IR 05000483/2013005
| ML14038A380 | |
| Person / Time | |
|---|---|
| Site: | Callaway  (NPF-030) |
| Issue date: | 02/06/2014 |
| From: | O'Keefe N NRC/RGN-IV/DRP/RPB-B |
| To: | Diya F Union Electric Co |
| O'Keefe N | |
| References | |
| IR-13-005 | |
| Download: ML14038A380 (75) | |
Text
February 6, 2014
SUBJECT:
CALLAWAY PLANT - NRC INTEGRATED INSPECTION REPORT 05000483/2013005
Dear Mr. Diya:
On December 31, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Callaway Plant. On January 8, 2014, the NRC inspectors discussed the results of this inspection with Mr. C. Reasoner, Vice President, Engineering, and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.
NRC inspectors documented three findings of very low safety significance (Green) in this report.
All three of these findings involved violations of NRC requirements. Further, inspectors documented a licensee-identified violation which was determined to be of very low safety significance. The NRC is treating this violation as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy.
If you contest the violations or significance of these non-cited 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, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Callaway Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV; and the NRC resident inspector at the Callaway Plant.
U N IT E D S TA TE S NUC LEAR RE GULATOR Y C OM MI S SI ON R E G IO N I V 1600 EAST LAMAR BLVD AR L INGTON, TEXAS 76011-4511 In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management 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/
Neil OKeefe, Branch Chief Project Branch B Division of Reactor Projects
Docket Number: 50-483 License Number: NPF-30
Enclosure: Inspection Report 05000483/2013005 w/ Attachment 1: Supplemental Information
Attachment 2: Radiation Safety Inspection Request for Information
cc w/ encl: Electronic Distribution
SUMMARY
IR 05000483/2013005; 09/28/2013 - 12/31/2013; Callaway Plant, Integrated Resident and
Regional Report; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications, Surveillance Testing, and Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and Transportation.
The inspection activities described in this report were performed between September 28, and December 31, 2013, by the resident inspectors at the Callaway Plant and inspectors from the NRCs Region IV office. Three findings of very low safety significance (Green) are documented in this report. All of these findings involved violations of NRC requirements. Additionally, NRC inspectors documented one licensee-identified violation of very low safety significance. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310,
Components Within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a non-cited violation of Technical Specification 5.4.1(d),
Procedures, which states, in part, that written procedures shall be established, implemented, and maintained covering fire protection program implementation. Specifically, prior to October 10, 2013, Procedure OTO-ZZ-00001, Control Room Inaccessibility,
Revision 38, steps D31 and D32, directed operators to use incorrect process indication and action criteria to diagnose and prevent icing conditions in the cooling tower after the control room was evacuated. This condition was entered into the corrective action program as Callaway Action Request 201307709. As a result, the licensee revised the action criteria to direct the operators to take action to prevent icing in the cooling tower.
The inspectors determined that the failure of Procedure OTO-ZZ-00001 to contain the appropriate criteria for operator action to control the ultimate heat sink cooling tower bypass valve after a control room evacuation was a performance deficiency. The inspectors determined that the performance deficiency was associated with the Mitigating Systems Cornerstone and was more than minor, and therefore a finding, because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the action criteria would not direct the operators to correctly diagnose, prevent, or mitigate icing in the ultimate heat sink cooling tower after a control room evacuation. Using NRC Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding had very low safety significance (Green) because this finding would not have prevented the reactor from reaching and maintaining safe shutdown.
This finding was not assigned a cross-cutting aspect because the most significant contributor was not reflective of current licensee performance. (Section 1R17.2)
- Green.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, Corrective Action, for a failure to correct a condition adverse to quality.
Specifically, the licensees calculations to determine acceptable oil leak rates of safety related pumps in response to Callaway Action Requests 201102434 and 201207677 failed to ensure that the pumps mission times could be met. As a result, non-conservative limits were added to Procedure ODP-ZZ-0016E, Appendix 1, Operations Technician General Inspection Guide. Specifically, the licensee revised the leakage limits to correctly account for the mission time, but failed to account for the time between observations during operator rounds. The licensee entered this issue into the corrective action program in Callaway Action Request 201308127. NRC inspectors determined there was never a time within the last several years when any of these pumps exceeded the revised oil leakage limits, once corrected.
The inspectors determined that the failure to correct non-conservative oil leakage specifications used in operator logs is a performance deficiency. The inspectors determined that the performance deficiency was associated with the Mitigating Systems Cornerstone and was more than minor, and therefore a finding, because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to evaluate and determine appropriate lube oil leakage to maintain safety related equipment could have impacted the availability of mitigating systems if left uncorrected. The finding was assessed using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and was determined to be of very low safety significance since the inadequate specifications did not result in a loss of operability for the affected systems. This finding has a cross-cutting aspect in the decision-making component of the human performance cross-cutting area because the licensee failed to utilize a systematic process to ensure safety is maintained, specifically the authority and roles for decisions affecting nuclear safety in the corrective actions for Callaway Action Requests 201102434 and 201207677 were not formally defined resulting in inadequate engineering conclusions being incorporated into operations procedures H.1(a). (Section 1R22)
Cornerstone: Public Radiation Safety
- Green.
The inspectors reviewed a self-revealing non-cited violation of 10 CFR 71.5(a) for the failure to describe a radioactive material shipment correctly and the failure to include all required hazard communication information in a radioactive shipment document. The licensee entered this issue into the corrective action program as Callaway Action Request 201204454 and licensee representatives stated procedure guidance will be revised as a corrective action.
Failure to correctly describe a radioactive material shipment correctly and to include all required hazard communication information in accordance with federal hazardous material transportation regulations was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process (transportation program) and adversely affected the associated cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation, in that it resulted in the incorrect identification and quantification of radioactive material transported in the public domain. Using Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, the inspectors determined this to be of very low safety significance (Green) because the violation did not involve the radioactive effluent release program or the radiological environmental monitoring program, but the violation did involve the transportation of radioactive material. The violation was not (1) in excess of radiation limits, (2) a breach of package during transit, (3) a certificate of compliance issue, (4) a low-level burial ground noncompliance, or (5) a failure to make notifications or provide emergency information. The violation has a human performance cross-cutting aspect associated with licensee resources, in that the licensee had insufficient instruction in its shipping guidance to ensure package contents were verified H.2(c).
(Section 2RS08)
Licensee-Identified Violations
A violation of very low safety significance that was identified by the licensee has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and associated corrective action tracking numbers are listed in Section 4OA7 of this report.
PLANT STATUS
Callaway operated at 100 percent power for the duration of the inspection period with the exception of planned power reductions for routine surveillance testing.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
Readiness for Seasonal Extreme Weather Conditions
a. Inspection Scope
On October 17, 2013, the inspectors completed an inspection of the stations readiness for seasonal extreme weather conditions. The inspectors reviewed the licensees adverse weather procedures for cold weather conditions and evaluated the licensees implementation of these procedures. The inspectors verified that prior to the onset of cold weather the licensee had corrected weather-related equipment deficiencies identified during the previous winter.
The inspectors selected four risk-significant systems that were required to be protected from cold weather:
essential service water
refueling water storage tank
condensate storage tank
The inspectors reviewed the licensees procedures and design information to ensure the systems or components would remain functional when challenged by cold weather. The inspectors verified that operator actions described in the licensees procedures were adequate to maintain readiness of these systems. The inspectors walked down portions of these systems to verify the physical condition of heat trace piping, space heaters, and weatherized enclosures.
These activities constituted one sample of readiness for seasonal adverse weather, as defined in Inspection Procedure 71111.01.
b. Findings
No findings were identified.
