IR 05000275/2006011
| ML070560003 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/23/2007 |
| From: | William Jones NRC Region 1 |
| To: | Keenan J Pacific Gas & Electric Co |
| References | |
| IR-06-011 | |
| Download: ML070560003 (24) | |
Text
ary 23, 2007
SUBJECT:
DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2, NRC COMPONENT DESIGN BASIS INSPECTION REPORT 05000275/2006011; 05000323/2006011
Dear Mr. Keenan:
On January 11, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Diablo Canyon Nuclear Power Plant, Units 1 and 2. The preliminary findings were discussed on October 6, 2006, with you and other members of your staff. After additional in-office inspection, a final telephonic exit meeting was conducted on January 11, 2007, with Ms. D. Jacobs, Vice President, Nuclear Services, and other members of your staff. The enclosed report documents the inspection findings.
The team examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Based on the results of this inspection, the NRC has identified two issues that were evaluated under the risk significance determination process as having very low safety significance (green). The NRC has also determined that violations are associated with these issues. These violations are being treated as noncited violations, consistent with Section VI.A of the Enforcement Policy. These noncited violations are described in the subject inspection report. If you contest the violation or significance of these noncited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Diablo Canyon Nuclear Power Plant, Units 1 and 2.
Pacific Gas and Electric Company -2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
William B. Jones, Chief Engineering Branch 1 Division of Reactor Safety Dockets: 50-275; 50-323 Licenses: DPR-80; DPR-82
Enclosures:
Inspection Report 05000275/2006011; 05000323/2006011 w/Attachment Supplemental Information
REGION IV==
Dockets: 50-275; 50-323 Licenses: DPR-80; DPR-82 Report No.: 05000275/2006011; 05000323/2006011 Licensee: Pacific Gas and Electric Company Facility: Diablo Canyon Nuclear Power Plant, Units 1 and 2 Location: 7 1/2 miles NW of Avila Beach Avila Beach, California Dates: September 11, 2006, through January 11, 2007 Team Leader: W. Sifre, Senior Reactor Inspector, Engineering Branch 1 Inspectors: P. Gage, Senior Operations Engineer R. Lantz, Senior Emergency Preparedness Inspector J. Nadel, Reactor Inspector, Engineering Branch 1 J. Reynoso, Reactor Inspector, Engineering Branch 1 Contractors: F. Baxter, Electrical, Beckman and Associates M. Shlyamberg, Mechanical, Nuenergy, Inc.
Accompanied By: A. Fairbanks, Reactor Inspector (NSPDP)
M. Young, Reactor Inspector (NSPDP)
Approved By: William B. Jones, Chief Engineering Branch 1 Division of Reactor Safety-1- Enclosure
SUMMARY OF FINDINGS
IR 05000275/2006011; 05000-323/2006011; September 11 through October 6, 2006; Diablo
Canyon Nuclear Power Plant, Units 1 and 2: NRC Inspection Procedure 71111.21, Component Design Basis Inspection.
The report covered a period of inspection by a team of seven inspectors and two contractors.
Two findings of very low safety significance were identified. The significance of most findings is indicated by its color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609,
Significance Determination Process. Findings for which the significance determination process does not apply may be green or be assigned a severity level after NRC management review.
The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
NRC-Identified Findings
Cornerstone: Initiating Events
- Green.
The team identified a noncited violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, for the failure to translate design basis information into specifications and procedures. The team identified that a nonconservative flow rate was used as an input in engineering design calculations resulting in the potential for choked flow at the discharge valves for the Unit 1 auxiliary service water system. Choked flow turbulence is a wear concern for these components, and can result in system failure.
The licensee entered this finding into their corrective action program as Action Requests A0678338 and A0678472.
The finding is more than minor because the error affected the Mitigating System Cornerstone objective (Design Control attribute) of ensuring availability, reliability, and capability of the auxiliary service water systems to respond to initiating events to prevent undesired consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the issue screened as having very low safety significance because 1) did not represent a loss of system safety function; and 2) did not represent an actual loss of safety function of one or more non-technical specification trains of equipment; and did not screen as potentially risk significant because of a seismic, flooding, or sever weather initiating event (Section 1R21b.1.).
- Green.
