IR 05000275/1999012
| ML20212B176 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 09/13/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20212B173 | List: |
| References | |
| 50-275-99-12, 50-323-99-12, NUDOCS 9909200004 | |
| Download: ML20212B176 (22) | |
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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket Nos.: 50-275 50-323 License Nos.: PDR-80 PDR-82 Report No.: 50-275/99 12 50-323/99-12 Licensee: Pacific Gas and Electric Company Facility: Diablo Canyon Nuclear Power Plant, Units 1 and 2 Location: 7 % miles NW of Avila Beach Avila Beach, California Dates: July 11 through August 21,1999 Inspectors: David L. Proulx, Senior Resident inspector Dyfe G. Acker, Resident inspector Gregory A. Pick, Senior Project Engineer John G. Kramer, Resident inspector, San Onofre Approved By: Linda J. Smith, Chief, Project Branch E ATTACHMENT: Supplemental Information 9909200004 990913 PDR ADOCK 05000275 G PDR
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EXECUTIVE SUMMARY Diablo Canyon Nuclear Power Plant, Units 1 and 2 NRC Inspection Report No. 50-275/99-12; 50-323/99-12 This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report documents inspection performed during a 6-week period by the resident inspector Operations
During routine operations, the performance of plant operators reflected a focus on safety and was generally characterized by self- and peer-checking (Section O1.1). l
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I The licensee responded appropriately to indications ef voiding in the emergency core I cooling system. The licensee demonstrated good initiative, by plant walkdowns and systems analysis, in identifying potential locations in which gas voids could have I developed (Section O2.1).
Maintenance
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Routine maintenance and surveillance activities observed were performed properl The licensee appropriately integrated risk insights into the work planning process (Sections M1.1 and M1.2).
- The inspectors observed that technicians left a test cart unrestrained near a panel l containing Seismic Category 1 equipment. The licensee subsequently determined that the test cart in question was made of lightweight plastic and would not damage plant equipment during a seismic event. However, the inspectors noted that the technicians erroneously believed that leaving unrestrained equipment near Seismic Category i I equipment for less than a shift was an acceptable practice. The Technical Maintenance supervisor stated that technicians routinely implemented this improper practice. This
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item is in the corrective action program as Action Request A0489903 (Section M2.1).
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= A violation of Technical Specifications 4.7.3.1.a and 4.5.3.2 resulted from two separate l examples of failing to seal valves. Both examples had existed since initial startup of the units in 1984 and 1985. On June 28,1998, the licensee determined that they had not
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l sealed the component cooling water heat exchanger crosstie valves open as specified in Technical Specification 4.7.3.1 a. In addition, on July 25,1997, the licensee determined that personnel f ailed to seal the safety injection pump manual discharge valves during testing, as specified in Technical Specification 4.5.3.2. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Policy. The licensee placed these deficiencies in their corrective action program as Nonconformance Reports N0002065 and N0002031, respectively (Sections M and M8.4).
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- A violation of Technical Specification 3.3.3.10, Action B, resulted because personnel f ailed to comply with the requirement to take 4-hour grab samples whenever the gas analyzer was out of service. Specifically, in 1998, during performance of a calibration procedure, personnel determined that the waste gas system oxygen analyzer alarm / trip
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2-functions were not in service. Review determined that the system was out of service for a total of 26 days. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Policy. The licensee included this deficiency in the corrective action program as Nonconformance Report N0002073 (Section M8.5).
- A violation of Technical Specification 4.3.1.1 occurred when personnel failed to properly test the reactor coolant pump undervoltage and underfrequency trip functions. On March 6,1998, a system engineer determined that a procedure change implemented in February 1997 changed the test methodology such that not all relay contacts were tested. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Policy. The licensee included this {
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deficiency in the corrective action program as Nonconformance Report N0002057 (Section M8.6).
Enaineerina
- A violation (EA 99-219) of 10 CFR 50.59 was identified for failure to perform a safety evaluation for a change to the facility. Specifically, in March 1999, the licensee modified the auxiliary saltwater pump vault drain lines, as described in Final Safety Analysis Report Update Drawing 3-17, but did not perform a written safety evaluation to ,
determine if an unreviewed safety question existed. The licensee installed y-strainers and low point drains upstream of the auxiliary saltwater vault drain line check valves, which had the safety function of mitigating the consequences of a design basis tsunam The licensee subsequently performed a safety evaluation that demonstrated that an unreviewed safety question did not exist. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Polic This item is in the corrective action program as Action Request A0487899 (Section E1.1).
- The licensee demonstrated a sensitivity to safety and good engineering judgement
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during resolution of two separate technicalissues. Both issues resulted from notification.c by off-site design vendors. The first issue involved the potential to fail the pressurizer safety valves during an inadvertent safety injection at power because the water would be too cool. The second issue involved a reduction in the design margins '
related to integral fuel bumable assemblies manufactured with Zircaloy-4 cladding on i
the fuel rods. The licensee included these deficiencies in the corrective action program ;
l I as Nonconformance Report N0002048 and Action Request A0446521, respectively l (Sections E8.2 and E8.3).