1R04 Equipment Alignment
.1 Partial Walkdown
a. Inspection Scope
The inspectors performed partial system walk-downs of the following risk-significant systems:
November 4, 2013, essential switchgear ventilation train B during train A maintenance
December 18, 2013, visual inspection of tendon gallery and accessible external portions of containment
The inspectors reviewed the licensees procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the systems or trains were correctly aligned for the existing plant configuration.
These activities constituted two partial system walk-down samples as defined in Inspection Procedure 71111.04.
b. Findings
No findings were identified.
.2 Complete Walkdown
a. Inspection Scope
On December 27, 2013, the inspectors performed a complete system walk-down inspection of the instrument air system. The inspectors reviewed the licensees procedures and system design information to determine the correct system lineup for the existing plant configuration. The inspectors also reviewed open condition reports, in-process design changes, temporary modifications, and other open items tracked by the licensees operations and engineering departments. The inspectors then visually verified that the system was correctly aligned for the existing plant configuration.
These activities constituted one complete system walk-down sample, as defined in Inspection Procedure 71111.04.
b. Findings
No findings were identified.
1R05 Fire Protection
Quarterly Inspection
a. Inspection Scope
The inspectors evaluated the licensees fire protection program for operational status and material condition. The inspectors focused their inspection on five plant areas important to safety:
October 15, 2013, emergency diesel generator A room, fire area D-1
November 1, 2013, north electrical penetration room 1410, fire area A-18
November 5, 2013, train B battery and switchboard rooms, fire area C-15
November 5, 2013, train A battery and switchboard rooms, fire area C-16
November 12, 2013, letdown heat exchanger and valve compartment rooms, fire area A-1D For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensees fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions.
These activities constituted five quarterly inspection samples, as defined in Inspection Procedure 71111.05.
b. Findings
No findings were identified.
1R06 Flood Protection Measures
a. Inspection Scope
On December 18, 2013, the inspectors completed an inspection of the stations ability to mitigate flooding due to internal causes. After reviewing the licensees flooding analysis, the inspectors chose the following plant areas containing risk-significant structures, systems, or components that were susceptible to flooding:
Rooms 1411, 1412, 3101, 3301, 3302, 3501, and the potential flood paths in turbine building stairwells and communications corridor
The inspectors reviewed plant design features and licensee procedures for coping with internal flooding. The inspectors walked down the selected areas to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether operator actions credited for flood mitigation could be successfully accomplished.
In addition, on December 12, 2013, the inspectors completed an inspection of underground manholes susceptible to flooding. The inspectors selected MH01B, essential service water train B electrical manhole, which contained risk-significant or multiple-train cables whose failure could disable risk-significant equipment.
The inspectors observed the material condition of the cables and splices contained in the vault and looked for evidence of cable degradation due to water intrusion. The inspectors verified that the cables and vault met design requirements.
These activities constitute completion of one flood protection measures sample and one bunker/manhole sample, as defined in Inspection Procedure 71111.06, with the exception that only one manhole was inspected because the licensee only opened one manhole during the course of the year.
b. Findings
No findings were identified.
1R07 Heat Sink Performance
Triennial Inspection
a. Inspection Scope
The inspectors reviewed licensee programs to verify heat exchanger performance and operability for the following heat exchangers:
EEJ01A, residual heat removal heat exchanger A
EKJ06A, diesel generator jacket water heat exchanger A
SGN01C, containment cooler C
The inspectors verified whether testing, inspection, maintenance, and chemistry control programs were adequate to ensure proper heat transfer. The inspectors verified that the periodic testing and monitoring methods, as outlined in commitments to NRC Generic Letter 89-13, utilized proper industry heat exchanger guidance. Additionally, the inspectors verified that the licensees chemistry program ensured that biological fouling was properly controlled between tests. The inspectors reviewed previous maintenance records of the heat exchangers to verify that the licensees heat exchanger inspections adequately addressed structural integrity and cleanliness of their tubes.
These activities constitute completion of three triennial heat sink inspection samples as defined in Inspection Procedure 71111.07-05.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
.1 Review of Licensed Operator Requalification
a. Inspection Scope
On November 25, 2013, the inspectors observed a portion of an annual requalification test for licensed operators. The inspectors assessed the performance of the operators and the evaluators critique of their performance. The inspectors also assessed the modeling and performance of the simulator during requalification activities.
These activities constitute completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
.2 Review of Licensed Operator Performance
a. Inspection Scope
On the following dates, the inspectors observed the performance of on-shift licensed operators in the plants main control room. At the time of the observations, the plant was in a period of heightened activity or risk due to the included reasons. The inspectors observed the operators performance of the following activities:
November 21, 2013, heightened activity for turbine-driven auxiliary feedwater pump post-maintenance testing and loose parts monitor surveillance with under instruction watches while turbine-driven auxiliary feedwater pump was inoperable for maintenance
December 14, 2013, heightened activity for reactor coolant system dilution and annunciator response for a ground fault on a non-safety bus during backshift hours In addition, the inspectors assessed the operators adherence to plant procedures, including Procedure ODP-ZZ-00001, "Operations Department - Code of Conduct," and other operations department policies.
These activities constitute completion of two quarterly licensed operator performance samples, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed two instances of degraded performance or conditions of safety-related structures, systems, or components:
July 15, 2013, reserve auxiliary transformer exceeded the maintenance rule performance criteria, Callaway Action Request 201305640
December 6, 2013, failure of solenoid actuation valve for one condenser steam dump, Callaway Action Request 201308530 The inspectors reviewed the extent of condition of possible common cause structure, system, or component failures and evaluated the adequacy of the licensees corrective actions. The inspectors reviewed the licensees work practices to evaluate whether these may have played a role in the degradation of the structures, systems, or components. The inspectors assessed the licensees characterization of the degradations in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.
These activities constitute completion of two maintenance effectiveness samples, as defined in Inspection Procedure 71111.12.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed three risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk:
October 8, 2013, yellow risk during train A emergency diesel generator and essential service water outage for planned maintenance, Job 09512424
October 22, 2013, yellow risk during train B emergency diesel generator and essential service water outage for planned maintenance, Job 10008706 November 19, 2013, yellow risk during turbine-driven auxiliary feedwater pump outage for planned maintenance, Job 12505697 The inspectors verified that these risk assessments were performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensees risk assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments.
The inspectors also observed portions of two emergent work activities that had the potential to cause an initiating event, to affect the functional capability of mitigating systems, or to impact barrier integrity:
October 3, 2013, train A load shed and emergency load sequencer 48 Volt DC power supply card replacement, Job 10503145
October 23, 2013, train A emergency diesel generator lube oil filter inner support tube installation, Job 13006266 The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on unaffected structures, systems, or components.
These activities constitute completion of five maintenance risk assessment and emergent work control inspection samples, as defined in Inspection Procedure 71111.13.
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments
a. Inspection Scope
The inspectors reviewed five operability determinations that the licensee performed for degraded or nonconforming structures, systems, or components:
August 18, 2013, component cooling water pump bearing modification issues, Callaway Action Requests 201306459, 201306867 and 201306866
October 2, 2013, train A load shed and emergency load sequencer 48 Volt DC power supply failure, Callaway Action Request 201307473
October 7, 2013, component cooling water cross-connect valve leakage, Callaway Action Requests 201307672 and 201303782 October 15, 2013, essential service water through-wall leak on component cooling water heat exchanger, Callaway Action Request 201307915
November 21, 2013, turbine-driven auxiliary feedwater pump outboard bearing dark oil, Callaway Action Request 201308914
The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded structures, systems, and components to be operable, the inspectors verified that the licensees compensatory measures were appropriate to provide reasonable assurance of operability. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability of the degraded structures, systems, and components.
These activities constitute completion of five operability review samples, as defined in Inspection Procedure 71111.15.
b. Findings
No findings were identified.
