the volume control tank makeup inlet check valve failed its acceptance criteria. The inspectors identified that engineering personnel used a sledge hammer to mechanically agitate the check valve which was contrary to management expectations. Following mechanical agitation, the
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valve was retested and declared operable. When asked if the valve could be relied on to perform the function that caused it to be within the scope of the inservice testing program, the licensee indicated there was not a good basis for the check valve to be in the program. This issue will remain unresolved pending the completion of an evaluation to determine if the check valve should remain within the inservice testing program.
Ill. Enaineerina E1 Conduct of Engineering E1.1 Faihire to Evaluate Chances to Fit and Form Durino the Reolacement of Capacitors and Rectifiers Used in the Control Rod Drive Motors a.
Insoection Scope (37551)
The inspectors identified that engineering personnel did not evaluate changes to the fit and form of replacement capacitors and rectifiers during maintenance on the control rod drive motors.
b.
Observations and Findinas On March 22,1999, maintenance personnel began replacing rectifiers and capacitors in the control rod drive motor circuitry. The control rod drive motors move the 49 control rods. Failures which had occurred with these components were minimal. They were being replaced as a preventive measur.
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the control rod drive motor circuitry. The control rod drive motors move the 49 control i
rods. Failures which had occurred with these components were minimal. They were being replaced as a preventive measure.
On March 23, the inspectors were informed that the replacement rectifiers and capacitors were not identical replacement parts. The system engineer stated that the fit, form, and function of the replacement parts were different. The inspectors asked what type of analysis was performed that evaluated the differences in fit, form, and function.
The system engineer stated that the analysis was performed using the nonsignificant configuration change process as authorized by Standing Order SO-M-101,
" Maintenance Work Control." The system engineer provided the inspectors with a copy of each evaluation documented on Form 1173D as required by the procedure. The
inspectors noted that the system engineer had only evaluated the change in function for the rectifiers and capacitors. No evaluation for fit or form was performed. The inspectors asked why fit and form were not evaluated using the nonsignificant configuration change process. The system engineer stated that fit and form did not need to be evaluated using the nonsignificant configuration change process because the replacement parts were being used in a nonsafety-related application. The inspectors stated that configuration control was not limited to safety-related applications.
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When questioned again, the system engineer stated that not evaluating fit or form was acceptable because of the guidance outlined in Standing Order SO-M-101. The inspectors asked what effect a failure of the capacitors or re :tifiers would have on the centrol rods. Engineering personnel stated that, depending on how the failure occurred, it could prevent routine control rod movement.
A few days later, the inspectors asked engineering personnel if the control rod drive motors were described in the Updated Safety Analysis Report. Engir eerMg personnel stated that the motors were not described. Following this discussion, the inspectors searched the Updated Safety Analysis Report and noted that paragraph 3.7.2.3 described the control rod drive motor package. Procedure PED-OP-2, " Configuration Change Control," step 4.1.3, listed the components for which configuration control was required to be maintained. It included structures, systems, or components attached or connected structurally, mechanically, or electrically to any portion of a system, structure, or component described in the Updated Safety Analysis Report. Since the control rod drive motors were described in the Updated Safety Analysis Report, the inspectors concluded that configuration control was required to be maintained and an evaluation of fit and form should have been performed.
The inspectors also reviewed Standing Order SO-M-101and noted that step 5.1.10.B stated that, if a proposed maintenar ce activity will affect the fit, form, or function of a structure, system, or component, the system engineer shall be notified for evwuation and authorization of the configuration changt Step 5.1.10.D stated that if a configuration change is evaluated as a nonsignificant configuration change the system engineer shall document the evaluation on Form 1173D and attach the form to the maintenance work documents. Since the fit, form, and function were changed and cince
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l 10-the changes were evaluated as nonsignificant configuration changes, fit and form should have been evaluated before allowing the parts to be replaced.