Plant Succort
[ * Routine radiation protection activities were performed carefully with two exceptions. The inspectors identified that a surface contamination area barrier and rope were degraded and that tube oilleaked out of a contaminated area. Radiation protection personnel took appropriate action for these concerns (Section R1.1).
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- Because of inadequate communications among offsite licensee personnel, the county emergency sirens were inoperable for an excessive period of time. The computer that initiates the emergency warning system failed but was not restored to operable status irnmediately. The alarm signifying that the computer failed annunciated in the emergency operations facility but was not fully communicated to repair personnel for 1% hours. The licensee appropriately investigated and dispositioned this issue (Section P1.1),
- Routine security activities were performed well (Section S1.1).
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Report Details Summary of Plant Status Units 1 and 2 operated at essentially 100 percent power throughout this inspection perio l. Operations 01 Conduct of Operations 01.1 General Comments (71707)
The inspectors visited the control room and toured the plant on a frequent basis when on-site, including periodic backshift inspections. In general, the performance of plant operators reflected a focus on safety. Operator performance was generally characterized by self- and peer-checking. The utilization of three-way communications continued to improve, and operators responded promptly to alarm Operational Status of Facilities and Equipment
! O2.1 Gas Bubbles identified in Residual Heat Removal (RHR) Supolv to Safety iniection (SI)
! Pumo Suction insoection Scope (71707. 92901)
l The inspectors reviewed licensee actions to identify, evaluate, and eliminate a gas bubble located upstream of Valve SI-88048, Train B discharge to SI, and centrifugal charging pump suctions.
l Observations and Findinas
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l As part of a continuing review of void formation in the Si system (refer to NRC Inspection Report 50-275; 323/99-07), the licensee identified a void upstream of
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Valve SI-88048. The engineers determined that a vent for removing gas from this high j point had been installed at an existing test connection; however, the connection from the
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vent pipe was approximately 2 feet below the centerline of the valve, which prevented removing all of the trapped air. The licensee isolated the suction supply to the Si and centrifugal charging pumps and slowly opened Valve SI-8804B in an attempt to have the refueling water storage tank pressure move the air to a downstream vent valve; however, the void disappeared. The iicensee determined that, as personnel opened Valve SI-8804B, the air flowed around the valve plug and into the bonnet area for Valve SI-8804B. The inspectors reviewed piping isometric and valve configuration drawings and discussed this scenario with engineering personnel. The inspectors agreed with the conclusions of the engineers regarding this scenari Af ter identification of the void, operators requested that engineering perform a preliminary operability assessment of the as-found void configuration. This was necessary because Valve SI-8804B is opened during cold-leg recirculation. Engineers determined that the void volume was 0.43 cubic feet and would not have affected
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2-l assessment with engineering personnel. The engineers determined that, under maximum design flows, the trapped air would be very buoyant and not easily entrained into the fluid at this location. The inspectors identified no concerns with the engineers'
conclusion Conclusions The licensee responded appropriately to indications of voiding in the emergency core cooling system. The licensee demonstrated good initiative, by plant walkdowns and systems analysis, in identifying potential locations in which gas voids could have develope Quality Assurance in Operations 07.1 Institute of Nuclear Power Operations Evaluation (71707)
The inspectors reviewed the report of the Institute of Nuclear Power Operations
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evaluation of the Diablo Canyon facility that was performed in April 1999. The purpose of the review was to determine if the Institute of Nuclear Power Operations evaluation results identify safety or training issues not previously identified by NRC evaluations, which are significant enough to require NRC followup. No additional NRC followup is planne Miscellaneous Operations issues (90712)
08.1 (Closed) Licensee Event Report (LER) 323/1999-001-00: voluntary entry into Technical Specification 3.0.3 to open the containment recirculation sump sensor hatch to verify level transmitter operabilit For 2 minuter, on March 31, and for 10 minutes on April 7,1999, the licensee opened the containrv ent recirculation sump sensor hatch to remove and install, respectively, the sensor assembly for Level Transmitter 2-LT 940, narrow-range containment recirculation sump. Each evolution required vcluntarily entering Technical Specification 3.0.3. The licensee determined that removal and calibration of the Unit 2 sensor assembly was necessary after determining that similar assemblies had been found out-of-tolerance on Unit 1 during the refueling outage in April 1999. The licensee determined that a degraded electrolytic capacitor had contributed to the out-of-calibration condition. The inspectors concluded that the LER appropriately characterized the issu .