==1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications (71111.17)
==
.1 Evaluations of Changes, Tests, or Experiments
a. Inspection Scope
The inspectors reviewed seven evaluations to determine whether the changes to the facility or procedures, as described in the final safety analysis report, had been reviewed and documented in accordance with 10 CFR 50.59 requirements. The inspectors verified that when changes, tests, or experiments were made, evaluations were performed in accordance with 10 CFR 50.59 and licensee personnel had appropriately concluded that the change, test, or experiment could be accomplished without obtaining a license amendment. The inspectors also verified that safety issues related to the changes, tests, or experiments were resolved. The team compared the safety evaluations and supporting documents to the guidance and methods provided in Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, as endorsed by NRC Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, to determine the adequacy of the safety evaluations.
The inspectors reviewed 12 samples of changes, tests, and experiments that licensee personnel determined did not require evaluations. The inspectors verified that licensee personnels conclusions were correct and consistent with 10 CFR 50.59.
The inspectors also verified that calculations, analyses, design change documentation, procedures, the final safety analysis report, the technical specifications, and plant drawings used to support the changes were accurate after the changes had been made.
These activities constitute completion of 7 samples of evaluations and 12 samples of changes, tests, and experiments that were screened out by licensee personnel as defined in Inspection Procedure 71111.17-04.
b. Findings
Unresolved Item - Solid State Protection System Modifications
Introduction.
The inspectors identified an unresolved item associated with the implementation of the licensees process to comply with 10 CFR 50.59 for a digital modification of the solid state protection system (SSPS) logic and control boards. This item remains unresolved pending further review by the NRC staff to determine if this issue constitutes a violation of NRC requirements.
Description.
The SSPS logic and control boards provide the coincidence logic to produce actuation signals for operation of the reactor protection system and the engineered safety features actuation systems. Modification Package 10-0053, SSPS Printed Circuit Board Replacement, Version 000.2, evaluated a digital modification to the existing SSPS logic and control boards. This modification replaced existing obsolete printed circuit boards with replacement boards supplied by Westinghouse. The modification replaced universal logic printed circuit boards, safeguards driver printed circuit boards, undervoltage driver printed circuit boards, and semi-automatic tester printed circuit boards. The original circuit boards used fixed logic devices (i.e. transistor-transistor logic) whereas the replacement circuit boards used programmable logic devices (i.e. complex programmable logic devices (CPLD)) to perform the required logic operation for the design function of the SSPS.
The licensee performed a safety evaluation for this modification in accordance with Procedure APA-ZZ-00143, 10 CFR 50.59 Reviews. This procedure stated that its purpose was to describe the process for compliance with the requirements of 10 CFR 50.59 using the guidelines contained in NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, Revision 1. The procedure included the screening questions to be used to determine whether a plant change required an evaluation against the criteria in 10 CFR 50.59(c)(2).
Section 4 of NEI 96-07, Revision 1, states that a 10 CFR 50.59 evaluation is required when a change adversely affects the design function or the method of performing or controlling a design function. The guidance also states that an example that would require an evaluation is a change that introduces a new type of accident or malfunction.
The guidance also states that if a change has both positive and adverse effects, the change would require a 10 CFR 50.59 evaluation and should focus on the adverse effects.
Additionally, NEI 01-01, Guideline on Licensing Digital Upgrades, Revision 1, Section 4.3.2, states that most digital upgrades to redundant safety systems should be conservatively treated as adverse and should require an evaluation. This section also states that some examples of adverse effects that should be evaluated are those that change functionality in a way that increases complexity and introduces different behavior or potential failure modes.
The licensee concluded in their 10 CFR 50.59 evaluation that the replacement of SSPS cards did not meet the criteria in 10 CFR 50.59(c)(2), because the modification did not adversely affect the function of the SSPS as described in the final safety analysis report.
The basis for that conclusion consisted, in part, of the following statements in the 50.59 evaluation:
The CPLD contains no software or programmable code; rather, the CPLD is configured during manufacturing by loading a data file that programs the logic gates in the device. Thus, the CPLD is hardware-based and does not utilize software to perform its function. Hardware is distinguished from software by the degree of testability. The CPLD-based board does not have the characteristics associated with microprocessor based systems such as modifiable code, branches or interrupts, decision-making capability, lockups, and common-mode software failure susceptibility. Thus, failures can be treated as single random hardware failures. The failure of the CPLD will cause the failure of all logic circuits on the board, similar to other failure scenarios for the original-design board. The frequency of hardware failures for the new-design cards, including the CPLD, is therefore comparable to what it was for the original-design boards.
The three safety-related boards that support protective actuation functions have been fully tested as documented in Westinghouse WNA-TR-02644-SCP, Solid State Protection System New Design Circuit Boards Final Logic Test Report, which concludes that the new boards have the same output responses as the original boards such that the criteria for 100 percent testing is satisfied. The steady state operation for every possible logic input verifies the new design SSPS circuit boards operate identical to the original design circuit boards and that no unpredicted or unexpected outputs occur for any possible logic combination input.
The failure modes analyses, qualification processes, and testing of the new circuit boards do not indicate a more than minimal increase in the likelihood of occurrence of a malfunction as a result of this change. The new version circuit boards are still in compliance with the general design criteria. Overall, the replacement of the existing SSPS digital circuit boards with the new design SSPS digital circuit boards does not provide a trend toward increasing the likelihood of malfunction of the structures, systems, or components.
The inspectors reviewed the 10 CFR 50.59 evaluation and the Westinghouse supporting information for the replacement cards and identified various issues of concern associated with the design, testing, and operation of the replacement circuit boards, which could represent adverse effects, with a more than minimal increase in the likelihood of occurrence of a malfunction, to the design function of the SSPS as described in the final safety analysis report. These potential adverse effects would have required an evaluation against the criteria in 10 CFR 50.59(c)(2) as directed by site Procedure APA-ZZ-00143, and the self-imposed NEI guidance (NEI 96-07 and NEI 01-01). Specifically, the inspectors identified that:
While the licensee concluded that the CPLD-based circuit boards contained no software because the manufacturer used a data file or firmware set during initial configuration to program the logic gates in the device board, section 5.3.3.2 of NEI 01-01, defined that type of feature as Base Software. Additionally, NEI 01-01, section 4.3.2, Software Considerations, indicates that digital modifications that involve the use of software applications should be conservatively treated as an adverse effect, due to the potential introduction of new failure modes (software based failures, including common cause failures not previously evaluated, especially when modifications involve redundant safety systems (i.e. reactor protection system or engineered safety features actuation system). The 10 CFR 50.59 evaluation did not contain sufficient information to exclude the data file from the definition of Base Software and the associated design considerations in NEI 01-01.
Second party commercial vendors were involved in the manufacturing of the CPLDs as well as the development of the data file software. The inspectors found that there was not sufficient information in the 10 CFR 50.59 evaluation and supporting vendor information, to determine the level of quality assurance placed into the development of the CPLDs to ensure reliable operation of this logic device. Furthermore, licensee discussions with Westinghouse confirmed that the second party commercial vendors were not qualified to 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
The testing performed by the vendor for the development of the CPLDs only covered the combinations of inputs and outputs (hardware functional testing)required for the design function of the SSPS. However, the 50.59 evaluation and supporting vendor information did not contain sufficient information to determine if the testing performed by the vendor was sufficient enough to cover other possible sequence of device states due to the relative complexity of the CPLDs operation. This would include software-induced states associated with the CPLDs themselves and the embedded data file, which could result in malfunctions of the SSPS.
This issue remains unresolved pending further NRC review of additional information provided by Westinghouse to address the concerns described above, in order to determine the adequacy of the licensees 50.59 evaluation and whether or not the issue represents a violation of 10 CFR 50.59, Changes, Tests, and Experiments. The licensee entered this issue in the corrective action program as Callaway Action Request 201306081 to address operability of the SSPS and evaluate the need for a license amendment. The licensee completed a prompt operability determination. The inspectors reviewed the operability determination and did not identify any issues regarding the operability of the SSPS cards.
