IR 05000528/1999017
| ML17300B360 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 10/02/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17300B359 | List: |
| References | |
| 50-528-99-17, 50-529-99-17, 50-530-99-17, NUDOCS 9910120220 | |
| Download: ML17300B360 (23) | |
Text
ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket Nos.:
License Nos.:
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved By:
50-528; 50-529; 50-530 NPF-41; NPF-51; NPF-74 50-528/99-17; 50-529/99-17; 50-530/99-17 Arizona Public Service Company Palo Verde Nuclear Generating Station, Units 1, 2, and 3 5951 S. Wintersburg Road Tonopah, Arizona August 30 through September 3, 1999 M. Runyan, Senior Reactor Inspector Engineering and Maintenance Branch W. McNeill, Reactor Inspector Engineering and Maintenance Branch Dr. Dale A. Powers, Chief, Engineering and Maintenance Branch Division of Reactor Safety ATTACHMENT Supplemental Information 99iOi20220 99i002 PDR ADQCK OMOOS28
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-2-EXECUTIVE SUMMARY Palo Verde Nuclear Generating Station, Units 1, 2, and 3 NRC Inspection Report No. 50-528/99-17; 50-529/99-17; 50-530/99-17 Encnineering The licensee had established and implemented an effective program for performing 10 CFR 50.59 safety evaluations.
Safety evaluations reviewed were observed to be rigorous, comprehensive, and of high quality (Section E2.1).
The licensee's policy on increase in consequences of an analyzed accident was not consistent with the requirements of 10 CFR 50.59. The licensee defined an increase in consequences to indicate that the calculated dose exceeds an applicable regulatory limit established in 10 CFR Part 100 or the General Design Criteria. The NRC considers an increase in consequences to constitute any increase in dos'e beyond that previously calculated and reported in the Updated Final Safety Analysis Report. Over the past 3 years, no violation examples existed (Section E2.1).
A violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for failure, during initial plant design and construction, to apply adequate design controls to ensure that cables associated with the station batteries and the emergency lighting batteries in each unit were separated in accordance with the requirements of Regulatory Guide 1.75. This Severity Level IVviolation is being treated as a noncited violation, consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as Condition Report/Disposition Request 99-1 00622 (Section E2.1).
A violation of 10 CFR Part 50, Appendix B, Criterion XVI,"Corrective Action," was identified for failure to take prompt corrective action in response to the licensee's
- discovery of an electrical cable separation nonconformance associated with the station and emergency lighting batteries in Units 1 and 3. This Severity Level IVviolation is being treated as a noncited violation, consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as Condition Report/Disposition Request 99-100622 (Section E2.1).
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-3-Re ort Details Summa of Plant Status Units 1, 2, and 3 were in Mode 1 during the inspection.
E2 Engineering Support of Facilities and Equipment E2.1 10 CFR 50.59 Pro ram and Im tementation a. 'ns ection Sco e 37001 The inspectors reviewed the licensee's 10 CFR 50.59 safety evaluation program and its implementation.
The review included an examination of procedures, screenings, safety evaluations, and the training and qualification records of personnel.
b.
Observations and Findin s 10 CFR 50.59 Pro ram The licensee's 10 CFR 50.59 program was described in Procedure 93DP-OLC07,
"10 CFR 50.59 Screenings and Evaluations," Revision 2.
The licensee reviewed changes to the plant as described in the Updated Final Safety Analysis Report (UFSAR), first for applicability to 10 CFR 50.59, then screened changes that did not require evaluations, and lastly performed safety evaluations, if necessary.
Section 3.2,1 of Procedure 93DP-OLC07 listed criteria for excluding changes from a 10 CFR 50.59 screening.
This criteria included the presence of an existing evaluation or screening that adequately addresses the change, a change that merely restores conformance with the UFSAR, a change that was previously approved by the NRC, a change that was purely administrative or editorial in nature, a correction to UFSAR inconsistencies, or minor drawing corrections.
The inspectors considered the exclusion criteria to be appropriate.
For those changes determined to be applicable to 10 CFR 50.59, Procedure 93DP-OLC07 listed criteria used to determine whether safety evaluations were necessary and, if so, listed questions to determine whether unreviewed safety questions existed.
