IR 05000498/2019001
| ML19128A334 | |
| Person / Time | |
|---|---|
| Site: | South Texas, 07201041  |
| Issue date: | 05/08/2019 |
| From: | Nick Taylor NRC/RGN-IV/DRP/RPB-B |
| To: | Gerry Powell South Texas |
| References | |
| IR 2019001 | |
| Download: ML19128A334 (33) | |
Text
May 8, 2019
SUBJECT:
SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION, UNITS 1 AND 2, NRC INTEGRATED INSPECTION REPORT 05000498/2019001; 05000499/2019001, AND INDEPENDENT SPENT FUEL STORAGE INSPECTION 07201041/2019001
Dear Mr. Powell:
On March 31, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your South Texas Project Electric Generating Station, Units 1 and 2. On April 11, 2019, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
One of these findings involved a violation of NRC requirements.
The inspectors documented a licensee-identified violation which was determined to be of very low safety significance in this report. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest the violation or significance or severity of the violation documented in this inspection report, 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; and the NRC resident inspector at South Texas Project Electric Generating Station, Units 1 and 2.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC resident inspector at South Texas Project. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Nicholas H. Taylor, Branch Chief Project Branch B Division of Reactor Projects
Docket Nos. 05000498; 05000499 and 07201041 License Nos. NPF-76 and NPF-80
Enclosure:
Inspection Report 05000498/2019001; 05000499/2019001 and 07201041/2019001 w/attachment: NRC Request for Information
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Number(s):
05000498; 05000499 and 07201041
License Number(s):
Report Number(s):
05000498/2019001; 05000499/2019001, and 07201041/2019001
Enterprise Identifier: I-2019-001-0004 and 1-2019-001-0082
Licensee:
STP Nuclear Operating Co.
Facility:
South Texas Project Electric Generating Station, Units 1 and 2
Location:
Wadsworth, TX 77483
Inspection Dates:
January 1, 2019, to March 31, 2019
Inspectors:
B. Baca, Health Physicist
J. Choate, Resident Inspector
A. Sanchez, Senior Resident Inspector
L. Brookhart, Senior ISFSI Inspector, FCDB
C. Smith, Reactor Inspector, DRS, EB1
O. Masnyk Bailey, Health Physicist, RI, DIRHB
M. Learn, Reactor Engineer, RIII, MCID
R. Edwards, Senior Health Physicist, RIII, MCID
Approved By:
Nicholas H. Taylor
Chief, Project Branch B
Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a quarterly inspection at South Texas Project Electric Generating Station, Units 1 and 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. NRC and self-revealed findings, violations, and additional items are summarized in the table below. Licensee-identified non-cited violations are documented in report sections:
7115
List of Findings and Violations
Failure to Implement the Procedure for Equipment Failures Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events Green FIN 05000499/2019-01 Closed
[H.12] - Avoid Complacency 71152 Problem Identification and Resolution The inspectors identified a Green finding for the licensees failure to implement the procedure for the failure of a plant generation risk high risk component. Specifically, the licensee failed to implement and follow Procedure WCG-0008, Preventing Recurring Equipment Problems (PREP), Revision 7, following the failure of the Unit 2 steam generator 2C normal feedwater regulating control valve, FCV-553.
Failure to Provide Adequate Procedural Guidance for a Surveillance Test Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events Green NCV 05000499/2019-02 Closed
[H.12] - Avoid Complacency 71152 Problem Identification and Resolution The inspectors documented a Green self-revealed NCV for the failure to provide an adequate procedure for a Unit 2 qualified display processing system (QDPS) surveillance procedure that resulted in unnecessary troubleshooting, unnecessary replacement of properly operating circuit cards, and placing the loop 2 over-temperature delta-T, over-pressure delta-T, and low Tavg feedwater isolation bistables into a TRIPPED condition, a half-trip condition for the reactor. Following an investigation, the licensee discovered that incorrect biases were inputted into the QDPS surveillance procedure.
Additional Tracking Items
Type Issue number Title Inspection Procedure Status LER 05000498/2017-002-00 Unit 1 Condition Prohibited by Technical Specifications due to Inoperable Control Room Envelope Makeup Filtration System Heating Coil 71153 Follow-up of Events Closed LER 05000498/2018-001-00 Unit 1 Main Steam Safety Valve As Left Settings Outside of Required Range Contrary to Technical Specifications due to Inadequate Procedure 71153 Follow-up of Events Closed
PLANT STATUS
Unit 1 operated at rated thermal power for the entire inspection period.
Unit 2 began the inspection period at rated thermal power. On March 2, 2019, while performing main turbine governor valve testing at 90 percent power, a procedural error resulted in an unexpected decrease in reactor power to 86 percent power. The reactor was returned back to rated thermal power on March 3, 2019, and remained there for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed plant status activities described in IMC 2515 Appendix D, Plant Status, and conducted routine reviews using IP 71152, Problem Identification and Resolution. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown (IP Section 02.01) (5 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 1, train B chemical and volume control system during train A centrifugal charging pump maintenance on January 2, 2019
- (2) Unit 2, turbine driven auxiliary feedwater system with elevated unit trip risk during north bus outage on March 1, 2019
- (3) Unit 2, train C emergency core cooling system during train B emergency diesel generator maintenance on March 13, 2019
- (4) Unit 2, train A essential chilled water system during train B essential chilled water system maintenance on March 16, 2019
- (5) Unit 2, train A essential cooling water system during train C essential cooling water maintenance on March 22, 2019
71111.05Q - Fire Protection
Quarterly Inspection (IP Section 03.01) (5 Samples)
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Unit 2, safety injection pump cubicles, Fire Area 35 on February 2, 2019
- (2) Units 1 and 2, balance of plant diesel generator rooms and turbine generator building switchgear rooms, Fire Areas 78 and 90 on February 26, 2019
- (3) Unit 1, auxiliary feedwater pump cubicles, Fire Areas 48, 49, 50, and 51 on February 27, 2019
- (4) Unit 1, train C essential cooling water intake structure, Fire Area 55 on March 22, 2019
- (5) Unit 1, train C vital switchgear room, battery room, 125 vdc distribution room, and motor generator room, Fire Areas 52 through 54 on March 26, 2019
71111.06 - Flood Protection Measures
Inspection Activities - Internal Flooding (IP Section 02.02a.) (1 Sample)
The inspectors evaluated internal flooding mitigation protections in the Unit 1, train B essential cooling water pump room on March 26, 2019.