Failing to perform an evaluation of fit and form for the replacement parts (capacitors and l
rectifiers) used in the control rod drive motor circuitry is a violation of Standing i
Order SO-M-101 and Technical Specification 5.8.1(50-285/99003-03),
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Conclusions The inspectors identified that replacement parts (capacitors and rectifiers) installed in the control rod drive motor circuitry were not like-for-like replacements. Form, fit, and function of the replacement parts were different. The licensee only evaluated the change in function. Failing to evaluate the change in fit and form was a violation of
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Technical Specification 5.8.1. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Policy. This deficiency was included in the corrective action program as Condition Report 199900615.
E8 idiscellaneous Engineering issues (92903)
E8.1 Administrative Closure of Violations Based Upon Chanaes in the Enforcement Policy The inspectors performed an in-office review of the following violation in the engineering
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l area. The Severity Level IV violation listed below was issued in a Notice of Violation prior to the March 11,1999, implementation of the NRC's new policy for treatment of Severity LevelIV violations (Appendix C of the Enforcement Policy). Because Violation 50-285/98019-01, inadequate corrective actions, in response to a potential reactivity event due to inadequate dilution through the postaccident sampling system, would now be treated as a noncited violation in accordance with Appendix C, it is being closed in this report. The inspectors verified that the licensee had generated a corrective action reference, Condition Report 199900056, for this violation.
E8.2 (Closed) Licensee Event Report 50-285/97-016-00: Nuclear fuel potentially outside of I
the manufacturer's fuel design criteria.
On October 28,1997, the nuclear fuel vendor informed the licensee about concerns with the fuel rod internal pressure, the status of the fuel performance code, and concerns with fuel rod design criteria. On this date the fuel vendor indicated that the amount of fuel cladding oxidation estimated by the fuel performance code may be nonconservative because o' pellet clad isopening and increased f uel cladding oxidation, which could result in exceeding the 17 percent maximum fuel cladding oxidation limit specified in 10 CFR 50.46. This issue applied to facilities that had integral fuel burnable absorber rods with Zircaloy-4 cladding because of the increased oxidation. When this data was included in the fuel performance code gap, reopening was predicted for second cycle fuel assemblies with burnup from 35-45.000 MWr'JMT. The fuct performance code included a design criteria that no pellet-clad gap would rxcur.
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l Based on loss of coolant analyses performed, the fuel vendor established a 12 percent oxidation limit as a screening criteria for assessing compliance of affected plants to the 17 percent maximum oxidation criterion. The fuel vendor had never exceeded 5 percent
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maximum fuel cladding oxidation during loss of coolant accident modeling. The j
inspectors confirmed, with personnel knowledgeable about this issue in the Office of Nuclear Reactor Regulation, that the NRC had no concerns with the fuel vendor's approach to assessing the amount of fuel clad degradation, as documented in the generic justification for continued operation issued October 27,1997.
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During this inspection, the inspectors determined that the fuel vendor provided the licensee with information indicating the maximum oxidation experienced in the Cycle 17 core would not have exceed 3d 11.8 percent. From discussions with reactor engineers the inspectors determined that the fuel vendor had used the actual burnup history for each of the fuel rods in the core in Cycle 17. The inspectors reviewed the operability evaluation and confirrned for Cycle 17 that: (1) Batches R, S, and T fuel assemblies had the Zircaloy-4 fuel cladding; (2) Batch T fuel assemblies were in the second most susceptible burnup cycle; (3) Batches R and S were in the third and fourth burnup cycle, respectively; and (4) Batch U fuel assemblies with ZlRCO fuel cladding had been loaded as fresh fuel. From review of the core reload map, the inspectors confirmed that the present Cycle 18 core load had no Batch R, S, or T fuel assemblies. The Cycle 18 fuel load had two batches (W and U) of fuel with ZlRCO cladding and previously burned Batches M and N fuel.
For integral fuel burnable assemblies with Zircaloy-4 cladding, new empirical data demonstrated greater than the design-predicted internal gas pressure and cladding oxidation. This new data resulted in decreased margin to exceeding the emergency core cooling system cladding oxidation limn. However, at no time were regulatory limits exceeded. The inspectors found the licensee evaluation and corrective actions appropriate.