The inspectors considered this procedural guidance to be consistent with 10 CFR 50.5 CFR 50.59 Screenin s
The inspectors reviewed 18 10 CFR 50.59 screenings where the licensee had concluded that safety evaluations were not required.
In each case, the screening document adequately justified the decision to.not perform a safety evaluation.
10 CFR 50.59 Safet Evaluations The inspectors reviewed 27 10 CFR 50.59 safety evaluations.
The inspectors found each of these evaluations to be rigorous, comprehensive, and of high quality.
10 CFR 50.59 Non-A licabilit Reviews The inspectors reviewed three safety-related changes that were determined by the licensee to not require a 10 CFR 50.59 screening or safety evaluation because they met a prescribed criterion for exclusion from the 10 CFR 50.59 process.
The inspectors identified a concern related to one of the non-applicability reviews.
Deficiency Work Order 00828838, dated January 29, 1998, identified that a Class 1E cable of the station battery in Unit 2 was within 36 inches (horizontal) of the dc power cable jumper of the non-Class 1E emergency lighting battery, without the presence of a damage limiting barrier. This configuration was in conflict with the requirements of Regulatory Guide 1.75, to which the licensee was committed. Condition Report/Disposition Request (CRDR) 1-8-0013 was written on January 8, 1998, to document this condition, which was observed to exist in two battery rooms in each of the-three units. Therefore, a total of six separation discrepancies existed.
The nonconforming cable separation distances ranged from approximately 27 to 30 inches, versus the 36-inch requirement.
During the inspection, the licensee's representative speculated that the cable separation problem originated in 1992 and 1993 when new round cell batteries were installed in each unit. At the time of installation, the emergency lighting batteries were physically covered for protection, and, therefore, the associated cables were not noted to be located within the Regulatory Guide 1.75 minimum separation distance.
Subsequent to the inspection, the licensee's representative informed the inspectors that a review of original specifications indicated that the discrepancy had existed since the time of initial plant licensing (Unit 1-December 31, 1984, Unit 2-December 9, 1985, and Unit 3-March 25, 1987).
Criterion III of Appendix B, 10 CFR Part 50, "Design Control," states, in part, that measures shall be established to assure that applicable regulatory requirements for safety-related components are correctly translated into specifications, drawings, procedures, and instructions. The UFSAR committed to Regulatory Guide 1.75, which required that safety-related cables be separated by at least 36 inches horizontally. The licensee's failure, during initial plant design and construction, to assure that these regulatory requirements were correctly translated to drawings and specifications associated with the station and emergency lighting battery cables was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III. This Severity Level IVviolation is
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-5-being treated as a noncited violation (50-528; -529, -530/9917-01), consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as CRDR 99-100622.
Deficiency Work Order 00828838 and the Deficiency Work Orders 00828834, 00828835, 00828837, 00828839, and 00828840, addressing the five other separation discrepancies, were all dispositioned as not requiring a 10 CFR 50.59 screening or evaluation because the licensee intended to restore compliance to Regulatory Guide 1.75 during the next refueling outage on each unit (Unit 1-Spring 1998, Unit 3-Fall 1998, and Unit 2 - Spring 1999). This work was to be accomplished in coordination with a replacement of the station batteries in each unit (installation o'f square cell batteries).
An operability evaluation was not performed in response to CRDR 1-8-0013 or any of the deficiency work orders.
However, CRDR 1-8-0013 included a statement that the "Regulatory Guide 1.75 separation requirements are not considered an equipment operability issue as detailed in CRDR 95-Q633.
However, identified deficiencies should be resolved in a timely manner.", Condition Report/Disposition Report 95-Q633 was noted by the inspectors to not contain an analytical assessment of operability, but rather referred to an NRC internal letter (Darrell G. Eisenhut to John A.
Olshinski, December 20, 1983) that contained the following text:
"We do not believe that the physical separation barriers of R.G. 1.75 represent a first order of safety importance.
More significant are those fire barriers of Appendix R which provide protection against an exposure fire affecting redundant safe shutdown capability and as such are included in the Technical Specifications.
Therefore, we do not believe that physical separation barriers of R.G. 1.75 should be included in the Technical Specifications."