71111.07A - Heat Sink Performance
Annual Review (IP Section 02.01) (1 Sample)
The inspectors evaluated readiness and performance of:
- (1) Unit 1, train C emergency diesel generator jacket water and lube oil heat exchangers the week of January 14, 2019
- (2) Unit 2, train B essential chilled water cooler the week of March 12, 2019
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
The inspectors observed and evaluated licensed operator performance in the Unit 1 control room during a secondary side electro-hydraulic leak which resulted in the closure of the 13 west reheat stop valve to the low pressure turbine on January 9, 2019.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
The inspectors observed and evaluated Unit 1 operations in the simulator during a loss of coolant accident followed by a loss of offsite power on February 27, 2019.
71111.12 - Maintenance Effectiveness
Routine Maintenance Effectiveness Inspection (IP Section 02.01) (2 Samples)
The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:
- (1) Unit 1, turbine driven auxiliary feedwater pump low governor oil level that resulted in inoperability of the train on January 3, 2019
- (2) Unit 1, electro-hydraulic control system leak that resulted in the closure of reheat stop valve 13 on January 9, 2019
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent work activities:
- (1) Units 1 and 2, planned north bus outage that resulted in entering the Configuration Risk Management Program on February 26, 2019
- (2) Unit 2, planned turbine driven auxiliary feedwater pump extent of condition maintenance for connecting rod engagement on January 31, 2019
- (3) Unit 1, train B essential cooling water unplanned pump replacement February 13 through 19, 2019
- (4) Unit 1, elevated risk due to unplanned entry into the Configuration Risk Management Program for exceeding the train B essential chilled water system 7 day limiting condition for operation due to maintenance activities on March 10, 2019
71111.15 - Operability Determinations and Functionality Assessments
Sample Selection (IP Section 02.01) (5 Samples)
The inspectors evaluated the following operability determinations and functionality assessments:
- (1) Unit 1, turbine driven auxiliary feedwater pump following the identification of a degraded conduit for the remote trip solenoid on January 16, 2019
- (2) Unit 1, train B control room envelope heating, ventilation, and air conditioning damper controller on February 6, 2019
- (3) Unit 2, cask loading pool leakage on February 12, 2019
- (4) Unit 1, train B emergency diesel generator shutdown when taken out of emergency mode on March 11, 2019
- (5) Unit 2, turbine driven auxiliary feedwater pump experiencing high main oil pump discharge pressure on March 15, 2019
71111.19 - Post Maintenance Testing
Post Maintenance Test Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the following post maintenance tests:
- (1) Unit 1, train C emergency diesel generator following fuel oil storage tank level transmitter replacement on January 31, 2019
- (2) Unit 2, turbine driven auxiliary feedwater pump following failure of connecting rod on January 31, 2019
- (3) Unit 1, train B essential cooling water pump following pump replacement on March 4, 2019
- (4) Unit 1, train B emergency diesel generator following replacement of shutdown air solenoid valve on March 13, 2019
- (5) Units 1 and 2, diesel fire pump number 3 following replacement of the pump impeller, diesel starter solenoid, and pump packing on March 18, 2019
- (6) Unit 2, train C residual heat removal pump following control room handswitch replacement on March 19, 2019
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance tests:
In Service Testing (IST) (IP Section 03.01)
- (1) Unit 1, turbine driven auxiliary feedwater pump surveillance test on January 24, 2019
- (2) Unit 1, train B essential cooling water system surveillance test on February 14, 2019
Surveillance Testing (IP Section 03.01) (4 Samples)
- (1) Unit 2, train B emergency diesel generator surveillance test on February 12, 2019
- (2) Unit 2, train A auxiliary feedwater regulating valve dynamic pressure test on March 4, 2019
- (3) Unit 1, train B emergency diesel generator fast start surveillance on March 13, 2019
- (4) Unit 1, train C emergency diesel generator fast start surveillance on March 15,
RADIATION SAFETY
71124.02 - Occupational ALARA Planning and Controls
Radiological Work Planning (IP Section 02.01) (1 Sample)
The inspectors evaluated the licensees radiological work planning by reviewing the following activities:
- 1RE21 - Reactor Cavity and Head Activities
- 1RE21 - Fuel Transfer Canal Activities
- 2RE19 - Reactor Cavity and Head Activities
- 2RE19 - Steam Generator Activities
Verification of Dose Estimates and Exposure Tracking Systems (IP Section 02.02) (1 Sample)
The inspectors evaluated dose estimates and exposure tracking. The inspectors reviewed the following ALARA planning documents:
- 18-276-7 ALARA Review Package: 2RE19 Non-Rapid Refuel, Revision 1
- 18-276-8 ALARA Review Package: 2RE19 Steam Generator Inspections
- 18-7926-4 ALARA Review Package: 1RE21 Non-Rapid Refuel
Additionally, the inspectors reviewed the following radiological outcome evaluations:
- 18-276-7 ALARA Close Out Review Package: 2RE19 Non-Rapid Refuel, Revision 1
- 18-276-8 ALARA Close Out Review Package: 2RE19 Steam Generator Inspections
- 18-7926-4 ALARA Close Out Review Package: 1RE21 Non-Rapid Refuel
- 18-7926-4 ALARA Close Out Review Package Supplement 1: 1RE21 Non-Rapid Refuel
71124.04 - Occupational Dose Assessment
External Dosimetry (IP Section 02.02) (1 Sample)
The inspectors evaluated the external dosimetry program implementation.
Internal Dosimetry (IP Section 02.03) (1 Sample)
The inspectors evaluated the internal dosimetry program implementation.
Three contaminated workers' whole body counts were reviewed for potential uptakes. The licensee had no positive whole body counts indicating an internal uptake of radioactive material.
The licensee had no occurrences to observe or review in-vitro internal monitoring or dose assessments performed using air sampling and DAC-hr. monitoring.
Source Term Categorization (IP Section 02.01) (1 Sample)
The inspectors evaluated the licensees characterization of the source term and use of scaling factors for the use of hard-to-detect radionuclide activity.
Special Dosimetric Situations (IP Section 02.04) (1 Sample)
The inspectors evaluated special dosimetric situations, as detailed below.