This issue is being tracked as URI 05000483/2013005-01, Solid State Protection System Modifications.
.2 Permanent Plant Modifications
a. Inspection Scope
The inspectors verified that calculations, analyses, design change documentation, procedures, the final safety analysis report, the technical specifications, and plant drawings used to support the modifications were accurate after the systems, structures, and components had been modified. The inspectors verified that modifications were consistent with the plants licensing and design bases. The inspectors confirmed that revised calculations and analyses demonstrated that the modifications did not adversely impact plant safety. Additionally, inspectors interviewed design and system engineers to assess the adequacy of the modifications.
These activities constitute completion of eight samples of permanent plant modifications as defined in Inspection Procedure 71111.17-04.
.2.1 Install Safety Related Tie-in for Hardened Condensate Storage Tank
The inspectors reviewed Modification Package 12-0002, which implemented a modification to the auxiliary feedwater system pressure boundary for future addition of a condensate storage tank that would withstand both seismic and tornado missile events.
The modification added safety related ASME Class 3 piping, tees, and valves to the existing auxiliary feedwater system to establish safety related tie-ins for a future hardened condensate storage tank. The inspectors reviewed the in-process nondestructive examination results and the post-modification nondestructive examination results to determine that the ASME pressure boundary was maintained.
Additionally, the inspectors reviewed changes to supporting drawings, procedures, calculations, and design basis documents to verify that the associated programs and processes were updated.
.2.2 Replace Tobar Pressurizer Pressure Transmitters with Rosemount
The inspectors reviewed Modification Package 08-0054, implemented to replace obsolete pressure transmitters for the pressurizer instrumentation loops. The modification replaced obsolete Tobar pressure transmitters with Rosemount model 1154 pressure transmitters. Replacement of the new pressure transmitters had the effect of reducing the allowable tolerance of the transmitters and their corresponding instrument loops. The inspectors reviewed the transmitter and instrumentation loop tolerance calculations to ensure that the reduced tolerances were addressed in the calculations.
Additionally, the inspectors reviewed the calibration procedures and the results of the post-modification calibration to ensure that the affected acceptance criteria were updated to include the new allowable tolerance.
.2.3 Ultimate Heat Sink Temperature Issue Solution
The inspectors reviewed Modification Package 11-0004, which implemented a modification to the essential service water system and associated cooling towers. The modification raised the minimum ultimate heat sink level and level setpoints, added new circuitry for cooling tower fan speed indication in the control room, added new control logic to interface the essential service water temperature loops to the cooling tower fan speed and bypass valve control, and added new control room hand-switches to remotely close the cooling tower bypass valves. The inspectors reviewed changes to supporting drawings, procedures, calculations, and design basis documents to verify that the associated programs and processes were updated. Additionally, the inspectors reviewed operating procedures and completed walkdowns of the affected safety related equipment to determine if operator actions would maintain the equipment within their design basis.
.2.4 Installation of Emergency Core Cooling and Containment Spray System Vents
The inspectors reviewed Modification Package 08-0016. The purpose of the modification was to design a vent assembly that can be used in any location of the emergency core cooling system or containment spray system designed to ASME Section III Subsection NC (Class 2) to vent trapped gasses identified through ultrasonic test examination or some other inspection means. The subsystems that compromise the emergency core cooling system are as follows: chemical and volume control, residual heat removal, high pressure coolant injection, and accumulator safety injection. The modification also included the design of a bypass line and a clamp assembly to alleviate excessive vibration on a newly installed vent assembly. The inspectors additionally reviewed design and piping and instrumentation drawings.
.2.5 Component Cooling Water Heat Exchanger Divider Plate Repair to Prevent Bypass Flow
The inspectors reviewed Modification Package 09-0056. The purpose of the modification was to repair the component cooling water heat exchangers to restore their design margin and to eliminate bypass flow between the front cover plate and the partition plate. Approximately a one-quarter inch gap existed between the cover plate and the partition plate, which allowed cooling water to traverse the gap and bypass the heat exchanger tubes. The inspectors also reviewed thermal performance tests of the heat exchangers, seismic and structural calculations, and field change notices associated with the modification.
.2.6 Replacement of Emergency Diesel Generator Jacket Water and Lube Oil Heat
Exchangers The inspectors reviewed Modification Package 09-0076. The purpose of the modification was to replace the emergency diesel generator jacket water and lube oil heat exchangers with new units that are less susceptible to corrosion and/or erosion. A new vent valve was also installed in the jacket water piping to replace a pipe plug. This was to address a safety concern of potentially venting hot (165°F) water. Additionally, an upgrade of an existing carbon steel vent valve on emergency diesel generator B jacket water heat exchanger essential service water out vent and vent line piping to stainless steel was performed. The inspectors also reviewed the heat exchanger data sheets for the new units, updated calculations of heat exchanger thermal performance, preventive maintenance procedures, and one field change notice associated with the modification.
.2.7 Installation of 30-Inch Check Valves to Minimize Essential Service Water System Water
Hammer The inspectors reviewed Modification Package 10-0003. The purpose of the modification was to minimize the occurrence of water hammer that occurred in the essential service water and service water systems during engineered safety features actuation system testing and loss of offsite power events. This was accomplished by installing a 30-inch check valve in each service water supply line that ties into the essential service water supply headers. Two 2-inch drain valves in the service water supply lines were also installed. The inspectors also reviewed calculations for piping supports and anchors, hazards reviews, and field change notices associated with the modification.
.2.8 Evaluation of New Design for Emergency Diesel Generator Intercooler Heat Exchangers
The inspectors reviewed Modification Package 10-0008. The purpose of the modification was to replace the emergency diesel generator intercooler heat exchangers with new units that are less susceptible to corrosion and/or erosion. The modification replaced the tubing materials, channel heads, and tube sheets of the intercooler heat exchangers. The inspectors also reviewed the heat exchanger data sheets for the new units, updated calculations of heat exchanger thermal performance, preventive maintenance procedures, the change assessment matrix, and field change notices associated with the modification.
b. Findings
Ultimate Heat Sink Cooling Tower - Incorrect Action Criteria in Control Room Evacuation Procedure
Introduction.
The inspectors identified a Green non-cited violation of Technical Specification 5.4.1(d), Procedures, for the licensees failure to maintain procedures as required by the fire protection program. Specifically, the operators were directed to use incorrect process indication and action criteria to diagnose and prevent icing conditions in the cooling tower after the control room was evacuated.
Description.
The inspectors reviewed Procedure OTO-ZZ-00001, Control Room Inaccessibility, Revision 38, to understand the manual operator actions necessary to operate the ultimate heat sink cooling towers after the evacuation of the control room. In D, Control Room Supervisor Actions with Fire, the procedure directed the control room supervisor to prepare, for manual operation, the essential service water system train and associated cooling tower credited for fire safe shutdown. After steps to remove power to remotely control essential service water cooling tower bypass valve EFHV0066, and cooling tower fan control circuit, and running a train of cooling tower fans in low speed, step D31 and D32 stated:
D31. MONITOR UHS Cooling Twr B & D Return Temp Ind
EFTI0084 D32. PERFORM One Of The Following:
If Indicated Temperature Is Greater Than 32°F Manually CLOSE EFHV0066, ESW UHS Cool-TWR Trn B Byp HV OR
If Indicated Temperature Is Less Than 32°F Manually OPEN EFHV0066, ESW UHS Cool-TWR Trn B Byp HV
These steps directed the operator to monitor the temperature of the essential service water returning to the cooling tower to prevent icing in the cooling tower. Icing in the cooling tower would degrade the heat transfer capability of the tower and could damage critical internal tower components. If the temperature of the essential service water was greater than 32°F, the essential service water would be cooled by evaporative cooling through the cooling towers. If the essential service water temperature was lower than 32°F, the essential service water would bypass the cooling tower and be injected directly into the ultimate heat sink pond.