-3-11. Maintenance M1 Conduct of Maintenance M1.1 General Comments on Maintenance Activities insoection Scope (62707)
The inspectors observed portions of work activities covered by the following work orders:
R0190968 Auxiliary Feedwater Pump Turbine 1-1 bearing oil sample
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C0163083 Auxiliary Feedwater Pump Turbine 1-1 replacement of Suction Relief I
Valve FW-1-RV-536 R0095634 480 volt circuit breaker preventative maintenance R0181384 Component Cooling Water Pump 1 1 auxiliary tube oil preventative
! maintenance R0180963 Component Cooling Water Pump 1-1 contactor preventative maintenance I Observations and Findinas l i The inspectors noted that the maintenance activities observed were performed properl Operators established clearance boundaries properly and performed risk assessments l as appropriate to the circumstances.
l M1.2 Surveillance Observations Insoection Scoce (61726)
The inspectors observed performance of all or portions of the following surveillance test procedures (STP):
STP P 23A " Acceleration Timing of Safety-Related Pumps Actuated by Solid State Protection System Train A," Revision 9 STP M-81 A " Diesel Engine Generator Inspection (Every Refueling Outage)," l Revision 16 l STP l-2C1 " Removal of Power Range Channel From Service," Revision 18 STP l-2C3 " Reinstatement of Power Range Channel to Service," Revision 19
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-4-STP l-2D " Nuclear Power Range incore/Excore Calibration," Revision 42 STP M-89A " Emergency Core Cooling System Venting," Revision 17 Observations and Findinas The inspectors found that the surveillance tests were conducted properly and in accordance with procedures, except as noted in Section M2.1 below. Risk assessments for portions of the surveillances that caused safety related systems to be inoperable were performed properl M2 Maintenance and Materiel Condition of Facilities and Equipment M2.1 Seismic Control Of Eauipment - Unit 1 insoection Scope (61726. 62707)
The inspectors observed personnel perform portions of Procedures STP l-2C1,
" Removal of Power Range Channel From Service," Revision 18; STP l-2C3,
" Reinstatement of Power Range Channel to Service," Revision 19; STP l-2D," Nuclear Power Range incore/Excore Calibration," Revision 42. In addition, the inspectors reviewed Procedure AD4.lD3," Seismic System Interaction Program Housekeeping Activities," Revision 3, and the " Seismically Induced Systems Interaction Manual,"
Revision Observations and Findinas On July 27,1999, the inspectors observed technicians perform a calibration of nuclear power range Channel N41. The inspectors questioned the technicians about the program for securing the cart and test equipment used for the calibration when the equipment was not being used and no personnel were present for extended periods of time, such as lunch breaks. The technicians indicated that since a calibration of the nuclear instruments was in progress, there was no requirement to seismically secure the equipmen The technicians concluded that restraining the cart was unnecessary when unattended because Procedure AD4.lD3 specified that, upon discovery of unrestrained equipment in the plant, action must be taken before the end of the shift to restrain the item. The inspectors noted that this statement covered the discovery of equipment placed in the plant by persons unknown and was not intended to relieve knowledgeable personnel of their responsibility to prevent seismic interaction The inspectors determined that Procedure AD4.lD3, step 5.1.1 required, in part, that individuals who bring transient equipment into the plant shall position or restrain the transient equipment so that it cannot impact and damage targets. Procedure AD4.lD3 defined transient equipment, in part, as tools and test equipment. A target is a structure, system, or component required during a specific plant operating mode or
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5-condition for safe cold shutdown and for mitigating certain accidents following an earthquake. The Seismically induced Systems Interaction Manualincluded Computer input Racks RNCII and RNCl2 as target equipment. The inspectors notej the cart and test equipment were located within a couple of inches of the computer inpJt rack Although the Seismically Induced Systems interaction Manual included computer Racks RNCl1 and RNCl2 as seismic targets, the licensee stated that Pncedure AD4.lD3 was not violated. Subsequent licensee analysis revealed that the equipment cart used in this case was made from a lightweight plastic and would not damage plant equipment during a seismic event. The inspectors reviewed the licensee analysis and determined that it was reasonabl In addition, the inspectors discussed this issue with Technical Maintenance supervisio Technical Maintenance personnel stated that the practice of leaving unattended maintenance equipment unrestrained near seismic targets for break periods was a common practice. The inspectors noted that this practice did not meet the intent of Procedure AD4.lD3. Licensee management stated that they would evaluate enhancements to worker practices with respect to seismic interaction of temporary equipment. This item is in the corrective action program as Action Request A048990 Conclusions The inspectors observed that technicians left a test cart unrestrained near a panel containing Seismic Category 1 equipment. The licensee subsequently determined that the test cart in question was made of lightweight plastic and would not damage plant equipment during a seismic event. However, the inspectors noted that the technicians erroneously believed that leaving unrestrained equipment near Seismic Category I equipment for less than a shift was an acceptable practice. The Technical Maintenance supervisor stated that technicians routinely implemented this improper practice. This item is in the corrective action program as Action Request A048990 M8 Miscellaneous Maintenance issues (92700,90712)