- The inspectors reviewed approximately five declared pregnant workers' documentation from January 1, 2017, to March 21, 2019.
- The licensee did not perform any EDEX assessments from July 31, 2017, to March 21, 2019.
- The inspectors reviewed one shallow dose equivalent assessment (Condition Report 2019-1215).
- The inspectors reviewed the licensee's use of Mirion neutron thermoluminescent dosimetry and electronic alarming dose meters (Model EPDN2) for neutron dosimetry. The inspector compared radiological surveys, RWP and WAN assignments, and worker exposures to confirm the licensee is appropriately assigning neutron dose to occupational workers.
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below.
=
- (1) IE01: Unplanned Scrams per 7000 Critical Hours Sample (Units 1 and 2, January - December 2018
- (2) IE03: Unplanned Power Changes per 7000 Critical Hours Sample (Units 1 and 2, January - December 2018
- (3) IE04: Unplanned Scrams with Complications Sample (Units 1 and 2, January - December 2018)
71152 - Problem Identification and Resolution
Annual Follow-up of Selected Issues (IP Section 02.03) (4 Samples)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) Unit 2, pressurizer master controller card failure, Condition Report 2017-658.
Inspectors reviewed the plant response to the event, operator actions, corrective actions, and interviewed licensee personnel.
- (2) Unit 1, low head safety injection breaker failure, Condition Report 2017-20444.
Inspectors reviewed the apparent cause investigation, operability evaluation, and corrective actions.
- (3) Unit 2, train C steam generator main control valve failed low in automatic mode, Condition Report 2017-17659. Inspectors reviewed the plant response to the event, operator actions, maintenance and engineering troubleshooting repair of the issue, interviewed licensee personnel and corrective actions.
- (4) Unit 2, train D qualified display processing system calibration procedure that resulted in exceeding the Technical Specification allowed outage time and placed the bistables into a trip condition, Condition Reports 2016-11346 and 2016-11257. The inspectors reviewed the prompt investigation, interviewed maintenance and engineering personnel, and evaluated the corrective actions.
71153 - Follow-up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event reports which can be accessed at https://lersearch.inl.gov/LERSearchCriteria.aspx:
- (1) LER 05000498/2017-002-00, "Unit 1 Condition Prohibited by Technical Specifications due to Inoperable Control Room Envelope Makeup Filtration System Heating Coil,"
on January 22, 2018:
- (2) LER 05000498/2018-001-00, "Unit 1 Main Steam Safety Valve As Left Settings Outside of Required Range Contrary to Technical Specifications due to Inadequate Procedure," on December 4, 2018:
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
60854 - Preoperational Testing of an Independent Spent Fuel Storage Inspection (ISFSI)60855 - Operation of an ISFSI at Operating Plants 60857 - Review of 10 CFR 72.48 Evaluations
Inspections of dry cask storage operations were conducted at STP on January 21-25 and on January 28 through February 2, 2019, by inspectors from NRC Regions I, III, and IV. The STP ISFSI was licensed as a general 10 CFR Part 72 license and utilized the Holtec HI-STORM FW Certificate of Compliance (COC) 1032 Amendment 2 and HI-STORM FW Final Safety Analysis Report (FSAR) Revision 5.
In prior NRC ISFSI inspections at STP (ADAMS Accession Nos. ML18221A385 and ML19032A079), the NRC staff documented the observations and evaluations of many of the required pre-operational demonstrations from License Condition No. 9 of the COC performed by STP. Those inspections included an in-depth review of the licensees dry cask storage program implementation into existing 10 CFR Part 50 programs to support ISFSI operations at the site. The purpose of the NRC inspections in January and February 2019 were to:
- (1) observe and evaluate the remaining pre-operational demonstrations and training exercises which STP was required to complete prior to commencing loading operations, and
- (2) observe and evaluate the licensees first loading operations. On January 21-25, 2019, the NRC staff observed STP complete the remaining demonstrations listed in License Condition No. 9 of the COC to allow use of the Holtec HI-STORM FW storage system. South Texas Project had successfully completed all the required pre-operations activities and fully demonstrated that the procedures, programs, and training related to those dry cask storage operations had been successfully integrated into its site operations. The subsequent week, from January 28 to February 2, 2019, NRC staff observed the licensees first canister loading operations.
The ISFSI activities specifically reviewed during the on-site inspections and during in-office review included:
1. Evaluated and observed dry run demonstrations from License Condition No. 9 of the
COC, which included: preparation of the canister storage system for fuel loading; loading a dummy fuel assembly into the canister with appropriate independent verification; selection and verification of the specific fuel assemblies to ensure type conformance; remote installation of the canister lid; removal of the canister from the spent fuel pool; and unloading operations that included remotely removing a lid from the canister to support fuel unloading.
2. Reviewed the licensees structural and seismic calculations for dry cask storage
operations which included: over-head spent fuel building crane with loaded Transfer Cask (TC); loaded TC in the spent fuel pools loading area; loaded TC in the canister wash-down area for processing operations; loaded TC in the stack-up configuration to support downloading of a canister into the concrete over-pack; loaded over-pack on low-profile transporter; and loaded over-pack being carried by the vertical cask transporter to the ISFSI pad.
3. Reviewed two 10 CFR 72.48 safety evaluations performed by the licensee for changes
made to the ISFSI program since the previous NRC ISFSI inspection per Inspection Procedure 60857.
4. Reviewed spent fuel documentation for the first canister to verify the fuel met all
Appendix B Technical Specifications requirements for storage.
5. Evaluated and observed fuel selection and fuel loading operations associated with dry
fuel storage canister No. 1.
6. Evaluated and observed welding of the canister, non-destructive testing of the welds,
forced helium drying, helium backfill, and heavy load movements from the spent fuel pool to the sites vertical cask transporter.
7. Reviewed the licensee's loading, processing, and heavy load procedures associated
with their current dry fuel storage campaign.
8. Reviewed the licensees corrective action programs condition reports of issues identified
for resolution since the previous NRC ISFSI inspection performed in November 2018.
9. Evaluated and observed the licensees radiation protection implementation during the
dry run and loading operations.
10. Evaluated and observed the licensees implementation of foreign material exclusion process during the dry run and loading operations.
The inspectors did not identify any issues of concerns requiring documentation.