The team identified a noncited violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, for the failure to demonstrate that the acceptance criteria for surveillance tests had conservatively accounted for uncertainties in determination of the minimum allowed ultimate heat sink temperature. Specifically, the team identified that the acceptance criteria specified in the Surveillance Test Procedure STP I-1A,
Routine Shift Checks Required by the Licensee, Revision 101, did not correctly account for instrument uncertainty. The licensee entered this finding into their corrective action program as Action Request A0682398.
The finding is more than minor because the error affected the Mitigating System cornerstone objective (Design Control attribute) of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Using the Manual Chapter 0609, Significance Determination Process,
Phase 1 screening worksheet, the issue screened as having very low safety significance because 1) did not represent a loss of system safety function; and 2) did not represent an actual loss of safety function of one or more non-technical specification trains of equipment; and did not screen as potentially risk significant because of a seismic, flooding, or severe weather initiating event (Section 1R21b.2).
B. Licensee-Identified Findings None.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events/Mitigating Systems/Barrier Integrity
REACTOR SAFETY
Inspection of component design bases verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected components and operator actions to perform their design bases functions. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The plant risk assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems and Barrier Integrity cornerstones for which there are no indicators to measure performance.
1R21 Component Design Bases Inspection
The team selected risk-significant components and operator actions for review using information contained in the licensees probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1E-6.
a. Inspection Scope
To verify that the selected components would function as required, the team reviewed design basis assumptions, calculations, and procedures. In some instances, the team performed independent calculations to verify the appropriateness of the licensee engineers' analysis methods. The team also verified that the condition of the components was consistent with the design bases and that the tested capabilities met the required criteria.
The team reviewed maintenance work records, corrective action documents, and industry operating experience information to verify that licensee personnel considered degraded conditions and their impact on the components. For the review of operator actions, the team observed operators during simulator scenarios associated with the selected components simulated actions in the plant.
The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions because of modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items, such as, failed performance test results; significant corrective actions; repeated maintenance; 10 CFR 50.65(a)1 status; operable, but degraded, conditions; NRC
resident inspector input of problem equipment; system health reports; industry operating experience; and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins.
The inspection procedure requires a review of 15-20 risk-significant and low design margin components, three to five relatively high-risk operator actions, and 4 to 6 operating experience issues. The sample selection for this inspection was 18 components, 8 operator actions, and 6 operating experience items.
The components selected for review were:
- Component cooling water surge tank
- Component cooling water motor-operated valves
- Component cooling water piping integrity
- Component cooling water heat exchangers
- Component cooling water check valves
- Switchgear ventilation (fans, louvers, etc.), 480v, (Fans 43,44)
- 125vdc batteries (capacity, fuses)
- Emergency diesel generators (engines, output breakers)
- Battery chargers (electrolytic capacitors)
- Auxiliary feedwater turbine driven pumps
- Auxiliary salt water (ultimate heat sink) traveling screens
- Auxiliary salt water pumps
- Auxiliary salt water motor-operated valves
- Auxiliary salt water check valves
- Power operated relief valves
- Residual heat removal containment isolation valves The selected operator actions were:
- Response to a spurious safety injection actuation at power
- Transfer to cold leg recirculation with failure of one residual heat removal pump to trip
- Transfer to hot leg recirculation
- Isolation of design leakage into the component cooling water system
- Reduction of component cooling water heat loads because of a malfunction in the component cooling water system, establishment of an alternate cooling source for the component cooling water system
- Diagnosis and response to a loss of secondary heat sink
- Establishment of alternate cooling to the 480 Volt AC switchgear
- Restoration of auxiliary saltwater cooling via the unit auxiliary salt water system cross-connect valve (FCV-601) during a station blackout The operating experience issues were:
- Auxiliary feedwater pump recirculation line orifice fouling - potential common cause failure at Point Beach
- Sizing and setting of molded case circuit breakers and thermal overload heaters
- Safe shutdown potentially challenged by unanalyzed internal flooding events
- Vibration induced degradation of butterfly valves (Fisher valves)
- Guidance on developing acceptable inservice testing programs
- Periodic verification of design-basis capability of safety-related motor-operated valves
b. Findings
b.1. Failure to Use Correct Design Inputs in Determination of a Potential for Choking Flow/Cavitation Across the Auxiliary Service Water Throttled Butterfly Valves
Introduction.