M8.1 (Closed) LER 275/1999-002-00: Technical Specification 3.3.1 not met because of inadequate knowledge and communicatio l NRC previously identified this issue as Noncited Violation 275; 323/99003-02, as ;
documented in NRC Inspection Report 50-275; 323/99-03, Section 0 M8.2 (Closed) LER 275/1999-001-00: Engineered safety features actuation during main turbine generator testing because of inadequate procedure guidanc NRC previously identified this issue as Noncited Violation 275/99003-03, as documented in NRC Inspection Report 50-275; 323/99-03, Section M M8.3 (Closed) LER 275: 323/1998-008-00: Technical Specification 3.7.3.1 not met because of personnel erro _ _ _ _ _ _ _ _ _ _ _ _
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6-During review of an industry operating event, the licensee determined on June 26,1998, that the component cooling water heat exchanger crosstie valves on Units 1 and 2 were not sealed or verified to be open within the 31-day interval specified in Technical Specification 4.7.3.1.a and Units 1 and 2 Procedures OP K-10E4, " Sealed Valve Checklist for Component Cooling Water Vital Headers A and B." The licensee determined that this condition had existed since initial startup of the units in February 1984 and July 1985, respectively. Corrective actions included: (1) verifying the valves were open and installing valve seals, as required, and (2) reviewing portions of other emergency core cooling systems to determine whether valves in flow paths were properly controlled. The licensee identified no discrepancies on the other systems and revised Procedure OP K 10E4 for each uni The failure to seal the component cooling water crosstie valves in the open position while in Mode 4 or above since February 1984 and July 1985 for Units 1 and 2, respectively, is a violation of Technical Specification 4.7.3.1. This Severity Level IV violation is being treated as the first example of a noncited violation for failure to proper y seal valves, consistent with Appendix C of the Enforcement Policy. The licensee included this deficiency'in the corrective action program as Nonconformance Report N0002065 (275; 323/99012-01, Example 1).
M8.4 (Closed) LER 275: 323/1997-008-00: Technical Specification 4.5.3.2 not met because of inadequate procedures.
! On July 25,1997, while verifying that test procedures required valve seal installation for low temperature overpressure protection, as specified in Action Request A0352439, operators identified specific surveillance procedures that did not have the required guidance. The licensee initiated Nonconformance Report N0002031 and determined from record reviews that Unit 2 operators had not complied with the requirements of Technical Specification 4.5.3.2 on October 20,1994 Procedure STP M 15," Integrated Test of Engineered Safeguards and Diesel Generators," directed that operators close the manual discharge valves for the SI pumps without specifying how they were to be administratively controlled. Consequently, operators closed and caution tagged the valves closed rather than seal the valves as specified by Technical Specification 4.5. The licensee properly sealed the valves as of July 25,199 The licensee attributed the root cause to inadequate procedure guidance. The corrective actions included: (1) specifying in design documents the specific valves that require seals to ensure compliance with Technical Specifications 4.1.2.3.2,4.1.2.4.2, and 4.5.3.2 and (2) revising appropriate procedures to ensure valve seals are specifie The inspectors confirmed the licensee had developed the design documents and sampled selected procedures to verify that valve seals were specified. The failure to seal manual discharge valves after testing on October 20,1994, resulted in a violation of Technical Specification 4.5.3.2. This Severity Level IV violation is being treated as the second example of a noncited violation for failure to properly seal valves, consistent with Appendix C of the Enforcement Policy. The licensee documented this deficiency in the corrective action program as Nonconformance Report N0002031 and Action Requests A0445533 and A0445549 (275; 323/99012-01, Example 2).
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7-M8.5 (Closed) LER 323/1998-004-00: Technical Specification 3.3.3.10, Action B, not met because of personnel erro On August 25,1998, a maintenance technician determined that the Unit 2 waste gas system oxygen analyzer alarm / trip functions were out of service during performance of Procedure STP l-24-A75.B. " Calibration Test of Waste Gas System Oxygen Analyzer Channels 75 and 76." The licensee determined that the waste gas system oxygen analyzer had been out of service for 26 days, completed Procedure STP l-24-A7 and restored the waste gas system oxygen analyzer to operable. Operators had verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during rounds that the oxygen concentration in the gaseous radioactive waste system had not exceeded 2 percent, as specified in Technical Specification 3.11.2.5. The licensee attributed the root cause to personnel error, determined that the restoration steps required clarification, and determined that the tailboard needed improvement. The licensee counseled the individual, clarified the procedure, and developed a case study emphasizing the importance of tailboard Technical Specification 3.3.3.10, Action B requires that grab samples be taken every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> whenever the alarm / trip function of the waste gas system oxygen analyzer is out of service. The failure to comply with the action requirement of Technical Specification 3.3.3.10, Action B resulted in a violation. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Policy. The licensee documented this deficiency in the corrective action program as Nonconformance Report N0002073 and Action Request A0467164 (323/99012-02).