On this basis, the licensee concluded that [quoted from CRDR 95-Q633] "ft]herefore, it appears that the NRC does not consider nonconformances to Regulatory Guide 1.75 to be an 'operability'ssue."
The inspectors considered the above by itself to not necessarily constitute an adequate operability basis.
However, the stationing of fire watches in these areas (which had been established previously for other reasons)
provided justification for operability. Additionally, the inspectors did not consider the separation deficiencies to be a significant technical concern because of the existing separation (at least 27 inches) and the low probability of faults or fires propagating across this distance.
As of September 3, 1999, approximately 20 months after initial discovery of the problem, only two of the six separation problems had been corrected.
This occurred during the Unit 2 refueling outage in Spring 1999, when the round cell station batteries were replaced.
Because the battery replacement itself did not correct the separation problem, damage limiting barriers were erected in each of the Unit 2 battery rooms to restore Regulatory Guide 1.75 compliance.
The separation problems in Units 1 and 3 were not corrected during the two refueling outages in 1998 because of vendor supply problems
-6-that delayed by one operating cycle the battery replacement in these units. The licensee did not reconsider the corrective action plan for the Unit 1 and 3 separation deficiencies in response to the delay in replacing the batteries in these units. The licensee's representative stated that installation of the new batteries and correction of the separation issues would occur during the Fall 1999 refueling outage for Unit 1 and the Spring 2000 refueling outage for Unit 3.
The inspectors concluded that the licensee's actions in this instance were not in violation of 10 CFR 50.59 because Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions," Revision 1, does not require a safety evaluation for nonconforming conditions that the licensee intends to restore to compliance.
However, within this allowance was an expectation that compliance will be restored at the next available opportunity.
In this case, the licensee failed to re-evalute the corrective action plan for the cable separation problems when the original corrective action schedule was delayed by one operating cycle. 'During the inspection and.in response to the inspectors'inding, the licensee issued CRDR 99-100622, which was generated to review the corrective action issues and to perform a formal operability review of the separation issue.
As noted above, the inspectors did not consider. the nonconformances to constitute an immediate safety or operability concern.
Criterion XVIof Appendix B, 10 CFR Part 50, "Corrective Action," states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
The licensee failed to correct in a prompt manner nonconformances associated with Regulatory Guide 1.75 separation criteria in Units 1 and 3. This failure was identified as a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This Severity Level IVviolation is being treated as a noncited violation (50-528; -530/9917-02), consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as CRDR 99-100622.
Conse uences of an Accident The inspectors noted that the licensee's definition of an increase in accident consequences endorsed Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Safety Evaluations," Revision 0, dated September 1997. This document identified that an increase in calculated dose resulting from a plant change would not be considered an increase in consequences under 10 CFR 50.59 unless the revised doses exceeded regulatory limits as defined in 10 CFR Part 100 or the General Design Criteria. The only exception was where the licensing document specifically addressed and credited the existing margin between the calculated and limiting doses.
The licensee's policy on increase in consequences was not consistent with the NRC position, which establishes the existing dose calculations in the UFSAR as the basis for determining if an increase in consequences has occurred.
Use of the licensee's policy, could result in NRC enforcement action. The inspectors noted that during the time period reviewed within this inspection (3 years), a violation did not exis erator Res onse Times The inspectors did not find in any procedure or lesson plan where the licensee had emphasized to reviewers the possibility of a 10 CFR 50.59 impact resulting from changes to environmental conditions (physical barriers, heat, oxygen, radiation, lighting, radiography, steam, etc.), as well as procedure changes that add operator actions to recovery efforts, or modifications that cause equipment response to be slower (such as, increasing motor-operated valve gear ratios). The concern was that assumed operator response times in the UFSAR may be affected by subtle changes of this nature.
One site-specific example that appeared relevant to this concern was a change within UFSAR, Section 9.5.3, to reflect that areas not routinely inhabited by personnel may not be equipped with emergency lighting (Safety Evaluation 97-00140).
The licensee intended to use this UFSAR change to permit removal of batteries from emergency lighting fixtures in containment while the unit was in Modes 1 through 4. In cases where a containment entry was needed during or immediately following operation in these modes, portable lighting was to be used to provide necessary illumination, if offsite power sources supplying normal lighting is not available.