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Critieria III, Design Control, for the failure to translate design basis information into specifications and procedures. Specifically, the team identified that a non-conservative flow rate was used as an input in engineering design calculations (M-988, ASW System Flows, Pressures and Temperatures, Revision 6 and Surveillance Test Procedure 1&2STP-M 26, ASW System Flow Monitoring, Revision 25B and M-885, Determine ASW System Flow in Various ASW/CCW Configurations and Conditions, Revision 3).
Description.
The team identified that conditions for choked flow in the auxiliary salt water system are established when the auxiliary salt water system is aligned in a two pump and one heat exchanger configuration. Since mid-1990 the auxiliary salt water trains have been aligned in this configuration when the ultimate heat sink temperature was in excess of 64°F and under an accident condition because of the normally open cross-train connecting Valves FCV-495 and FCV-496. The team determined that the auxiliary salt water system maximum auxiliary salt water system flow rate across the component cooling water heat exchanger outlet throttled butterfly valves following a design basis event would be significantly higher than the licensee assumed in Calculation M-988. In response to the team's questions the licensee issued Action Requests A0678338, Calculation M-988 Discrepancies, September 26, 2006; and A0678472, Incipient Cavitation of ASW Operation w/2PP, 1HX, September 27, 2006.
Action Request A0678472 documented that Calculation M-988 used a non-conservative flow rate for the cavitation determination and changed Unit 1 valves from 55 degrees open to greater than 70 degrees open (Unit 2 valve positions were greater than 70 degrees open). The licensee provided the following justification for the operability of the valves under choking conditions. "With 1 minute of 2 pumps, one heat exchanger operation per swap, and 50 swaps per year, we have accumulated at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation in the alignment of concern over the last decade. The maintenance history of these valves does not indicate any unusual damage or parts replaced because of cavitation damage. In addition, these valves are inspected when the heat exchangers are opened for cleaning every outage and no signs of damage or degradation that would interfere with valve operation have been noted." The team determined that the evidence of no "unusual damage or parts replaced because of cavitation damage" was based on a system operation under normal conditions where the auxiliary service water temperature is rarely above 70EF. Under accident conditions for which the choked flow was predicted, the auxiliary service water temperature was expected to be at least 140EF. An elevated temperature could significantly increase the cavitation potential.
Also given the transient nature of the swap over, the time period is not sufficiently long enough to evaluate for a cumulative effect. Based on the team's review of this action request, the licensee revised the operability determination and issued Action Request A0679175, Review Performance of Operability Determinations During CDBI, October 5, 2006.
Analysis.
The failure to use a conservative design input in the engineering analysis was a performance deficiency. This issue is more than minor because the error affected the Mitigating System cornerstone objective (Design Control attribute) of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it did not represent a loss of system safety function and it did not represent an actual loss of safety function of one or more non-technical specification trains of equipment; and it did not screen as potentially risk significant because of a seismic, flooding, or severe weather initiating event.
Enforcement.
Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established to assure that design basis are correctly translated into specifications and procedures. Contrary to the above, in Calculation M-988, the licensee did not use a conservative auxiliary service water flow rate to determine the outlet valve position necessary to prevent choke flow and subsequent component damage or even a catastrophic failure of the valves during a design basis event.
Because the finding is of very low safety significance (Green) and has been entered into the licensee's corrective action program (Action Requests A0678338 and A0678472), it is a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy:
NCV 05000275; 323/2006011-01, Failure to Use Correct Design Input in Determination of a Potential for Chocking Flow/Cavitation Across the Auxiliary Service Water Throttled Butterfly Valves.
b.2. Failure to Consider Instrument Uncertainty in Surveillance Requirements for Technical Specifications LCO 3.7.9
Introduction.
A noncited violation of very low safety significance (Green) was identified for the failure to demonstrate that the acceptance criteria for surveillance tests had conservatively accounted for uncertainties in the determination of the minimum allowed ultimate heat sink temperature. Specifically, the team identified that the acceptance criteria specified in the Surveillance Test Procedure STP I-1A, Routine Shift Checks Required by the Licensee, Revision 101, did not correctly account for the instrument uncertainty.
Description.