M8.6 (Closed) LER 275: 323/1998-002-00: Technical Specification 4.3.1.1 not met because of inadequate surveillance test procedur On March 6,1998, a system engineer determined, during review of a Unit 2 relay failure (contacts failed to open on demand), that Procedure STP 1-9A, "12 kV Bus RCP U/F l and U/V Trip Actuating Device Operational Test," did not adequately verify operabilit )
Since Unit 2 was in an outage, the licensee completed the surveillance during refueling outage surveillance testing. Unit 1 operators entered Technical Specification 4.0.3 and satisfactorily completed the surveillance. The licensee determined that a revision to Procedure STP 1-9A in February 1997 had changed the methodology used to verify contact state because of test equipment concerns. Specifically, the licensee had added a light emitting diode to demonstrate contact position without recognizing that the output of the light emitting diode indicated actuation demand and not the actual contact positio Corrective actions included: (1) reviewing other procedures for similar problems, (2) rescinding the change to Procedure STP 1-9A, and (3) providing a case study of this event to surveillance authors and independent technical reviewers to emphasize lessons-leamed. The licensee determined that Technical Specification 3.3.1, Table 4.3-1, " Reactor Trip System Instrumentation Surveillance Requirements,"
Items 15 and 16 (tests of the reactor coolant pump undervoltage and underfrequency relays for the reactor trip system) had not been met since March 1997. The failure to comply with the requirements of Technical Specification 3.3.1 resulted in a violatio This Severity Level IV violation is being treated as a noncited violation, consistent with
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-8-Appendix C of the Enforcement Policy. The licensee included this item in the corrective action program as Nonconformance Report N0002057 and Action Request A0455728 (275;323/99012-03).
M8.7 (Closed) LER 275: 323/1997-017-00: Technical Specification 4.5.2.1.b not met because of inadequate procedure NRC previously identified this issue as Noncited Violation 275; 323/97019-03, as documented in NRC Inspection Report 50-275: 323/97-19, Section M Ill. Enoineerino E1 Conduct of Engineering E Modification of Auxiliary Saltwater (ASW) Pumo Vault Drain Lines Inspection Scope (37551)
The inspectors reviewed the 10 CFR 50.59 screening for a modification to the ASW vault drains in accordance with Design Change Package DCP P-4939 Observations and Findinas During Refueling Outage 1R9, the licensee implemented Design Change Package DCP P-49392 to modify the ASW pump vault drain lines. These drain lines included !
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check valves to prevent back flow into the ASW pump vaults in the event of a tsunami or internal intake structure flooding that were credited in the safety analysis. The ASW pump vault check valves had a previous history of sticking open because of foreign materialin the drain lines. To preclude the entry of this foreign material into the check valve seating surfaces, the licensee modified the drain lines to include y-strainers and l low-point drains upstream of the check valves.
l l As a part of the design change package, engineers performed a Licensing Basis impact Evaluation screen in accordance with Procedure TS3.lD2," Licensing Basis impact Evaluations," Revision 7. This procedure provided direction for determining if a 10 CFR 50.59 safety evaluation or other licensing basis evaluation was require l Procedure TS3.lD2 included four screening questions to determine if a safety evaluation was required. Any yes answer would result in a 10 CFR 50.59 safety evaluation. One of the questions stated,"Does it involve a change to the facility design, function, or method of performing the function as described in the safety analysis report, including text, tables, and figures, including the Fire Protection program and Quality Assurance Program?
The licensee answered no to each question, including the question above, and did not perform a safety evaluation. The licensee documented that the change required a revision to Final Safety Analysis Report (FSAR) Update Drawing 3-17, which indicated the ASW pump vault drain line configuration. The licensee concluded that: (1) the
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revision to the FSAR Update was insignificant, (2) the ASW pump vault drain lines were Class li, and (3) the intake flooding analysis was not changed, therefore, a safety evaluation was not require However, the inspectors noted that the licensee credited the Class 11 ASW pump vault drain line check valves with preventing back flow into the watertight ASW pump vaults during a design basis tsunami or internal flooding event. The installation of the y-stainers and low-point drains upstream of the check valves introduced a potential for bypassing the check valves. Specifically, if the y-strainers were not properly controlled, they could be inadvertently left open for cleaning, bypassing the passive check val /e .
Therefore, the inspectors concluded that the modification to the ASW pump vault c rain l
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lines was not insignificant or trivial and required a 10 CFR 50.59 safety evaluation implement. The inspectors informed the licensee of this concern, and the licenseo initiated an action reques The inspectors noted that 10 CFR 50.59 (b) (1) states that the licensee shall maintain records of changes to the facility to the extent that these changes constitute changes to the facility as described in the FSAR. These records must include a written safety evaluation that provides the bases to determine that the change does not involve an unreviewed safety question. The failure to perform a written safety evaluation for the modification to the ASW pump vault drain lines as described in the FSAR is a violation j of 10 CFR 50.59. However, this Severity Level IV violation is tung treated as a noncited violation, consistent with Appendix C of the Enforcement Policy. This item is in the corrective action program as Action Request A0487899 (275; 323/99012-04).