The inspectors determined that these procedure steps would not direct the operators to correctly diagnose, prevent, or mitigate icing in the cooling tower after a control room evacuation. Specifically, essential service water that is returning to the cooling tower, as it is carrying heat from the plant, would never fall below the freezing point of water, 32°F.
Additionally, operating the bypass valve based on lower essential service water return indicator EFTI0084 does not consider the effects of operating the cooling tower during extreme cold weather. The vendor manual for the cooling tower states:
Whenever wet-bulb temperature is below freezing, regardless of the dry-bulb temperature, ice may form on the relatively dry parts of the cooling tower where fine drops of water splash out into the entering air stream. Ice may form on the structural framing and outer filling but will not occur in the flooded portions of the tower.
Request for Resolution 200703556 further states:
Ice formation on the tower fill will reduce heat transfer capacity and could damage critical internal tower components.
These conclusions and the cooling tower performance curves were correctly translated into the operating procedure for the essential service water system, Procedure OTN-EF-00001, Essential Service Water System. This procedure does not permit the operators to manually close the essential service water cooling tower bypass valve, which sends essential service water flow through the cooling tower, when wet-bulb ambient temperature is below 32°F. If air temperature is less than or equal to 32°F, the essential service water return temperature must be greater than or equal to 61°F.
Analysis.
The inspectors determined that the failure of Procedure OTO-ZZ-00001, to contain the appropriate criteria for operator action to control the ultimate heat sink cooling tower bypass valve after a control room evacuation was a performance deficiency. The inspectors determined that the performance deficiency was associated with the Mitigating Systems Cornerstone and was more than minor, and therefore a finding, because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the action criteria would not direct the operators to correctly diagnose, prevent, or mitigate icing in the ultimate heat sink cooling tower after a control room evacuation. Using NRC Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding had very low safety significance (Green) because this finding would not have prevented the reactor from reaching and maintaining safe shutdown. This finding was not assigned a cross-cutting aspect because the most significant contributor was not reflective of current licensee performance.
Enforcement.
The inspectors identified a Green non-cited violation of Technical Specification 5.4.1(d), Procedures, which states, in part, written procedures shall be established, implemented, and maintained covering fire protection program implementation. Contrary to this requirement, the licensee failed to maintain procedures covering fire protection program requirements. Specifically, prior to October 10, 2013, Procedure OTO-ZZ-00001, Control Room Inaccessibility, Revision 38, steps D31 and D32, directed operators to use incorrect process indication and action criteria to diagnose and prevent icing conditions in the cooling tower after the control room was evacuated. As a result, the licensee revised the action criteria to direct the operators to take action to prevent icing in the cooling tower. This violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy. This violation was entered into the corrective action program as Callaway Action Request 201307709: NCV 05000483/2013005-02, Ultimate Heat Sink Cooling Tower - Incorrect Action Criteria in Control Room Evacuation Procedure.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed five post-maintenance testing activities that affected risk-significant structures, systems, or components:
October 9, 2013, emergency diesel generator A post-maintenance test following air start solenoid replacement, Job 13504846
October 22, 2013, ultimate heat sink fan B post-maintenance test following motor preventative maintenance, Job 12504141
October 23, 2013, essential service water B return isolation valve (EFHV0038)retest following motor operator preventative maintenance, Job 13006258 October 30, 2013, containment cooler C containment isolation valves post-maintenance test following breaker replacement, Job 10513514
November 20, 2013, turbine-driven auxiliary feedwater pump post-maintenance test following regularly scheduled outage, Job 12506030 The inspectors reviewed licensing-and design-basis documents for the structures, systems, or components and the maintenance and post-maintenance test procedures.
The inspectors observed the performance of the post-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected structures, systems, or components.
These activities constitute completion of five post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed three risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the structures, systems, or components were capable of performing their safety functions:
In-service tests:
October 2, 2013, centrifugal charging pump A surveillance test, Job 13508617
Other surveillance tests:
October 21, 2013, emergency exhaust system train B operability test, Job 13511347
November 12, 2013, containment spray system train B surveillance test, Job 13510195 The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected structures, systems, or components following testing.
These activities constitute completion of three surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.
b. Findings
Introduction.
The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to correct a condition adverse to quality. Specifically, the licensees calculations to determine acceptable oil leak rates of safety related pumps in response to Callaway Action Requests 201102434 and 201207677 failed to ensure that the pumps mission times could be met. As a result, non-conservative limits were added to Procedure ODP-ZZ-0016E, Appendix 1, Operations Technician General Inspection Guide.
Description.
On October 2, 2013, while monitoring the centrifugal charging pump train A in-service test, NRC inspectors reviewed Procedure ODP-ZZ-0016E, Appendix 1, guidance for operations technicians and determined that each centrifugal charging pumps allowed bearing oil leak rate was 2.5 gallons per hour. Inspectors found that this rate was determined in response to Callaway Action Request 201207677, written to address concerns with the allowed lubricating oil volumes required for operability for those particular pumps.
The inspectors reviewed the oil leakage evaluation enclosed in Callaway Action Request 201207677 and found that engineers bounded the time limit to only the design basis mission time and did not consider the frequency of operator rounds when checking leaks and oil levels. As such, the pump could have a leak less than the maximum allowed and oil greater than the minimum level; however within approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, less than one operations shift, the pump would not be able to meet its mission time and would therefore be inoperable. The inspectors then questioned the accuracy of the operability limits for the other safety related equipment. Those were determined in an attachment to Callaway Action Request 201102434, as a corrective action for NRC NCV 05000483/2011007-04, which documented the failure to provide adequate oil leakage specifications for safety related equipment, primarily due to inadequate mission times.
Safety related equipment inspected for oil leakage by operations technicians during normal rounds at Callaway Plant include the safety injection pumps, centrifugal charging pumps, component cooling water pumps, and auxiliary feedwater pumps. The inspectors determined that none of the calculations took into account the time period between operator rounds, during which time the oil level would be lowering due to the leak. The resulting quantities were then incorporated into ODP-ZZ-0016E guidance for operations technicians rounds without a critical review by engineering or operations management. The resultant worst case scenario could have caused the auxiliary feedwater pumps bearing oil to drain below the necessary levels to provide lubrication within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 40 minutes, well short of the time an operations technician would be expected to observe the pump during their next normal rounds. The licensee addressed this issue in Callaway Action Request 201308127 and planned to revise the allowable leakage values. NRC inspectors determined there was never a time within the last several years when any of these pumps were inoperable due to oil leakage.
Analysis.
The inspectors determined that the failure to correct non-conservative oil leakage specifications used in operator logs was a performance deficiency. Specifically, the inspectors determined that the licensee had revised the leakage limits to correctly account for the mission time of each pump, but failed to account for the time between observations on operator rounds. The inspectors determined that the performance deficiency was associated with the Mitigating Systems Cornerstone and was more than minor, and therefore a finding, because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to adequately evaluate lubricating oil leakage on safety related equipment could have impacted the availability of mitigating systems if left uncorrected. The finding was assessed using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and was determined to be of very low safety significance since the inadequate specifications did not result in a loss of operability for the affected systems. This finding has a cross-cutting aspect in the decision-making component of the human performance cross-cutting area because the licensee failed to utilize a systematic process to ensure safety is maintained. Specifically, the authority and roles for decisions affecting nuclear safety in the corrective actions for Callaway Action Requests 201102434 and 201207677 were not formally defined resulting in inadequate engineering conclusions being incorporated into operations procedures H.1(a).
Enforcement.
Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that conditions adverse to quality such as deficiencies are promptly identified and corrected. Contrary to this, from May 23, 2013 to December 12, 2013, the licensee failed to correct non-conservatisms in the calculational bases for operator logs maximum allowable oil leakage limits of the emergency core cooling pumps, auxiliary feedwater pumps, and component cooling water pumps in their corrective actions for Callaway Action Request 201102434. This violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy. This violation was entered into the corrective action program as Callaway Action Request 201308127: NCV 05000483/2013005-03, Failure to Correct Non-conservative Safety Related Equipment Oil Leakage Criteria.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
Training Evolution Observation
a. Inspection Scope
On November 26, 2013, the inspectors observed simulator-based licensed operator requalification training that included implementation of the licensees emergency plan.
The inspectors verified that the licensees emergency classifications, off-site notifications, and protective action recommendations were appropriate and timely. The inspectors verified that any emergency preparedness weaknesses were appropriately identified by the evaluators and entered into the corrective action program for resolution.
These activities constitute completion of one training observation sample, as defined in Inspection Procedure 71114.06-05.
b. Findings
No findings were identified.
RADIATION SAFETY
Cornerstones: Public Radiation Safety and Occupational Radiation Safety
2RS5 Radiation Monitoring Instrumentation
a. Inspection Scope
The inspectors verified the accuracy and operability of the radiation monitoring equipment used by the licensee
- (1) to monitor areas, materials, and workers to ensure a radiologically safe work environment, and
- (2) to detect and quantify radioactive process streams and effluent releases. The inspectors interviewed licensee personnel, walked down various portions of the plant, and reviewed licensee performance in the following areas:
Selected plant configurations and alignments of process, post-accident, and effluent monitors with descriptions in the final safety analysis report, and the offsite dose calculation manual
Select instrumentation, including effluent monitoring instrument, portable survey instruments, area radiation monitors, continuous air monitors, personnel contamination monitors, portal monitors, and small article monitors to examine their configurations and source checks
Calibration and testing of process and effluent monitors, laboratory instrumentation, whole body counters, post-accident monitoring instrumentation, portal monitors, personnel contamination monitors, small article monitors, portable survey instruments, area radiation monitors, electronic dosimetry, air samplers, continuous air monitors
Audits, self-assessments, and corrective action documents related to radiation monitoring instrumentation since the last inspection
These activities constitute completion of one required sample as defined in Inspection Procedure 71124.05-05.
b. Findings
No findings were identified.
2RS6 Radioactive Gaseous and Liquid Effluent Treatment
a. Inspection Scope
The inspectors verified that the licensee maintained gaseous and liquid effluent processing systems and properly mitigated, monitored, and evaluated radiological discharges with respect to public exposure. The inspectors verified that abnormal radioactive gaseous or liquid discharges and conditions, when effluent radiation monitors are out-of-service, were controlled in accordance with the applicable regulatory requirements and licensee procedures. The inspectors verified that the licensees quality control program ensured radioactive effluent sampling and analysis adequately quantified and evaluated discharges of radioactive materials. The inspectors verified the adequacy of public dose projections resulting from radioactive effluent discharges. The inspectors interviewed licensee personnel and reviewed or observed the following items:
Radiological effluent release reports since the previous inspection and reports related to the effluent program issued since the previous inspection
Effluent program implementing procedures, including sampling, monitor setpoint determinations and dose calculations
Equipment configuration and flow paths of selected gaseous and liquid discharge system components, filtered ventilation system material condition, and significant changes to their effluent release points, if any, and associated 10 CFR 50.59 reviews
Selected portions of the routine processing and discharge of radioactive gaseous and liquid effluents (including sample collection and analysis)
Controls used to ensure representative sampling and appropriate compensatory sampling
Results of the inter-laboratory comparison program
Effluent stack flow rates
Surveillance test results of technical specification-required ventilation effluent discharge systems since the previous inspection
Significant changes in reported dose values
A selection of radioactive liquid and gaseous waste discharge permits
Part 61 analyses and methods used to determine which isotopes are included in the source term
Offsite dose calculation manual changes Meteorological dispersion and deposition factors
Latest land use census
Records of abnormal gaseous or liquid tank discharges
Ground water monitoring results
Changes to the licensees written program for identifying and controlling contaminated spills/leaks to groundwater
Identified leakage or spill events and entries made into 10 CFR 50.75(g) records, if any, and associated evaluations of the extent of the contamination and the radiological source term
Offsite notifications and reports of events associated with spills, leaks, or ground water monitoring results
Audits, self-assessments, reports, and corrective action documents related to radioactive gaseous and liquid effluent treatment since the last inspection These activities constitute completion of one required sample, as defined in Inspection Procedure 71124.06-05.
b. Findings
No findings were identified.
2RS7 Radiological Environmental Monitoring Program
a. Inspection Scope
This area was inspected to:
- (1) ensure that the radiological environmental monitoring program verifies the impact of radioactive effluent releases to the environment and sufficiently validates the integrity of the radioactive gaseous and liquid effluent release program;
- (2) verify that the radiological environmental monitoring program is implemented consistent with the licensees technical specifications and/or offsite dose calculation manual, and to validate that the radioactive effluent release program meets the design objective contained in Appendix I to 10 CFR Part 50; and
- (3) ensure that the radiological environmental monitoring program monitors non-effluent exposure pathways, is based on sound principles and assumptions, and validates that doses to members of the public are within the dose limits of 10 CFR Part 20 and 40 CFR Part 190, as applicable. The inspectors reviewed and/or observed the following items:
Annual environmental monitoring reports and offsite dose calculation manual
Selected air sampling and dosimeter monitoring stations Collection and preparation of environmental samples
Operability, calibration, and maintenance of meteorological instruments
Selected events documented in the annual environmental monitoring report which involved a missed sample, inoperable sampler, lost thermoluminescence dosimeter, or anomalous measurement
Selected structures, systems, or components that may contain licensed material and has a credible mechanism for licensed material to reach ground water
Records required by 10 CFR 50.75(g)
Significant changes made by the licensee to the offsite dose calculation manual as the result of changes to the land census or sampler station modifications since the last inspection
Calibration and maintenance records for selected air samplers, composite water samplers, and environmental sample radiation measurement instrumentation
Inter-laboratory comparison program results
Audits, self-assessments, reports, and corrective action documents related to the radiological environmental monitoring program since the last inspection These activities constitute completion of one required sample as defined in Inspection Procedure 71124.07-05.
b. Findings
No findings were identified.
2RS8 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage,
and Transportation (71124.08)
a. Inspection Scope
The inspectors verified the effectiveness of the licensees programs for processing, handling, storage, and transportation of radioactive material. The inspectors interviewed licensee personnel and reviewed the following items:
The solid radioactive waste system description, process control program, and the scope of the licensees audit program
Control of radioactive waste storage areas including container labeling/marking and monitoring containers for deformation or signs of waste decomposition Changes to the liquid and solid waste processing system configuration including a review of waste processing equipment that is not operational or abandoned in place
Radio-chemical sample analysis results for radioactive waste streams and use of scaling factors and calculations to account for difficult-to-measure radionuclides
Processes for waste classification including use of scaling factors and 10 CFR Part 61 analysis
Shipment packaging, surveying, labeling, marking, placarding, vehicle checking, driver instructing, and preparation of the disposal manifest
Audits, self-assessments, reports, and corrective action reports radioactive solid waste processing, and radioactive material handling, storage, and transportation performed since the last inspection These activities constitute completion of one required sample as defined in Inspection Procedure 71124.08-05.
b. Findings
Introduction.
The inspectors reviewed a self-revealing, Green, non-cited violation of 10 CFR 71.5(a) for the failure to describe a radioactive material shipment correctly and the failure to include all required hazard communication information in a radioactive shipment document.
Description.