INSPECTION RESULTS
Failure to Implement the Procedure for Equipment Failures Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events
Green FIN 05000499/2019-01 Closed
[H.12] - Avoid Complacency 71152 Problem Identification and Resolution The inspectors identified a Green finding for the licensees failure to implement the procedure for the failure of a plant generation risk (PGR) high risk component. Specifically, the licensee failed to implement and follow Procedure WCG-0008, Preventing Recurring Equipment Problems (PREP), Revision 7, following the failure of the Unit 2 steam generator 2C normal feedwater regulating control valve FCV-553.
Description:
On June 19, 2017, Unit 2 reactor operators responded to control room alarms and indications for steam flow/ feed flow mismatch for steam generator 2C. Operators identified that the steam generator 2C main feedwater regulating valve had failed closed and that steam generator level was decreasing for that steam generator. Operators entered the appropriate off-normal procedure and within 28 seconds regained control of steam generator levels by manually controlling feedwater flow through the main feedwater regulating valve. As a result of the failure, reactor power lowered to approximately 97 percent, average reactor coolant system temperature (Tavg) reached a maximum valve of 592.9 degrees F (normally Tavg is 592.0), and steam generator 2C narrow range level lowered to 52.8 percent (normally 70 percent).
Following the failure, engineering and maintenance discussed and implemented changing out a suspected nuclear tracker-driver (NTD) circuit card. This type of activity in the 7300 control cabinets is a high risk evolution and required a work activity risk review to help ensure no mistakes were made and that operations prepared for the worst case outcome. South Texas Project (STP) has site specific operating experience where card calibration and card replacement activities have resulted in reactor trips. Engineering was confident that the NTD circuit card was the problem. Maintenance personnel suggested measuring voltage readings at various locations inside the panel to ensure the faulty card was identified, however these measurements were not taken due to engineerings position that the NTD card was faulty.
The inspectors attended the pre-job brief and observed the field work. Maintenance performed the card replacement successfully, but the issue remained. Maintenance and engineering regrouped and decided to take voltage measurements to identify the circuit card failure. Maintenance identified the nuclear control board (NCB) as the problem and replaced the NCB without issue. STP performed the high-risk evolution twice due to inadequate troubleshooting and over-reliance on engineering assumptions.
The licensee evaluated the operators response to the main feedwater regulating valve failure and engineering and maintenances response to troubleshooting and repair of the failed circuit card. The inspectors interviewed engineering and maintenance and determined that the licensee should have entered and implemented Procedure WCG-0008 following the failure of the Unit 2 steam generator 2C normal feedwater regulating control valve, FCV-553.
The procedure would have ensured the appropriate degree of rigor was applied when pursuing resolution of high risk equipment issues. Specifically, Step 1.3 states, in part, The PREP process entry criteria are: Loss of Key Equipment function/ Critical Attribute.
The failure of the main feedwater regulating valve was a loss of key equipment function. The PREP process would have engaged more management oversight, directed the gathering of facts through Addendum 3, Preventing Recurring Equipment Problems (PREP) checklist, and may have led to the development of a troubleshooting plan.
Corrective Actions: The licensee replaced the failed card and developed a lessons-learned presentation for engineering to use human performance tools, validate assumptions, and challenge recommendations (use technical conscience).
Corrective Action References: Condition Reports 2017-17659 and 2019-4368
Performance Assessment:
Performance Deficiency: The failure to enter and use station Procedure WCG-0008, Preventing Recurring Equipment Problems (PREP), Revision 7, following the failure of the Unit 2 steam generator 2C normal feedwater regulating control valve, FCV-553, was a performance deficiency. Specifically, the licensee failed to enter WCG-0008 upon the failure of the main feedwater regulating valve, a loss of key equipment function due to the component being classified as a high PGR. The result of not implementing this procedure was an increased probability of tripping the reactor due to possible human error while in the 7300 control cabinet.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the human performance attribute and adversely affected the Initiating Events Cornerstone objective to limit the likelihood of events that can upset plant stability during power operations and is therefore a finding. Specifically, engineering relied on past experience and did not consider entering and implementing the PREP process procedure, which would have applied more rigor to troubleshooting and resolving the issue.
Significance: The inspectors used IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, issued June 19, 2012, to determine the finding was of very low safety significance, Green. Specifically, one finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
Cross-cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
Failure to Provide Adequate Procedural Guidance for a Surveillance Test Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events
Green NCV 05000499/2019-02 Closed
[H.12] - Avoid Complacency 71152 Problem Identification and Resolution The inspectors documented a Green self-revealed NCV for the failure to provide an adequate procedure for a Unit 2 qualified display processing system (QDPS) surveillance procedure that resulted in unnecessary troubleshooting, unnecessary replacement of properly operating circuit cards, and placing the loop 2 over-temperature delta-T (OTDT), over-pressure delta-T (OPDT), and low Tavg feedwater isolation bistables into a TRIPPED condition, a half-trip condition for the reactor. Following an investigation, the licensee discovered that incorrect biases were inputted into the QDPS surveillance procedure.
Description:
On September 15, 2016, operations declared the loop 2 OTDT, OPDT, and the low Tavg feedwater isolation inoperable for loop calibration surveillance testing and bypassed the channel per Technical Specifications 3.3.1, items 8 and 9, action 6, and 3.3.2, item 5F, action 20. If not restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the channel must be placed into a TRIPPED condition. During the performance of Surveillance Procedure 0PSP05-RC-0420, Delta T and T average loop 2 Set Calibration (T-0420), Revision 53, maintenance identified an out-of-tolerance condition. Engineering recommended maintenance perform a QDPS output calibration using Procedure 0PMP08-AM-APCD2, QDPS Train D SGWLCS/TAS Input/Output Calibration, Revision 10. During this calibration, the licensee again identified an out-of-tolerance condition which could not be resolved. The licensee performed troubleshooting and began to replace circuit boards in the train D QDPS cabinet without success. On September 18, 2016, the bistables for the OTDT, OPDT, and Low Tavg feedwater isolation were placed into a TRIPPED condition as required by technical specifications.
The licensee subsequently discovered that an engineering error resulted in incorrect biases being placed into the Calibration Procedures 0PSP05-RC-0420 and 0PMP08-AM-APCD2.