Subsequently, the license performed a Licensing Basis impact Evaluation, in accordance with Procedure TS3.lD2, to meet the requirements of 10 CFR 50.59. The licensee determined that leaving the strainers open following cleaning and thus permitting the bypassing of the ASW pump vault drain line check valves was highly unlikely, given the procedural and administrative controls over the strainers. Therefore, )
the possibility of a malfunction of a different type did not exist. No new accidents or consequences were identified. The inspector reviewed the Licensing Basis impact Evaluation and agreed with the licensee's conclusion The inspectors identified that Procedure TS3.lD2 contained questionable direction in explaining the screening criterion. Section 5.4.1 of Procedure TS3.lD2 stated that safety evaluations were not required for minor drawing or figure changes. The l
inspectors noted that no further definition of " minor" was provided. Using the direction I of Section 5.4.1, a preparer could screen out activities that made an important design j change to the f acility but required minimal effort to change the FSAR Update, such as in ;
the case of the modification to the ASW pump vault drain line check valves. The i l
licensee was evaluating enhancements to Procedure TS3.lD2 at the end of the inspection period, Conclusions l l
A violation (EA 99-219) of 10 CFR 50.59 was identified for failure to perform a safety evaluation for a change to the facility. Specifically, in March 1999, the licensee
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l- - 10-modified the ASW pump vault drain lines as described in FSAR Update Drawing 3-17 but did not perform a written safety evaluation to determine if an unreviewed safety question existed. The licensee installed y-strainers and low-point drains upstream of the ASW vault drain line check valves, which had the safety function of mitigating the consequences of a design basis tsunami. The licensee subsequently performed a safety evaluation that demonstrated that an unreviewed safety question did not exis This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the Enforcement Policy. This item is in the corrective action program as Action Request A048789 E8 Miscellaneous Engineering issues (90712,92700)
E8.1 (Closed) LER 323/1998-001-00: defect in Class 1 piping spoo On December 17,1997, a welder questioned the black color of a 1%-inch nominal diameter, stainless steel ASME Class 1 piping spool. The welder v/as preparing to install the pipe in a high head Si cold-leg branch line. The licensee had this one-of-a-kind piping spool fabricated by a vendor who had the capabilities to upgrade the pipe from ASME Class 2 stoc The licensee initiated Nonconformance Report N0002046 to identify a root cause and affect corrective actions. The licensee determined that the piping spool had not been properly cleaned to ensure corrosion resistance following the bending and annealing process. Proper cleaning allows for formation of a passive layer that assures resistance to corrosion. The licensee attributed the root cause to lack of clear technical requirements in the purchase order and f ailure of the vendor to document / communicate needed clarifications. In addition, the vendor informed the licensee that some discoloration was possible, misleading procurement personnel into believing that the black color was acceptable. As immediate corrective actions, the licensee had the pipe properly cleaned and prepared, prior to installing the pipe in February 1998. Long term corrective actions included: (1) auditing other material shipped from the vendor; (2) developing additional guidance for procurement personnel; and (3) performing an audit at the vendor facilities. The licensee submitted a 10 CFR Part 21 report for this defec E8.2 (Closed) LER 275: 323/1998-001-01 and -00: reactor coolant system outside design basis for inadvertent emergency core cooling system actuation at power because of nonconservative assumptions in pressurizer safety valve operation.
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On January 22,1998, the licensee (after being notified by the nuclear steam supply l
system vendor) determined that the reactor coolant system was outside its design basis.