E8.3 (Closed) Licensee Event Report 50-285/97-015-00: Unanalyzed condition for station batteries.
On October 16,1997, during implementation of the direct current system review as part of the Updated Final Safety Analysis Report verification described in the licensee response to the October 6,1996,10 CFR 50.54(f) letter, engineers determined that a
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calculation did not exist to demonstrate the capability of the batteries to operate for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a design basis accident (large break loss of coolant accident).
Specifically, the Updated Final Safety Analysis Report, Section 8.4.2.1, specified, in part, that analysis has demonstrated that the installed station batteries have adequate
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capacity to meet the present 8-hour load demand and have additional margin. The licensee determined that the existing station battery load calculation was for the 4-hour and 8-hour station blackout.
The licensee immediately declared the station batteries inoperable and initiated an operability evaluation (Safety Analysis for Operability 97-002," Station 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Battery Capacity for a Design Basis Accident"). Because the licensee presumed that the loads for the batteries would be similar for the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the design basis accident and
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-12-during a station blackout, the licensee concluded that the batteries could be considered operable in the interim. The licensee implemented compensatory measures in order to make the determination of operable batteries; specifically, the licensee issued Operations Memorandum 97-11, "DC Bus Actions During a Design Basis Accident,"
Revision 0, which informed the operators of the loads required to be minimized on the loss of a battery charger. Operations Memorandum 97-11 included Attachment 6,
" Minimizing DC Loads," of Emergency Operating Procedures / Abnormal Operating Procedures. Another immediate action included providing guidance specific to inoperability of Battery Charger 2 combined with accident conditions and a single tailure of Diesal Generator 1.
The inspectors determined that the licensee completed Calculation FC06680, " Battery Load Profile and Capacity Calculation for Design Basis Accident," Revision 0, on January 23,1998. Calculation FC06680 concluded that the batteries had sufficient capacity for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a design basis accident; however, the calculation included the assumption that operators reduced the direct current loads in accordance with Attachment 6 to the Emergency Operating Procedures / Abnormal Operating Procedures.
Tne licensee closed Safety Analysis for Operability 97-002 based on completion of Calculation FC06680 and compensatory actions in eight affected emergency operating procedures. The compensatory actions included: (1) minimiz;ng direct current loads, (2) adding notes to have direct current loads minimized within 15 minutes of a loss of a battery charger, (3) adding a check to the safety function status checks, and (4) adding floating steps to reference loss of battery charger power.
The inspectors noted that the licensee initiated Condition Report 199701430 to document this deficiency, initiated Root Cause Anr.iysis FC-155, and performed an operability evaluation in addition to the corrective actions described above, the licensee provided training to engineers on this evor t in an effort to sensitize them to the types of errors that could be identified in the Updated Final Safety Analysis Report reviews. This licensee-identified Severity Level IV violation of 10 CFR Part 50, Appendix B, Criterion !!!, is beino treated w a noncited violation, consistent with Appendix C of the Enforcement TW.;f (50-285/99003-04). This deficiency was included in the corrective action program as Condition Report 199701430.
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I IV. Plant Support
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R1 Radiological Protection and Chemistry Controls R1.1 Genere! Comments (71750)
The inspectors observed health physics personnel, including supervisors, routinely touring the radiologically controlled areas. Licensee personnel working in radiologically controlled areas exhibited good radiation worker practices.
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-13-R8 Miscellaneous Radiological Protection and Chemistry Controls issues (92904)
R8.1 Administrative Closure of Violations Based Uoon Chanaes in the Enforcement Policy The inspectors performed an in-office review of the following violation in the plant support area. The Severity Level IV violation was issued in a Notice of Violation prior to the March 11,1999, implementation of the NRC's new po' icy for treatment of Severity LevelIV violations (Appendix C of the Enforcement Policy). Because Violation 50-285/98024-01, use and control of temporary lead shielding, would now be treated as a noncited violation in accordance with Appendix C, it will be closed in this report. The inspectors verified that the licensee had generated a corrective action program reference, Condition Report 190801978. This violation already had a docketed response.