The two procedures were corrected with the proper biases and completed satisfactorily. The manual substitution of the resistance temperature detector (RTD) biases in the two calibration procedures had never been performed before and, due to the complex nature of the QDPS system, only three people on site could perform and peer check these biases. The licensee performed a prompt investigation and a review for organizational learnings. The licensee determined that the engineer made changes to the procedure to account for the two RTD operation, but failed to get the values peer checked for accuracy. This failure led to excessive out-of-service time, unnecessary QDPS maintenance, and having to place the loop 2 protection bistables into a TRIPPED condition, i.e. half reactor trip.
Corrective Actions: The licensee immediately corrected the procedures and re-performed the calibrations and satisfactorily performed analog channel operational testing. The licensee also developed a lessons-learned document and has an outstanding action to develop an engineering guideline when either unit is in two RTD operation.
Corrective Action References: Condition Reports 2016-11346, 2016-11257, 2019-4367
Performance Assessment:
Performance Deficiency: The failure to provide adequate procedures for maintenance on safety-related equipment was a performance deficiency. Specifically, engineering supplied inaccurate information to calibration procedures for the Unit 2 loop 2 and train D QDPS that resulted in unnecessary maintenance in QDPS and TRIPPED bistables associated with OTDT, OPDT, and low Tavg feedwater isolation which caused Unit 2 to be in a half reactor trip.
Screening: The inspectors determined the performance deficiency was more than minor because it is associated with the Procedure Quality attribute and adversely affected the Initiating Event Cornerstone objective to limit the likelihood of events that can upset plant stability during power operations and is therefore a finding. Specifically, the loop 2 bistables associated with OTDT, OPDT, and low Tavg were placed into a TRIPPED condition which placed Unit 2 one spurious reactor trip signal away from a true reactor trip due to the performance deficiency.
Significance: The inspectors used IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, issued June 19, 2012, to determine the finding was of very low safety significance, Green. Specifically, one finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
Cross-cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. The inspectors determined that the finding had a crosscutting aspect in the area of human performance associated with avoiding complacency. Specifically, engineering provided inaccurate biases for the procedure changes without utilizing appropriate error reduction tools [H.12].
Enforcement:
Violation: Technical Specification 6.8.1.a requires, in part, that written procedures shall be established, implemented, and maintained in accordance with Appendix A of Regulatory Guide 1.33, Revision 2, February 1978, Section 8.b.(L) reactor protection system tests and calibrations. The licensee established Procedure 0PSP05-RC-0420, Delta T and T average loop 2 Set Calibration (T-0420), Revision 53 to meet the Regulatory Guide 1.33 requirement.
The Unit 2 data package loop 2 set 2 (T-0420) provides appropriate biases to be used to test and calibrate the loop.
Contrary to the above, on September 15, 2016, the licensee failed to ensure that the Unit 2 data package loop 2 set 2 (T-0420) provided the appropriate biases to be used to test and calibrate the loop. Specifically, engineering failed to provide accurate biases for procedures used for surveillance and calibration of the OTDT, OPDT, and low Tavg feedwater isolation trip setpoints.
Enforcement Action: This violation is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.
Minor Violation 71153 Follow-up of Events Minor Violation: Technical Specification 3.7.7, Control Room Makeup and Cleanup Filtration System, states, in part, that if one Control Room Makeup and Cleanup Filtration System is inoperable, it must be restored to operable in Modes 1, 2, 3, and 4 within 7 days, or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Contrary to the above, between September 27 and November 24, 2017, the licensee failed to restore one control room makeup and cleanup filtration system to operable. Specifically, the licensee failed to properly jumper a circuit board for the control room makeup and cleanup filtration system heating coil.
Screening: The inspectors determined the performance deficiency was minor. The inspectors found this violation to be minor in accordance with IMC 0612 Appendix B, Issue Screening, issued December 13, 2017, because the violation did not satisfy any of the four More-than-Minor screening questions.
Enforcement:
This failure to comply with Technical Specification 3.7.7 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
Licensee-Identified Non-Cited Violation 71153 Follow-up of Events This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Technical Specification 3.7.1.1, Safety Valves, states, in part, that if one or more main steam line code safety valves are inoperable, they must be restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or reduce the Power Range Neutron Flux High Trip Setpoint or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Contrary to the above, between October 2 and 6, 2018, the licensee failed to restore two of the Unit 1 steam generator B main steam safety valves to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, failed to reduce the Power Range Neutron Flux High Trip Setpoint, and failed to be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Specifically, the licensee failed to identify that two of the Unit 1 Steam Generator B main steam safety valves were left outside of their Technical Specification required lift settings following testing.
Significance: Green.
The inspectors determined the performance deficiency was more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined this finding did not represent an actual loss of function. Therefore, the inspectors determined the finding was of very low safety significance (Green).
Corrective Action References: Condition Reports 2018-11811 and 2019-4374
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 11, 2019, the inspector presented the quarterly resident inspector inspection results to Mr. G. T. Powell, President and Chief Executive Officer, and other members of the licensee staff.
- On March 21, 2019, the inspector presented the baseline radiation safety inspection to Mr. J. Connolly, Executive Vice President and Chief Nuclear Officer, and other members of the licensee staff.
- On March 21, 2019, the inspector presented the pre-operational dry run and first loading inspection to Mr. J. Connolly, Executive Vice President and Chief Nuclear Officer, and members of the licensee staff.