On August 10, 2011, the licensee shipped a sea/land container of dry active waste to a waste processor. The shipment included empty shielded drums within B25 boxes loaded into the sea/land container. The container was shipped and labeled as Low Specific Activity (LSA) -II. Low Specific Activity (LSA-II) is defined by 49 CFR 173.403 as radioactive material in which the activity is distributed throughout the package contents. The licensee had originally determined the shipment contained 709 millicuries and 44,574 pounds of dry active waste.
On June 20, 2012, the waste processor contacted the licensee regarding an abnormality with the shipment. While preparing the shipment contents for processing, the processor discovered one of the drums contained mechanical filter pieces. The processor characterized the filter pieces and determined the pieces would be greater than Class C waste, if disposed. Even with waste averaging, the entire shipment would have been Class C waste, and one of the processors acceptance criteria was that wastes received from a generator could not be greater than Class A. Therefore, the waste processor returned the shielded drum with the mechanical filter pieces to the licensee.
The licensee subsequently calculated the activity of the mechanical filter pieces to be approximately 5.13 curies (including 1.58 curies of cobalt-60). In addition, the 5.13 curies of activity was concentrated in approximately 15 pounds of the mechanical filter pieces, and contained within the shielded drum. Thus, the activity was not distributed throughout the package contents and the shipment could not be described as low specific activity material, as defined by 49 CFR 173.403. Because licensee representatives were not aware of the filter pieces being included in the shipment, they also did not describe all required hazard communication information in the shipping documents as specified by 49 CFR 172.203(d), i.e. the name of each radionuclide, a description of the physical form of the material, and the activity contained.
The licensee placed the event into the corrective action program as Callaway Action Request System 201204454. The licensee investigated the occurrence to determine how the mechanical filter pieces were included in the shipment and found inaccurate radioactive material storage documentation caused the licensee to believe the shielded drum was empty when in fact it was not. The initial characterization and labeling of the shielded drum occurred in 1993, before the licensee instituted a program to more accurately track and maintain radioactive material storage containers onsite. Licensee representatives had an additional opportunity to discover the situation; however, they did not confirm the contents of the drum, assumed it was empty, and shipped the drum as part of the LSA-II shipment. Licensee representatives stated the procedural guidance for radioactive shipments did not instruct them to confirm the contents of the containers before shipping. Licensee representatives stated the procedure would be revised as a corrective action.
The inspectors reviewed the processors license and confirmed the radioactivity possession limits were not exceeded. The radiation from the mechanical filter pieces was shielded within the drum and the drum readings were not different from the empty shielded drums also shipped for waste processing. The dose rates for the shipment to the processor did not exceed the regulatory limits of 49 CFR 173.441.
Analysis.
The failure to describe a radioactive material shipment correctly and to include all required hazard communication information in accordance with hazardous material transportation regulations was a performance deficiency. The finding was determined to be more than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process (transportation program) and adversely affected the associated cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation, in that it resulted in the incorrect identification and quantification of radioactive material transported in the public domain.
Using IMC 0609 Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, the inspectors determined this to be a finding of very low safety significance (Green) because the finding did not involve the radioactive effluent release program or the radiological environmental monitoring program, but the finding did involve the transportation of radioactive material. The finding was not
- (1) in excess of radiation limits,
- (2) a breach of package during transit,
- (3) a certificate of compliance issue,
- (4) a low-level burial ground noncompliance, or
- (5) a failure to make notifications or provide emergency information.
The finding has a human performance cross-cutting aspect associated with licensee resources, in that the licensee had insufficient instruction in its shipping guidance to ensure package contents were verified H.2(c).
Enforcement.
Title 10 CFR 71.5(a) requires, in part, that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in Title 49 CFR Parts 107, 171 through 180, and 390 through 397, appropriate to the mode of transport. Title 49 CFR 173.22(a) states, in part, the person shall describe the hazardous material in accordance with 49 CFR 172 and 49 CFR 173.
Contrary to the above, on August 10, 2011, the licensee failed to describe the hazardous material in accordance with 49 CFR 172 and 49 CFR 173. Specifically, the licensee described the contents of Shipment 2011-20 as low specific activity, although the activity was not distributed throughout the package contents as defined by 49 CFR 173.403.
Additionally, the licensees description of the shipment failed to accurately include
- (1) the name of each radionuclide in the Class 7 (radioactive) material that is listed in 49 CFR 173.435,
- (2) a description of the physical and chemical form of the material, and
- (3) the activity contained in each package of the shipment, as required by 49 CFR 172.203(d).
Because this finding was of very low safety significance (Green) and was documented in the licensees corrective action program as Callaway Action Request System 201204454, it is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000483/2013004-04, Failure to describe a radioactive material shipment correctly and failure to include all required hazard communication information in a radioactive shipment document.
OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security
4OA1 Performance Indicator Verification
.1 Mitigating Systems Performance Index:
High Pressure Injection Systems (MS07)
a. Inspection Scope
The inspectors reviewed the licensees mitigating system performance index data for the period of fourth quarter 2012 through third quarter 2013 to verify the accuracy and completeness of the reported data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data.
These activities constituted verification of the mitigating system performance index for high pressure injection systems for Callaway Plant, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
.2 Mitigating Systems Performance Index:
Residual Heat Removal Systems (MS09)
a. Inspection Scope
The inspectors reviewed the licensees mitigating system performance index data for the period of fourth quarter 2012 through third quarter 2013 to verify the accuracy and completeness of the reported data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data.
These activities constituted verification of the mitigating system performance index for residual heat removal systems for Callaway Plant, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution
.1 Routine Review
a. Inspection Scope
Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensees corrective action program and periodically attended the licensees condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified. The inspectors also reviewed the licensees problem identification and resolution activities during the performance of the other inspection activities documented in this report.
b. Findings
No findings were identified.
.2 Semiannual Trend Review
a. Inspection Scope
The inspectors reviewed the licensees corrective action program, performance indicators, system health reports, quality assurance audit/surveillance reports, self-assessment reports, and other documentation to identify trends that might indicate the existence of a more significant safety issue. The inspectors verified that the licensee was taking corrective actions to address identified adverse trends. The inspectors nominally considered the 6-month period of June 2013 through November 2013 although some examples expanded beyond those dates where the scope of the trend warranted.
These activities constitute completion of one semiannual trend review sample, as defined in Inspection Procedure 71152.
b. Findings
No findings were identified.
.3 Annual Follow-up of Selected Issues
a. Inspection Scope
The inspectors selected one issue for an in-depth follow-up:
On October 10, 2013, the inspectors reviewed Callaway Action Request 201307816, for failure to enter Technical Specification 3.8.3 when safety-related equipment was rendered inoperable. Specifically, the licensee did not identify and address the technical specification applicability for the air start system becoming inoperable when the associated emergency diesel generator was already inoperable.
The inspectors assessed the licensees problem identification threshold and cause analyses. The inspectors verified that the licensee appropriately prioritized the corrective actions and that these actions were adequate to correct the condition.
These activities constitute completion of one annual follow-up sample, as defined in Inspection Procedure 71152.
b. Findings
No findings were identified.
4OA5 Other Activities
(Closed) Unresolved Item (URI) 05000483/2012005-03: Determine Licensing Basis and Capability of One Vital Air Conditioning Unit to Cool Both Trains of Class 1E Electrical Equipment The inspectors identified an unresolved item involving the licensing basis and cooling capacity of the safety-related air conditioning units and the ability to cool both trains of safety-related switchgear, batteries, battery chargers, and inverters with a single train of cooling as documented in Inspection Report 05000483/2012005. Specifically, the licensee had relied on compensatory measures to open the doors between trains and then allowed 7 days to restore the cooling unit.