l Specifically, the vendor indicated that nonconservative assumptions, related to a l temperature coefficient in a calculation, modified the response of the pressurizer safety valves for a spurious actuation of the Si system at power. The vendor determined that the alpha temperature coefficient was not a constant, as previously used in WCAP 11677," Safety Relief Valve Operation for Water Discharge During a Feedwater Line Break," but varied with temperature. The vendor applied this new data to the i
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l pressurizer safety valves and determined the nonconservative calculation error resulted in an increase in the lowest temperature of saturated water allowed to pass through the pressurizer safety valves. The valves were initially analyzed for 603*F; however, the new low temperature was 613" The licensee had the vendor run several cases to determine interim measures to prevent passing water through the pressurizer safety valves at too cool of a temperature. The licensee determined that the increased minimum temperature value eliminated the ability of the operators to use manual actions; consequently, the licensee implemented administrative control for the positive displacement pumps when Units 1 and 2 operated in Modes 1 and 2. The licensee will only operate the positive displacement pumps in Modes 1 and 2 to verify operability. This administrative control reduces the rate of fill for the pressurizer and allows time for operators to secure the centrifugal charging pumps if a spurious actuation occur The licensee included this issue in the corrective action program as Nonconformance Report N0002048 and Action Request A0449600. As long-term corrective action, the licensee planned to submit a license amendment request to credit power operated relief valve operation. The power operated relief valves lif t at lower pressures than the pressurizer safety valves and can be blocked if they were to fail. The licensee will restrict operation of the positive displacement pumps until a license amendment request is approved. The inspectors concluded that the licensee implemented appropriate corrective action E8.3 (Closed) LER 323/1997-006-01 and -00: nuclear fuel potentially outside of the manufacturer's fuel design criteri On October 28,1997, the nuclear fuel vendor informed the licensee about concerns with fuel rod design criteria for burnable absorber rods with Zircaloy-4 cladding. Specifically, the fuel vendor indicated that the amount of fuel cladding oxidation the vendor had calculated was nonconservative because the vendor had not accounted for pellet clad reopening resulting from greater than predicted internal gas pressure. Without site specific analysis, the vendor reported that the increased fuel cladding oxidation could exceed the 17-percent maximum fuel cladding oxidation limit specified in 10 CFR 50.4 The fuel vendor established a 12-percent oxidation limit as a preaccident screening criteria for assessing compliance to the 17-percent maximum oxidation criterion. The 12 percent limit was based on modeling that established a maximum of 5 percent fuel cladding oxidation during loss-of-coolant accidents. The inspectors discussed the vendor analysis with personnelin the Office of Nuclear Reactor Regulation. The inspectors determined 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 27,199 Diablo Canyon Units 1 and 2 cores had recently included a large percentage of integrated fuel burnable absorber rods with Zircaloy-4 cladding. Therefore, the licensee required specific analyses for past, current, and future core loads. For past Unit 2 Cycle 8. the vendor performed a site-specific analysis that demonstrated that the
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12-preaccident screening criteria was exceeded (12.5 percent). However, the vendor determined that the increase in oxidation resulting from a loss-of-coolant accident would
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only increase by 3.3 percent, so the total oxidation did not exceed the 17-percent -
criterion. A similar analysis demonstrated that oxidation criterion was also not exceeded for Unit 1 Cycle 9. For current and future core loads, the vendor analyses demonstrated that the 12-percent screening criterion would not be exceede During onsite inspection, the inspectors reviewed the operability evaluations that had been performed to support each operating cycle for each unit. After Unit 1 Cycle 9 and Unit 2 Cycle 8, the chance that the 17 percent oxidation criterion would be exceeded l decreased because the licensee continued to replace the Zircaloy-4 clad integral fuel burnable absorber rods with ZlRLO integral fuel burnable absorber rods, which were less susceptible to oxidation. The licensee took action to decrease the percentage of oxidation including: (1) using annular pellets to increase the available volume within the fuel rods, (2) planning to use longer fuel rods to increase the available volume within the fuel rods, and (3) using a lower enrichment of the boron coating for the integral fuel burnable absorber, in summary, for integral fuel burnable rods with Zircaloy-4 cladding, new data demonstrated greater than the design-predicted internal gas pressure and cladding oxidation. This new data resulted in decreased margin to exceeding the fuel rod l cladding oxidation limit. However, site specific data demonstrated that the cladding L oxidation regulatory limit would not be exceeded during a loss-of cooling accident. The l inspectors concluded that the licensee's evaluation and corrective actions demonstrated
! a sensitivity to safety and good engineering performanc IV. Plant Support l
l l R1 Radiological Protection and Chemistry Controls R1.1 General Comments (71750)
The inspectors evaluated radiation protection practices during plant tours and work observation. The inspectors determined that personnel donned protective clothing and dosimetry properly, and that radiological barriers were properly maintained, with two exceptions.
l On July 15,1999, the inspectors identified that Centrifugal Charging Pump 1-1 lobe oil leaked inside of the surface contamination boundary, but spread to nearby clean area The inspectors notified radiation protection personnel, who cleaned up the potentially contaminated flui On August 5, during a tour of the facility, the inspectors identified a degraded surface contamination area boundary. Near the centrifugal charging pump room, one of the redundant surface contamination area signs and boundaries was face down on the deck. Adjacent to the degraded boundary, the area was properly posted. The inspectors notified radiation protection personnel, who corrected the degraded surface
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13-contamination radiation area boundary. The licensee initiated an event trending record to trend this occurrence of a degraded radiological posting. The inspectors concluded that the licensee's actions were appropriat P1 Conduct of Emergency Planning Activities P1.1 Loss of Emeraency Sirens Inspection Scope (71750)
The inspectors evaluated the response to Action Request A0489305, which discussed a significant loss of the emergency siren Observations and Findinas i
On August 6,1999, at 2 a.m., the Early Warning Siren System computer failed. This rendered all of the emergency sirens for San Luis Obispo County inoperable. However, the alarm for the computer failure did not actuate until 6:05 a.m. When the alarm was received, the watch officer at the emergency operations facility attempted to contact the appropriate maintenance personnel but only left a message. The technician did not receive the message until 7:30 a.m. Upon receiving the message that the Early Warning Siren System computer had failed, the technician went to the emergency operations facility and placed the backup computer on line, restoring the capability of initiating the emergency sirens. The licensee initiated an action request to place this item into the corrective action syste Licensee review of the event revealed that the capability to sound the emergency sirens had been lost for several hours. Subsequently, the licensee reporteo this event as a l major loss of offsite response capability. The licensee was considering improvements to l the communications of the offsite personnel with respect to the emergency sirens at the l end of the inspection period. The inspectors considered licensee disposition of this issue appropriate.