DOCUMENTS REVIEWED
71111.01-Adverse Weather
Procedures
Number
Title
Revision
Severe Weather Plan
71111.04-Partial Equipment Alignment
Condition Reports
CR-2018-2825
Procedures
Number
Title
Revision
Safety Injection System Initial Lineup
Essential Chilled Water Pump 11A(21A) Inservice Test
Essential Cooling Water Operations
Drawings
Number
Title
Revision
5R179F05007#1
Chemical and Volume Control System
5S142F00024#1
5N129F05015#2
Safety Injection System
5V119V10001#2
HVAC Essential Chilled Water System
5R289F05038#2
Essential Cooling Water System Train 2B
71111.05-Fire Protection
Procedures
Number
Title
Revision
Fire Preplan Fuel Handling Building Train B SI/CSS Cubicle 2
Fire Preplan Fuel Handling Building Train A SI/CSS Cubicle 2
STPEGS Fire Brigade
Fire Preplan for Make-Up Demineralizer Building
Fire Preplan Turbine Generator Building BOP Diesel
Generator Building
Fire Preplan for Turbine Generator Building 4.16 KV and
Electrical Equipment Rooms
Procedures
Number
Title
Revision
Fire Preplan Turbine Generator Building 13.8 KV Switchgear
Room and Cable Vault
Fire Preplan Isolation Valve Cubicle Pump Room Train C
Fire Preplan Isolation Valve Cubicle Pump Room Train B
Fire Preplan Isolation Valve Cubicle, Pump Room Train A
Fire Preplan Isolation Valve Cubicle, Pump Room Train D
Fire Preplan Electrical Auxiliary Building ESF Switchgear
Room Train C
Fire Preplan Electrical Auxiliary Building, Channel IV Battery
and Distribution Room
Fire Preplan Electrical Auxiliary Building, Motor Generator
Room
Fire Preplan Essential Cooling Water Intake Structure Pump
Room Train C
71111.07A - Annual Heat Sink
Procedures
Number
Title
Revision
York Chiller Inspection & Maintenance 300 Tons
Essential Cooling Water System Reliability Program
Work Authorization Numbers
443062
71111.11-Simulator/Control Room Observation
Condition Reports
CR-2019-364
CR-2019-423
Procedures
Number
Title
Revision
Main Turbine EHCS
Main Turbine Steam Inlet Valve Test
Number
Title
Revision
6S109F00017#1
Number
Title
Revision
7T089F10001#1
P&ID EH Fluid System
71111.12-Maintenance Effectiveness
Condition Reports
CR-2015-18885
CR-2017-15182
CR-2017-23544
CR-2018-995
CR-2018-8533
CR-2018-1108
CR-2018-12813
CR-2018-15270
CR-2018-15273
Work Authorization Numbers
499003
519391
21829
2330
540499
548908
568084
603714
Procedures
Number
Title
Revision
Auxiliary Feedwater Pump Turbine Maintenance
Auxiliary Feedwater Turbine Trip Throttle Valve Maintenance 36
Auxiliary Feedwater Pump 14(24) Inservice Test
Drawings
Number
Title
Revision
6S109F00017#1
Piping and Instrumentation Diagram Main Steam
7T089F10001#1
Piping and Instrumentation Diagram Electro Hydraulic Fluid
System
71111.13-Maintenance Risk and Emergent Work
Procedures
Number
Title
Revision
Configuration Risk Management Program
Risk Management Actions (RMAs)
Work Activity Risk (WAR)
2804
RasCal Sequences
3331
3336
3337
71111.15-Operability Determinations and Functionality Assessments
Condition Reports
CR-2019-47
CR-2019-347
CR-2019-1377
19-2777
19-637
Drawings
STI. Number
Title
Date
31390502
Cabinet-Control (Assembly Engine Control)
2/27/1978
298485
Control Schematic (Starting Sequence Control)
09/29/1977
71111.19-Post Maintenance Testing
Condition Reports
CR-2018-7676
CR-2018-9443
CR-2018-14727
CR-2019-111
CR-2019-893
CR-2019-1762
CR-2019-2357
CR-2019-2435
CR-2019-2777
CR-2019-1020
Work Authorization Numbers
545649
565703
571744
571827
2175
576480
2665
604208
604273
606453
573948
605423
Procedures
Number
Title
Revision
Standby Diesel Generator Fuel Oil Storage Tank Inspection 9
ESF Diesel Generator Fuel Oil Storage Tank Level
Calibration
Essential Cooling Water Pump 1B(2B) Reference Values
Measurement
Auxiliary Feedwater Pump 14(24) Inservice Test
Diesel Fire Pump Test
Fire Protection Water System Functional Test
Standby Diesel 12(22) Operability Test
Miscellaneous
Documents
Number
Title
Revision
VTD-T343-0001
Installation and Service Instructions for Tuthill Cartridge
Pumps
VTD-P104-0011
Patterson Operation and Maintenance Manual for Double
Suction Split Case Pumps
71111.22-Surveillances
Condition Reports
CR-2019-975
CR-2019-1762
Procedures
Number
Title
Revision
Auxiliary Feedwater Pump 14(24) Inservice Test
MOV Diagnostic Testing (VOTES) - Rising Stem Valves
Dynamic Stroke Testing of Train A Auxiliary Feed Motor
Operated Feed Regulating Valve
Standby Diesel 13(23) Operability Test
Standby Diesel Recording M&TE Installation
Standby Diesel 12(22) Operability Test
Essential Cooling Water System Train B Testing
71124.02 - Occupational ALARA Planning and Controls
Condition Reports
CR-2018-3840
CR-2018-3991
CR-2018-4050
CR-2018-4125
CR-2018-4555
CR-2018-4860
CR-2018-5849
CR-2018-6071
CR-2018-8097
CR-2018-9179
CR-2018-9284
CR-2018-11217
CR-2018-12339
CR-2018-13572
CR-2019-234
CR-2018-1066
Procedures
Number
Title
Revision
Radiation Protection Program
ALARA Program
ALARA Engineering and Procedure Review
Radiation Work Permits/Radiological Work ALARA Reviews 38
Procedures
Number
Title
Revision
Radiological Briefings
Radiological Risk Management
Conduct of Operations for Radiation Protection, Chapter 18:
ALARA Planning
Audits and Self-Assessments
Number
Title
Date
18-15289-1
Snapshot Self-Assessment using NRC 71124, Att. 2
2/28/2019
MN-19-0-107465
Radiation Protection Department Annual Evaluation
2/26/2019
Excessive Personnel Radiation Exposures; Unplanned
Radiation Exposures
2/05/2018
Radiation Work Permits
Number
Title
Revision
2018-1-296
1RE21 - Transfer Cart Repair in the FHB / RCB Fuel Transfer Canals
including RP Surveys and Decontamination
2018-1-296
1RE21 - Transfer Cart Repair in the FHB / RCB Fuel Transfer Canals
including RP Surveys and Decontamination
2018-2-129