The inspectors noted that the 10 CFR 50.59 evaluation did not address the impact that this change could have on assumed operator response times, noting that use of portable illumination would likely result in increased times to perform certain tasks.
The inspectors questioned the licensee's representative whether any operator action times in the UFSAR were applicable to this case and were informed that there were no operator actions necessary for safe shutdown inside containment for any accident scenarios.
Trainin and Qualifications The inspectors assured, through a spot check of selected safety evaluations, that the preparers and reviewers of 10 CFR 50.59 evaluations and screenings were included on the licensee's list of qualified personnel.
The inspectors reviewed two training documents (NGS03-02-RC-001-00, "50.59 Requalification (1998)," June 19, 1998, and NGT02-04-RC-001-008, "Engineering Support Personnel Training," September 23, 1998) and observed that they, with the exception of the definition of accident consequences (as discussed above), were consistent with the requirements of 10 CFR 50.59.
'Conclusions The licensee had established and implemented an effective program for 10 CFR 50.59 evaluations.
The inspectors found the 10 CFR 50.59 evaluations to be rigorous, comprehensive, and of high quality.
The licensee's policy on increase in consequences of an analyzed accident was not consistent with the requirements of 10 CFR 50.59. The licensee defined an increase in consequences to indicate that the calculated dose exceeds an applicable regulatory limit
-8-established in 10 CFR Part 100 or the General Design Criteria. The NRC considers an increase in consequences to constitute any increase in dose beyond that previously calculated and reported in the UFSAR. Over the past 3 years, no violation existed.
A noncited violation was identified for failure, during initial plant design and construction, to apply adequate design controls to ensure that cables associated with the station batteries and the emergency lighting batteries in each unit were separated in accordance with the requirements of Regulatory Guide 1.75.
A noncited violation was identified for failure to take prompt corrective actions in response to the discovery of an electrical cable separation nonconformance associated with the station batteries in Units 1 and 3.
V. Mana ement Meetin s X1 Exit Meeting Summary An exit meeting was conducted on September 3, 1999. A supplemental exit meeting was conducted by telephone on September 28, 1999. The licensee's management acknowledged the findings of the inspection.
The licensee's management stated that no proprietary information had been reviewed during the inspectio ATTACHMENT SUPPLEMENTAL INFORMATION PARTIALLIST OF PERSONS CONTACTED Licensee A. Abbate, Engineer S. Bauer, Licensing Section Leader, Regulatory Affairs B. Blackmore, Engineer L. Bullington, Engineer D. Fan, Department Leader, Design Engineering M. Grissom, Senior Engineer, Regulatory Affairs W. Ido, Vice President, Nuclear Production D. Marks, Section Leader, Regulatory Affairs T. Mitchell, Maintenance Engineer D. Oaks, Inservice Testing Section Leader G. Overbeck, Senior Vice President, Nuclear Production P. Paramithas, Section Leader Instrument L Control Design R. Stroud, Senior Consultant, Regulatory Affairs M. Winsor, Director, Nuclear Engineering R. Younger, Department Leader, Nuclear Assurance NRC N. Salgado, Resident Inspector INSPECTION PROCEDURES USED 37001 10 CFR 50.59 Safety Evaluation Program ITEMS OPENED AND CLOSED
~Oened 50-528; -529; -530/9917-01 NCV Failure to Assure Regulatory Guide 1.75 Separation Distances for Battery Cables (Section E2.1)
50-528; -530/9917-02 NCV Failure to Promptly Correct Electrical Cable Separation Problem (Section E2.1)
-2-Ctosed 50-528; -529; -530/9917-01 NCV 50-528; -530/9917-02 Failure to Assure Regulatory Guide 1.75 Separation Distances for Battery Cables (Section E2.1)
NCV Failure to Promptly Correct Electrical Cable Separation Problem (Section E2.1)
Procedures DOCUMENTS REVIEWED 93DP-OLC07, "10 CFR 50.59 Screenings and Evaluations," Revision 2 73DP-OZZ03, "System and Maintenance Engineering," Revision 9 30DP-9WP02, 'Work Document Development and Control," Revision 24 10 CFR 50.59 Re orts May 21, 1999 (January - December 1998)
Condition Re ort/Dis osition Re uests 95-Q633, December 8, 1995 1-8-0013, January 8, 1998 2-8-0074, March 25, 1998 99-1 00622, September 2, 1999 Classroom Lectures NGS03-02-RC-001-00, "50.59 Requalification (1998)," June 19, 1998 NGT02-04-RC-001-008, "Engineering Support Personnel Training," September 23, 1998 Indust Documents Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Safety Evaluations,"
September 1997 Safet Evaluations 95-00303, "Design Master Work Order 0071 5063," October 25, 1995 97-00022, "Design Master Work Order 00766255," December 24, 1996
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97-00049, "Design Master Work Order 00768502," March 4, 1997
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-3-97-00140, "Design Master Work Order 00773802 and Safety Analysis Report Change Notice 3752," July 2, 1997 97-00238, "Revision to PVNGS UFSAR and Technical Specification Bases for RWT ESF Reserve Volume," July 8, 1998 97-00256, "Deletion of UFSAR Commitment to Automatic Differential Pressure Isolation of AS System HELB," November 26, 1997 97-00271, "WO 0082541 7," December 22, 1996 97-00273, "Safety Analysis Report Change Notice 3822," January 13, 1998 98-00008, "Deletion of Certain DG Valves From The IST Program.