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I Conclusions Because of inadequate communications among off-site licensee personnel, the county emergency sirens were inoperable for an excessive period of time. The computer that initiates the emergency warning system failed but was not restored to operable status immediately. The alarm signifying that the computer failed annunciated in the emergency operations f acility but was not f ully communicated to repair personnel for 1% hours. The licensee appropriately investigated and dispositioned this issu .
-14-S1 Conduct of Security and Safeguards Activities S General Comments (71750.)
During routine tours, the inspectors noted that the security officers were alert at their posts, security boundarss were being maintained properly, and screening processes at the Primary Access Point were performed well. During backshift inspections, the inspectors noted that the protected area was properly illuminated, especially in areas
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where temporary equipment was brought i V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on August 24,1999. The licensee acknowledged the findings presente 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 SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee J. R. Becker, Manager, Operations Services W. G. Crockett, Manager, Nuclear Quality Services R. D. Gray, Director, Radiation Protection T. L. Grebel. Director, Regulatory Services D. B. Miklush, Manager, Engineering Services D. H. Oatley, Vice President and Plant Manager R. A. Waltos, Manager, Maintenance Services L. F. Womack, Vice President, Nuclear Technical Services INSPECTION PROCEDURES USED IP 37551 Onsite Engineering IP 61726 Surveillance Observations IP 62707 Maintenance Observation
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IP 71707 Plant Operations IP 71750 Plant Support Activities IP 90712 in-Office Review of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92700 Onsite Followup of Written Reports of Nontoutine Events at Power Reactor Facilities IP 92901 Followup - Operations IP 92902 Followup - Maintenance
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IP 92903 Followup - Engineering l
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ITEMS OPENED AND CLOSED
- Ooened None.
I Closed 323/1999-001-00 LER Voluntary entry into Technical Specification 3.0.3 to open the containment recirculation sump sensor hatch to verify level transmitter operability (Section 08.1)
275/1999-002-00 LER Technical Specification 3.3.1 not met because of inadequate knowledge and communication (Section M8.1)
275/1999-001-00 LER Engineered safety features actuation during main turbine generator testing because of inadequate procedure guidance (Section M8.2)
275;323/ LER Technical Specification 3.7.3.1 not met because of 1998-008-00 personnel error (Section M8.3)
275;323/ LER Technical Specification 4.5.3.2 not met because of 1997-008-00 inadequate procedures (Section M8.4)
323/1998-004-00 LER Technical Specification 3.3.3.10, Action B not met because of personnel error (Section M8.5)
275;323/ LER Technical Specification 4.3.1.1 not met because of 1998-002 00 inadequate surveillance test procedure (Section M8.6)
275;323/ LER Technical Specification 4.5.2.1.b not met because of 1997-017-00 inadequate procedures (Section M8.7)
323/98-001 00 LER Defect in Class 1 piping spool (Section E8.1) i 275; 323/ LER Reactor coolant system outside design basis for inadvertent !
1998-001-01 and emergency core cooling system actuation at power 00 because of nonconservative assumptions in pressurizer safety valve operation (Section E8.2)
323/1997-006-01 LER Nuclear fuel potentially outside of the manufacturer's fuel and -00 design criteria (Section E8.3)
Ooened and Closed 275:323/99012-01 NCV Two examples of violation for failure to properly seal valves in accordance with Technical Specifications 4.5.3.2 and 4.7.3.1 (Sections M8.3 and M8.4)
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3-323/99012-02 NCV Technical Specification 3.3.3.10 violation for failure to take grab samples when waste gas oxygen analyzer was inoperable (Section M8.5)
275:323/99012-03 NCV Technical Specification 4.3.1.1 not met because of inadequate procedure (Section M8.6)
275;323/99012-04 NCV Violation of 10 CFR 50.59 for failure to perform a safety evaluation upon modifying the ASW vault drain lines (Section E1.1)
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l LIST OF ACRONYMS USED ASME American Society of Mechanical Engineers
ASW auxiliary saltwater
! FSAR Final Safety Analysis Report t-LER licensee event report l
NCV noncited violation (. NRC- Nuclear Regulatory Commission RHR residual heat removal SI safety injection STP surveillance test procedure
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