2RE19 - Steam Generator-Primary Side Support and Equipment
Setup/Tear Down (HRA) - Medium Radiological Risk
2018-2-130
2RE19 - Steam Generator-Primary Side Manway/Inserts Removal
(LHRA) - High Radiological Risk
2018-2-131
2RE19 - Steam Generator-Primary Side Nozzle Dams (LHRA) - High
Radiological Risk
2018-2-131
2RE19 - Steam Generator-Primary Side Nozzle Dams (LHRA) - High
Radiological Risk
2018-2-132
2RE19 - Steam Generator-Eddy Current Testing (LHRA) - High
Radiological Risk
2018-2-133
2RE19 - Steam Generator-Re-Install Primary Side Manway (HRA) -
Medium Radiological Risk
2018-2-135
2RE19 - Steam Generator-Secondary Side Inspections (HRA)
ALARA In-Progress Reviews
Review
Number
Title
Date
249
Stud Detensioning
10/11/2018
250
HP Job Coverage
10/12/2018
ALARA In-Progress Reviews
Review
Number
Title
Date
254
HP Job Coverage in High Radiation Areas
10/21/2018
255
Request PMPI to provide craft support for the removal and
installation of mechanical snubbers per 0PMP04-SN-0001 in support
of snubber testing
10/22/2018
257
Contingency for Repairs to Unit 1 Fuel Handling Building Side Fuel
Transfer System
10/25/2018
259
Fuel Movement/Reactor Containment Building (RCB)
10/28/2018
ALARA Reviews
Number
Title
Date
18-276-4
ALARA Review: 2RE19 Snubber Inspections
03/02/2018
18-276-4
ALARA Close Out Review: 2RE19 Snubber Inspections
05/24/2018
18-276-8
ALARA Review: 2RE19 Steam Generator Inspections, Revision 1 04/04/2018
18-276-8
ALARA Close Out Review: 2RE19 Steam Generator Inspections
05/24/2018
18-7926-4
ALARA Review: 1RE21 Non-Rapid Refuel
10/02/2018
18-7926-5
ALARA Review: 1RE21 Flow Accelerated Corrosion Inspections -
In Service Inspections
09/12/2018
18-7926-6
ALARA Review: 1RE21 Room 003 Activities
09/26/218
19-234
ALARA Review: Dry Cask Storage
01/16/2019
19-234
ALARA Review: Initial DCS Campaign - Dry Cask Storage
Supplement 1
01/25/2019
Miscellaneous
Documents
Number
Title
Date
Refuel Outage 1RE20: ALARA Report
09/06/2017
Refuel Outage 2RE18: ALARA Report
05/18/2017
Refuel Outage 2RE19: ALARA Report
07/18/2018
71124.04Occupational Dose Assessment
Condition Reports
CR-2017-21141
CR-2018-2745
CR-2018-3052
CR-2018-4574
CR-2018-6875
CR-2018-8097
CR-2018-8247
CR-2018-10111
CR-2018-10591
CR-2018-10771
CR-2018-14460
CR-2018-14735
CR-2018-15144
CR-2019-1215
CR-2019-1254
Procedures
Number
Title
Revision
Personnel Dosimetry Program
Radiation Protection Program
Evaluation of Intakes
Personnel Exposure Investigation
Determination of Skin Dose
Dose to the Embryo/Fetus
Non-Routine Dosimetry Issue and Control
Calibration of the Siemens Environmental Systems Limited /
Thermo Electron Corporation Electronic Personal Dosimeter
Audits and Self-Assessments
Number
Title
Date
100555-0
Mirion Technologies (GDS), Inc. Onsite NVLAP Assessment Report 01/09/2018
Snapshot Assessment: Dosimetry Program 2018
2/04/2018
Snapshot Self-Assessment using NRC Inspection
Procedure 71124.04 Occupational Dose Assessment Criteria
2/19/2019
Personnel Exposure Records
2017-07
2017-28
2017-29
2017-30
2017-49
2018-08
2018-09
2018-10
2018-11
2018-23
2018-24
2018-35
2018-40
2018-41
2018-42
2018-43
2018-44
2018-45
71152-Issue Follow-up
Condition Reports
CR-2008-9639
CR-2019-893
CR-2019-1762
CR-2019-1954
CR-2019-2357
Work Authorization Numbers
335148
2175
576480
604273
606453
607307
Procedures
Number
Title
Revision
Essential Cooling Water Pump Maintenance (Product-
Lubricated Bearing Design)
Essential Cooling Water Operations
Miscellaneous
Documents
Number
Title
Revision
or Date
VTD-H127-0006
VSN Centrifugal Pump Installation and Operation
Manual
DCP 06-15147-45
Replace Essential Cooling Water Pump 2B
2/26/2009
4OA5.1 Other Activities (IP 60854, IP 60855, IP 60857)
Condition Reports
CR-2018-14418
CR-2018-14718
CR-201814770
CR-2019-540
CR-2019-630
CR-2019-631
CR-2019-665
CR-2019-706
CR-2019-816
Procedures
Number
Title
Revision
Fuel Selection for Dry Cask Storage
HI-STORM and MPC Pre-Use Inspections
MPC Loading Operations
MPC Closure Operations
MPC Transfer Operations
MPC Unloading Operations
HI-STORM Transport Operations
DCS Abnormal Response
PI-OP-HLTC-H-01
PCI Closure Welding of MPC Procedure
GQP - 9.2
PCI Liquid Penetrant Procedure
GQP - 9.6
PCI Visual Examination of Welds Procedure
Condition Reporting Process
Vendor Overview Activities (numerous)
Radiological Controls for Dry Cask Storage
N/A
ALARA Review Package
Dry Cask Storage Program Implementation
Procedures
Number
Title
Revision
10CFR72.48 Screening and Evaluations
WPS 8 MN-GTAW
ASME Section XI Welding Procedure Specification
WPS 8 MC-GTAW
ASME Section XI Welding Procedure Specification
PI-CNSTER-OP-
HLTC-H-01
Closure Welding of Holtec Multi-Purpose Canisters
Fuel Selection for Dry Cask Storage
Design Basis
Documents
Number
Title
Revision
HI-2114830
Final Safety Analysis Report on the HI-STORM FW MPC
Storage System
N/A
Certificate of Compliance No. 1032
2.212
STP ISFSI 10 CFR 72.212 Report
Miscellaneous
Documents
Number
Title
Revision
or Date
2.48 Screen
Initial Issue of STP ISFSI 10 CFR 72.212 Report
2.48 Eval
Initial Issue of STP ISFSI 10 CFR 72.212 Report
2.48 Eval
Non-Mechanistic Tipover Analysis for STP
VTI B056404
Design Report Stainless Steel Gate Cask Channel
58171-14-107
Whiting Corp Seismic Analysis Derived Values for 150 ton
Crane
04/13/2015
5817-14-050
Seismic Analysis for STP 150 ton Crane
CC09978
FHB Crane Seismic Analysis Decoupling Evaluation
B05315-0005
Seismic Stability HI-STORM/HI-TRAC Stack-up at STP
B05315-0003
STP Low Profile Transporter Analysis
B05315-00062
Time History Generation for STP
B05315-00066
Dynamic Analysis of HI-TRAC in Cask Wash-down area
B05315-00068
Dynamic Analysis of HI-TRAC in Cask Loading Pool
B05315-00069
Structural evaluation of Cask Loading and wash-down