P8 Miscellaneous Emergency Preparedness ibba (92904/82701)
P8.1 (Closed) Unresolved item 50-285/98-18-01: Emergency action level changes.nay result in a reduction in effectiveness of the emergency plan.
Changes to the emergency action levels in Procedure EPIP-OSC-1, Revision 27, were reviewed against the criteria for emergency action levels in NUREG-0654/ FEMA REP-1,
' Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Revision 1, and NUMARC/NESP-G07. The Nuclear Reactor Regulation evaluation concluded that, although the integration of NESP-007 technical basis into the previous NUREG-0654-based emergency action level scheme was generally appropriate, some emergency action levels were unacceptable and/or lacked an adequate technical basis for acceptance. Some emergency action level changes which were not fully consistent wah the guidance were recommended for acceptance. The Office of Nuclear Reactor Regulation determined that changes to the following 10 emergency action levels did decrease the effectiveness of the emergency plan and were not acceptable:
Emergency Action Level 1.1 - Reactor Coolant System Radioactivity exceeds
Technicat specification Limits.
Change: Deleted applicsbility to Modes 4 and 5.
Emergency Action Level 8.5 - General Area Radiation / Airborne Levels Outside of
Containment are > 1000 times background.
Change: Deleted the applicability to containment.
- Emergency Action Level 9.1 - A confirmed hostage situation is occurring in the protected area, but outside of the vital areas.
Change: Reclassified the event from an alert to notification of unusual event emergency classificatio,
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Emergency Action Level 9.3 - Bomb / sabotage device detected in the protected
area, but outside the vital areas.
Change: Reclassified the event from an alert to notification of unusual event emergency classification.
Emergency Action Level 9.4 - Confirmed protected area intrusion.
- Change: Reclassified the event from an alert to notification of unusual event emergency classification.
Emergency Action Level 9.5 - Confirmed hostage situation in a vital area.
- Change: Reclassified the event from a site area emergency to an alert emergency classification.
Emergency Action Level 9.7 - A confirmed bomb / sabotage device has been
detected in a vital area.
Change: Reclassified the event from a site area emergency to an alert emergency classification.
Emergency Action Level 9.8 - A confirmed vital area intrusion is occurring.
e Change: Reclassified the event from a site area emergency to an alert emergency classification.
Emergency Action Level 9.9 - Confirmed armed attack is occurring inside a vital i
area that affects the ability to maintain functions needed for cold shutdown.
Change: Deleted applicability to cperating modes and added system-based restrictions on classification.
Emergency Action Level 9.10 - Confirmed armed attack inside a vital area that I
affects the ability to maintain functions needed for hot shutdown.
Change: Deleted applicability to operating modes and added system-based restrictions on classification.
ll Part 50.54(q) of Title 10 of the Code of Federal Regulations permits licensees to make changes to their emergency plans without prior NRC approval only if the changes do not decrease the effectiveness of the plans and the plans continue to meet the standards of 10 CFR 50.47(b). Accordingly, the failure to receive prior NRC approval for an j
emergency plan change that decreased the plan's effectiveness was identified as a
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violation of 10 CFR 50.54(q). This violation is in the licensee's corrective action program as Condition Report 199801745. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the NRC Enforcement Policy (50-285/99003-05).
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As part of the Task Interface Agreement, the Office of Nuclear Reactor Regulation reviewed the emergency action level changes in Procedure EPI-OSC-1, Revision 28.
Changes in this revision did not decrease the effectiveness of the emergency plan, with the exception of those Revision 27 emergency action levels listed above that were left unrevised in Revision 28.