This Change WillBe Made By Revision 5 to 73DP-PXI01," February 3, 1998 98-00042, "Temporary Modification Implemented Via Procedure 31MT-9IA02," March 6, 1998 98-00050, "CRDR 9-8-0394," March 13, 1998 98-00068, "DFWOs 00830234, 00830235, 00830236, 008300237 - SG 11 and 12 Tube Plugging and Staking During U1R7," April 9, 1998 98-00076, "Temporary Modification 1-98-SH-009," April 22, 1998 98-00077, "DFWO 00837869," April23, 1998 98-00082, "WO 00838582," May 5, 1998 98-00114, "Deletion of Special Test Exceptions From the TRM," July 15, 1998 98-00120, "Reduction in Action Requirements for Post Accident Monitoring System, TRM 3.3.105," July 15, 1998 98-00139, "Procedure Revision," October 2, 1998 98-00142, "CRDR 9-8-1283," August 28, 1998 98-00146, 'WO 00852724," August 28, 1998 98-00155, "Evaluation of SDC Operation Requirements to Support Reduced Inventory Conditions for U3R7," September 17, 1998 98-00165, "Licensing Document Change Request 98-13012," September 24, 1998 98-00173, "WO 00849161," October 14, 1998 98-00176, "Work Order 00858695," no date noted
99-00063, "Configuration Calculation 13-MA-PC-992," April 28, 1999 99-00071, "Calculation 13-NC-CH-101, Revision 3," May 14, 1999 99-00083, "Calculation 13-MC-SI-215, Revision 3," July 14, 1999 Safet Evaluation Screenin s
Work Order 00885570, June 11, 1999 Maintenance Engineering Evaluation MEE 02445, May 18, 1999 Procedure 40OP-9ZZ05, Revision 28, May 11, 1999 Procedure 81DP-OEE02, Revision 0, July 11, 1999 Work Order 00871666, April 30, 1999 Work Order 00853062, January 14, 1999 Work Order 00870204, April 16, 1999 Maintenance Engineering Evaluation MEE 02365.00, May 18, 1999 Calculation 13-NC-PC-200, Revision 2, May 26, 1999 Calculation 13-MC-CH-533, Revision 8, June 4, 1999 Calculation 13-MC-CH-537, Revision 5, June 4, 1999 Calculation 03-MC-CH-535, Revision 3, May 27, 1999 USAR Tables 9.5-11 and 9.5-12, July 16, 1999 Calculation 13-NC-SI-203, Revision 1, May 27, 1999 Calculation 13-MC-ZZ-217, Revision 2, July 9, 1999 Design Master Work Order 00851801, Revision 2, August 10, 1999
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Calculation 13-JC-RC-206, Revision 4, May 13, 1999 Work Order 00878317, April 14, 1999 Non-A Iicabilit Screenin s
Deficiency Work Order 00713398, June 6, 1995 Deficiency Work Oider 00828838, January 29, 1998 Deficiency Work Order 00874909, May 11, 1999