areas
B05315-00071
Structural evaluation of Cask Loading Pool Pedestal
B05315-00073
Structural analysis of FHB Truck Bay Slab
B05315-00076
Safety Analysis of VCT Stability and Structural Evaluation
Miscellaneous
Documents
Number
Title
Revision
or Date
N/A
CK0001 - Fuel Selection for Dry Cask Storage
08/30/2018
N/A
CK0002 - Fuel Selection for Dry Cask Storage
08/30/2018
N/A
CK0003 - Fuel Selection for Dry Cask Storage
08/30/2018
N/A
CK0004 - Fuel Selection for Dry Cask Storage
08/30/2018
N/A
CK0005 - Fuel Selection for Dry Cask Storage
08/30/2018
RRTI 2319-11R0
Response to Request for Technical Information
2/12/2019
50.59 Screen 13-
6991-1 Supp 5 & 13-
6991-2 Supp 6
Fuel Handling Building Cask Loading Pool (CLP) and
Cask Connecting Channel (CCC) Gate and Liner
Modification
2/27/2018
DCP 13-6991-1
Supp. 0
Design Report Stainless Steel Gate Cask Connecting
Channel
10/08/2015
WO No. 96009469
HI Storm Overpack #6
01/15/2019
G. Powell
The following items are requested for the
Occupational Radiation Safety Inspection
at South Texas Project
Dates of Inspection: 03/18/2019 to 03/22/2019
Integrated Report 2019001
Inspection areas are listed in the attachments below.
Please provide the requested information on or before Monday, March 04, 2019.
Please submit this information using the same lettering system as below. For example, all
contacts and phone numbers for Inspection Procedure 71124.02 should be in a file/folder titled
2-A, applicable organization charts in file/folder 2-B, etc.
The information should be provided in electronic format or a secure document management
service. If information is placed on a secured document management system, please ensure
the inspection exit date entered is at least 30 days later than the onsite inspection dates, so the
inspectors will have access to the information while writing the report.
In addition to the corrective action document lists provided for each inspection procedure listed
below, please provide updated lists of corrective action documents at the entrance meeting.
The dates for these lists should range from the end dates of the original lists to the day of the
entrance meeting.
If more than one inspection procedure is to be conducted and the information requests appear
to be redundant, there is no need to provide duplicate copies. Enter a note explaining in which
file the information can be found.
If you have any questions or comments, please contact Bernadette Baca at 817-200-1235 or via
e-mail at Bernadette.Baca@nrc.gov.
PAPERWORK REDUCTION ACT STATEMENT
This letter does not contain new or amended information collection requirements subject
to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing information
collection requirements were approved by the Office of Management and Budget,
control number 3150-0011.
2.
Occupational ALARA Planning and Controls (71124.02)
Date of Last Inspection:
March 26, 2018
A.
List of contacts and telephone numbers for ALARA program personnel, as well as the
Licensing/Regulatory Affairs staff. Please include area code and prefix. If work cell
numbers are appropriate, then please include them as well.
B.
Applicable organization charts including position or job titles. Please include as
appropriate for your site, Site Management, RP, Chemistry, Maintenance (I&C),
Engineering, and Emergency Protection. (Recent pictures are appreciated.)
C.
Copies of audits, self-assessments, LARs, and LERs, written since the date of last
inspection, focusing on ALARA
D.
Procedure index for ALARA Program procedures and other related disciplines.
E.
Please provide specific procedures related to the following areas noted below.
Additional Specific Procedures may be requested by number after the inspector reviews
the procedure indexes.
1. ALARA Program
2. ALARA Planning
3. ALARA Reviews
4. ALARA Committee
5. Radiation Work Permit Preparation
F.
Please provide a list of NRC Regulatory Guides and NUREGs that you are currently
committed to relative to this program. Please include the revision and/or date for the
commitment and where this may be located in your current licensing basis documents.
G.
Please provide a summary list of corrective action documents (including corporate and
sub-tiered systems) written since the date of last inspection, related to the ALARA
program, including exceeding RWP Dose Estimates.
NOTE: These lists should include a description of the condition that provides
sufficient detail that the inspectors can ascertain the regulatory impact, the
significance level assigned to the condition, the status of the action (e.g., open,
working, closed, etc.) and the search criteria used. Please provide in document
formats which are sortable and searchable so that inspectors can quickly and
efficiently determine appropriate sampling and perform word searches, as
needed. (Excel spreadsheets are the preferred format.) If codes are used,
please provide a legend for each column where a code is used.
H.
List of work activities (RWPs) greater than 1 rem, since date of last inspection,
including the original dose estimates and actual doses accrued. (Excel format
preferred). Please provide all revisions/changes, as well as any related RWPs that
support the work activity.
SUNSI Review
ADAMS:
Non-Publicly Available
Non-Sensitive
Keyword:
By: NTaylor
Yes No
Publicly Available
Sensitive
OFFICE
SRI:DRP/B
RI:DRP/B
C:DRS/OB
C:DRS/IPAT
C:DRS/EB1
C:DRS/EB2
NAME
ASanchez
JChoate
GWerner
RKellar
VGaddy
JDrake
SIGNATURE
JMC
GEW
RLK
VGG
JFD
DATE
05/08/19
05/03/2019
04/30/2019
04/30/2019
04/30/2019
04/29/2019
OFFICE
C:DNMS/CHP
C:DRS/RCB
BC:DRP/B
NAME
GWarnick
NMakris
NTaylor
SIGNATURE
GXW
NJM
NHT
DATE
05/06/2019
4/29/19
5/8/19