The unresolved item was opened with four concerns:
- (1) Appropriateness and cooling capability of the compensatory measures
- (2) Use of a Final Safety Analysis Report administrative technical specification to address operability
- (3) Completeness of the calculation used as a technical basis
- (4) How the licensee addressed single failure criteria when creating the Final Safety Analysis Report administrative technical specification The licensee contracted a vendor to perform an analysis of the heat loads and cooling requirements for the class 1E electrical equipment, including modeling the temperature profile in each affected room for the 30-day mission time of the associated equipment.
The analysis took external and internal heat loads into account. It then analyzed the cooling capability of a single air conditioning unit with the room dimensions and airflow patterns. The results show that with only one air conditioning unit running and all the compensatory measures in place, both trains of class 1E electrical equipment will remain able to perform their credited functions. This review identified a number of heat terms that were not properly included in the previous evaluation, but the inclusion of these terms did not significantly alter the results. This new analysis addresses concerns 1 and 3.
In response to the inspectors concerns, the licensee changed their method of addressing operability if one train of cooling was lost. If either train of the cooling system becomes nonfunctional, they will immediately declare the associated class 1E electrical equipment inoperable and enter any required actions governed by technical specifications. They will then enter the issue into their corrective action program and perform an operability evaluation. Using the previously mentioned analysis, the licensee will perform an evaluation to verify that the existing conditions match those in the analysis and can support declaring the system operable but degraded for up to 7 days with the compensatory actions in place. After 7 days, the licensee will declare all affected equipment inoperable and take appropriate technical specification actions.
Operable but degraded is a condition that is described in NRC Inspection Manual Part 9900: Technical Guidance, Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety. Because the licensee now appropriately addresses operability through their corrective action process and implements compensatory actions, the single failure criteria is no longer a concern, as compensatory measures do not have to be single failure proof. The implementation of this process adequately addresses concerns 2 and 4.
The inspectors concluded that the licensee satisfactorily responded to the concerns identified in the unresolved item. Therefore, the unresolved item is closed. No findings were identified.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On August 22, 2013, the inspectors presented the radiation safety inspection results to Mr. C. Reasoner, Vice President, Engineering, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed. On October 2, 2013, the inspectors conducted a teleconference with Mr. B. Cox, Senior Director, Nuclear Operations, and other members of the licensee staff to discuss the re-characterization of the violation in Section 2RS8 of this report.
On October 3, 2013, the inspectors presented the preliminary heat sink performance inspection results to Mr. C. Reasoner, Vice President, Engineering and other members of the licensee staff.
On December 9, 2013, the inspectors telephonically presented the final inspection results to Mr. S. Maglio, Manager, Regulatory Affairs, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
On October 8, 2013, the inspectors presented the preliminary inspection results of the permanent plant modification and 10 CFR 50.59 inspection to Mr. D. Neterer, Senior Director, Engineering, and other members of the licensee staff. On October 29, 2013, the inspectors presented the final inspection results, by telephone, to Mr. B. Huhmann, Director, Engineering Design. The licensee acknowledged the results, as presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
On January 8, 2014, the resident inspectors presented the inspection results to Mr. C. Reasoner, Vice President, Engineering, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation.
Title 10 of the Code of Federal Regulations (10 CFR) 20.1501(a) requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present. Title 10 of the Code of Federal Regulations (10 CFR) 20.1301(a) requires each licensee to conduct operations so that:
- (1) the total effective dose equivalent to individual members of the public from the licensed operation does not exceed 0.1 rem in a year, and
- (2) the dose in any unrestricted area from external sources does not exceed 0.002 rem in any one hour.
Contrary to 10 CFR 20.1501(a), the licensee did not make surveys necessary to comply with 10 CFR 20.1301(a). Specifically, on June 20, 2011, the licensee shipped a Department of Transportation Type 7A Yellow III radioactive shipment by common carrier without evaluating the potential dose to the driver of the vehicle or the dose, in any one hour, in an unrestricted area. The licensee identified the error during a self-assessment, entered the issue into the corrective action program as Callaway Action Request 201305808, and evaluated the radiation levels to the driver and the dose, in any one hour, in the unrestricted area. The dose to the driver was determined to be 0.027 rem, which did not exceed the public dose limit of 10 CFR 20.1301(a). However, the dose, in any one hour, in an unrestricted area was determined by the licensee to be 0.003 rem, which did exceed the limit of 10 CFR 20.1301(a)(2). Using the Occupational Radiation Safety significance determination process, the inspectors determined the violation had very low safety significance because:
- (1) it was not an ALARA planning problem,
- (2) it was not an overexposure,
- (3) it did not represent a substantial potential for an overexposure, and
- (4) it did not compromise the licensees ability to assess dose.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- S. Banker, Director, Training
- D. Bond, Consulting Engineer, Nuclear Oversight
- P. Bott, System Engineer, Systems Engineering
- M. Covey, Assistant Operations Manager, Operations
- B. Cox, Senior Director, Nuclear Operations
- F. Diya, Vice President, Nuclear Operations
- L. Eitel, Supervising Engineer, Engineering Systems
- T. Elwood, Supervising Engineer, Regulatory Affairs/Licensing
- L. Franks, Engineer, Systems Engineering
- K. Gilliam, Supervisor, Radiation Protection
- C. Graham, Consulting Health Physicist
- L. Graessle, Senior Director, Operations Support
- M. Hoehn, Supervisor, Engineering Programs
- J. Houston, Senior Health Physicist, Radiation Protection
- B. Huhmann, Acting Director, Engineering Design
- G. Hurla, Supervisor, Radiation Protection
- J. Hutchinson, Supervisor, Engineering Design
- A. King, Senior Health Physicist, Radiation Protection
- G. Kremer, Director, Engineering Programs
- J. Little, Supervisor, Safety Analysis/Reactor Engineering
- S. Maglio, Manager, Regulatory Affairs
- J. Mayer, Supervisor, Radwaste Operations
- J. McLaughlin IV, NESM System Engineer, Systems Engineering
- M. McLachlan, Director, Engineering Systems
- S. Merciel, Engineer, Engineering - Plant Life Extension
- V. Miller, Supervisor, Radiation Protection
- D. Neterer, Senior Director, Engineering
- J. Patterson, Director, Construction
- S. Petzel, Engineer, Regulatory Affairs/Licensing
- L. Ptasnik, Program Manager, Regulatory Affairs/Licensing
- C. Reasoner, Vice President, Engineering
- D. Riegel, Engineer, Performance Engineering
- C. Smith, Manager, Radiation Protection
- V. Thomas, Supervising Engineer, Performance Engineering
- D. Thompson, Senior Health Physicist, Radiation Protection
- K. Tipton, Supervisor, Engineering Systems
- D. Traub, Technical Support Technician, Radiation Protection
- D. Waller, Supervising Engineer, Engineering Projects
- R. Wink, Supervisor, Regulatory Affairs
- T. Witt, Associate Engineer, Regulatory Affairs/Licensing
NRC Personnel
- R. Nease, Branch Chief, Division of Reactor Safety, Region II
- N. Carte, Senior Electronics Engineer, Office of Nuclear Reactor Regulation
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
- 05000483/2013005-01 URI Solid State Protection System Modifications (Section 1R17.1)
Opened and Closed
- 05000483/2013005-02 NCV Ultimate Heat Sink Cooling Tower - Incorrect Action Criteria in Control Room Evacuation Procedure (Section 1R17.2)
- 05000483/2013005-03 NCV Failure to Correct Non-conservative Safety Related Equipment Oil Leakage Criteria (Section 1R22)
- 05000483/2013005-04 NCV Failure to Class and Describe a Radioactive Material Shipment Correctly and Failure to Include all Required Hazard Communication Information in a Radioactive Shipment Document (Section 2RS8)
Closed
- 05000483/2012005-03 URI Determine Licensing Basis and Capability of One Vital Air Conditioning Unit to Cool Both Trains of Class 1E Electrical Equipment (Section 4OA5)