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The licensee submitted Procedure EPIP OSC-1 Revisions 29,30, and 31 on October 26 and December 10.1998, and February 24,1999, respectively. The changes in these revisions resolved issues from Revision 27. Emergency Action Levels 1.1, 8.5, 9.9 and 9.10 were corrected in Revision 29. Emergency Action Levels 9.1,9.3,9.4,9.5,
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9.7, and 9.8 were corrected in Revision 31. Revision 30 did not address emergency action levels changed in Revision 27. Inspectors determined that the changes in Revisions 29,30, and 31 did not decrease the effectiveness of the emergency plan.
V. Mar emment Meetinas X1 Exit Meeting Summary I
The inspectors presented the inspection results to members of licensee management at the exit meeting on April 23,1999. The licensee acknowledged the findings as presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie.
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ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee D. Bannister, Operations Supervisor -
J. Chase, Manager, Nuclear Assessment J. Cook, Shift Manager M. Core, Manager, System Engineering M. Frans, Licensing Manager M. Puckett, Manager, Radiation Protection R. Meng, Senior Emergency Planning Representative H. Sefick, Manager, Security and Emergency Services C. Simmons, Supervisor, Emergency Planning R. Short, Assistant Plant Manager J. Skiles, Manager, Design Engineering J. Solymossy, Manager, Fort Calhoun Station R. Clemens, Manager, Maintenance INSPECTION PROCEDURES USED 37551 Onsite Engineering 61726 Surveillance Observations 62707 Maintenance Observations
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71707 Plant Operations 71750 Plant Support Activities 82701 Operational Status of the Emergency Preparedness Program 92700 Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities 92901 Followup - Plant Operations 92902 Followup - Maintenance 92903 Followup - Engineering 92904 Followup - Plant Support
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-2-ITEMS OPENED. CLOSED. AND DISCUSSED Opened 99003-01 NCV Violation of Technical Specification resulting from an inoperable radiation monitor (Section 08.2).
99003-02 URI Mechanical agitation of volume control tank makeup inlet check valve (Section M2.1).
99003-03 NCV Failure to evaluate changes to fit and form (Section E1.1).
99003-04 NCV Unanalyzed condition for station batteries (Section E8.3).
99003-05 NCV Emergency action level revision (Section P8.1).
Closed 98028-01 IFl Undervoltage on 161 kV line resulting in a fast transfer (Section O8.1).
97-011-00 LER Violation of Technical Specification resulting from an inoperable radiation monitor (Section 08.2).
99003-01 NCV Violation of Technical Specification resulting from an inoperable radiation monitor (Section 08.2).
98009-01 VIO Administrative closeout, reactor coolant pump cavitation (Section 08.3).
97015-01 VIO Administrative closeout, inadequate corrective actions (Section 08.3).
99003-03 NCV Failure to evaluate changes to fit and form (Section E1.1).
98019-01 VIO Inadequate corrective actions (Section E8.1).
97-016-00 LER Fuel potentially outside design criteria (Section E8.2).
97-015-00 LER Unanalyzed condition for station batteries (Section E8.3).
99003-04 NCV Unanalyzed condition for station batteries (Section E8.3).
98024-01 VIO Use and control of temporary lead shielding (Section P8.1).
98018-01 URI Revisions to emergency action levels (Section P8.1).
99003-05 NCV Emergency action level revision (Section P8.1).
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-3-Discussed None LIST OF DOCUMENTS REVIEWED Emergency Plan Implementing Procedures:
. EPIP-OSC-1, Emergency Classification," Revision 26 EPIP-OSC-1, Emergency Classification," Revision 27 EPIP-OSC-1, Emergency Classification," Revision 28 EPIP-OSC-1, Emergency Classification," Revision 29 EPIP-OSC-1, Emergency Classification," Revision 30 EPIP-OSC-1, Emergency Classification," Revision 31 Other Documents:
Emergency Action Level Basis Document for Procedure EPIP-OSC-1, Emergency Classification," Revision 27 Nuclear Safety Evaluation ids 47784 and 47784A Condition Report 199801731 Memorandum FC-RP-119-98