Technical Specifications LCO 3.7.9 requires placing the second closed loop cooling system heat exchanger in service if the ultimate heat sink temperature exceeds 64EF. The team's review of Calculation Support STP I-1A, Indicated Temperature Uncertainties for TI-311 /-328, Revision 0, identified that the instrument correction for the maximum ultimate heat sink temperatures specified in this calculation and in the Surveillance Test Procedure STP I-1A (SR 3.7.9.2) were non-conservative.
The licensee's correction for the instrument uncertainty value was based on a standard deviation and not 2.4EF, the actual uncertainty of an individual instrument loop. The licensee used an instrument uncertainty correction of 0.9EF when two instruments were available and 0.4EF when three or four instruments were available. Use of standard deviation methodology is not an NRC and industry recognized methodology and was not in accordance with the licensee's programs and procedures for setpoint uncertainties (CF6.ID1, Setpoint Control Program; CF6.NE1, Instrument Channel Uncertainty and Setpoint Methodologies; AWP E 001, Development of PME Channel Uncertainty Calculations). Furthermore, use of standard deviation in this case is not scientifically valid, since this method accounts for only the statistical performance of the individual loops and cannot be used because of the variation in the monitored variable -
temperatures of the separate temperature channels. Therefore, the surveillance correction factors were non-conservative. The licensee issued Action Request A0682398, Evaluate UHS Instrument Uncertainty Used in STP I-1A, November 14, 2006. This action request stated that the STP acceptance criteria will be revised to a value that is based on the approved methodologies.
Analysis.
The failure to use a conservatively determined instrument uncertainty in the derivation of the acceptance criteria for the technical specifications surveillance values was a performance deficiency. This issue is more than minor because the error affected the Mitigating System cornerstone objective (Design Control attribute) of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Using the Manual Chapter 0609,Phase 1 screening worksheet, the issue screened as having very low safety significance because it did not represent a loss of system safety function and it did not represent an actual loss of safety function of one or more non-technical specification trains of equipment; and did not screen as potentially risk significant because of a seismic, flooding, or severe weather initiating event.
Enforcement.
Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, states, in part, that measures shall be established to assure that design basis are correctly translated into specifications and procedures. Contrary to the above, the licensee did not conservatively account for the effect of the instrument uncertainty in derivation of the acceptance criteria for the technical specifications surveillance values for Technical Specification LCO 3.7.9; thus, the maximum allowable ultimate heat sink temperature for a single component cooling water heat exchanger operation could not be guaranteed at the Technical Specification LCO values.
Because the finding is of very low safety significance (Green) and has been entered into the licensee's corrective action program (Action Request A0682398), it is a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000275/323/2006011-02, Failure to Consider Instrument Uncertainty in Surveillance Requirements for Technical Specifications LCO
OTHER ACTIVITIES
4OA6 Meetings, Including Exit
On January 11, 2007, the team leader presented the inspection results, via telephone, to, Ms. D. Jacobs, Vice President, Nuclear Services and other members of the Diablo Canyon Power Plants staff who acknowledged the findings. The inspectors confirmed that proprietary information was provided and examined during this inspection.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- T. Baldwin, Supervisor, Engineering
- J. Ballard, Engineer
- T. Chitwood, Senior Operations Engineer
- C. Dougherty, Senior Engineer, Regulatory Services
- J. Fields, Auditor, Quality Verification
- L. Fuseo, Manager, Engineering Services
- C. Harbor, Manager, Performance Improvement
- D. Jacobs, Vice President, Nuclear Services
- R. Klimezak, Manager, Engineering Services
- M. Mayer, Engineering Supervisor
- P. Nugent, Manager, Project Engineering
- L. Parker, Supervisor, Regulatory Services
- P. Roller, Director, Performance Improvement
- K. Schrader, Engineer, Regulatory Services
- R. Washington, Instrumentation and Controls Supervisor
- R. West, Engineering Supervisor
- S. Westcott, Manager, Engineering Services
- S. Zawalick, Engineer, Regulatory Services
NRC personnel
- T. Jackson, Senior Resident inspector
- T. Brown, Resident Inspector
- T. McConnell, Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000275; 323/2006011-01 NCV Failure to Use Correct Design Inputs in Determination of a Potential for Choking Flow/Cavitation Across the Auxiliary Service Water Throttled Butterfly Valves
- 05000275; 323/2006011-02 NCV Failure to Consider Instrument Uncertainty in Surveillance Requirements for Technical Specifications LCO 3.7.9 Attachment