IR 05000395/2017007: Difference between revisions

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{{Adams
{{Adams
| number = ML17335A032
| number = ML17150A455
| issue date = 11/30/2017
| issue date = 05/30/2017
| title = NRC Design Bases Assurance Inspection (Team) Report No. 05000395/2017007
| title = Notification of Virgil C. Summer Design Bases Assurance Inspection - NRC Inspection Report 05000395/2017007
| author name = Walker S
| author name = Bartley J
| author affiliation = NRC/RGN-II/DRS/EB1
| author affiliation = NRC/RGN-II/DRS/EB1
| addressee name = Lippard G
| addressee name = Lippard G
Line 11: Line 11:
| contact person =  
| contact person =  
| document report number = IR 2017007
| document report number = IR 2017007
| document type = Inspection Report, Letter
| document type = Inspection Plan, Letter
| page count = 24
| page count = 5
}}
}}


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=Text=
=Text=
{{#Wiki_filter:November 30, 2017
{{#Wiki_filter:May 30, 2017


==SUBJECT:==
==SUBJECT:==
VIRGIL C. SUMMER NUCLEAR STATION - NRC DESIGN BASES ASSURANCE INSPECTION (TEAM) REPORT NUMBER 05000395/2017007
NOTIFICATION OF VIRGIL C. SUMMER NUCLEAR STATION DESIGN BASES ASSURANCE INSPECTION - U.S. NUCLEAR REGULATORY COMMISSION INSPECTION REPORT 05000395/2017007


==Dear Mr. Lippard,==
==Dear Mr. Lippard:==
On October 20, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Virgil C. Summer Nuclear Station, Unit 1, and on November 20, 2017, the NRC inspectors discussed the results of this inspection with you and other members of your staff.
The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC)
Region II staff will conduct a Design Bases Assurance Inspection (DBAI) at your Virgil C.


The results of this inspection are documented in the enclosed report.
Summer Nuclear Station during the weeks of September 18 - 22, and October 2 - 6, 2017.


NRC inspectors documented three findings of very low safety significance (Green) in this report.
Mr. Marcus Riley, a reactor inspector from the NRCs Region II office, will lead the inspection team. The inspection will be conducted in accordance with Inspection Procedure 71111.21M, Design Bases Assurance Inspection (Teams), dated December 8, 2016 (ADAMS ML16238A320).


Three of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy. Further, inspectors documented a licensee-identified violation, which was determined to be of very low safety significance (Green) 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.
The inspection will evaluate the capability of components that have been modified and risk-significant/low-margin components to function as designed and to support proper system operation. The inspection will also include a review of selected operator actions, operating experience, and modifications.


If you contest the violations or significance of these NCVs, 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 II; the Director, Office of Enforcement; and the NRC resident inspector at the Virgil C. Summer Nuclear Station.
During a telephone conversation on May 23, 2017, with Mr. Bruce Thompson and Mr. Renard Perry, we confirmed arrangements for an information-gathering site visit and the two-week onsite inspection. The schedule is as follows:
* Information-gathering visit: Week of August 28 - September 1, 2017
*
Onsite weeks: Weeks of September 18 - 22, and October 2 - 6, 2017


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.
The purpose of the information-gathering visit is to meet with members of your staff to identify components that have been modified, risk-significant components and operator actions.


Sincerely,
Information and documentation needed to support the inspection will also be identified.
/RA/


Shakur A. Walker, Chief Engineering Branch 1 Division of Reactor Safety
Mr. George MacDonald, a Region II Senior Risk Analyst, will support Mr. Riley during the information-gathering visit to review probabilistic risk assessment data and identify components to be examined during the inspection. Additionally, during the onsite weeks, time will be needed on the plant-referenced simulator in order to facilitate the development of operator action-based scenarios. The enclosure lists documents that will be needed prior to the information-gathering visit.


Docket No.: 50-395 License No.: NPF-12
Please provide the referenced information to the Region II Office by Monday, August 14, 2017.


===Enclosure:===
Additional documents will be requested following the information-gathering visit. The inspectors will try to minimize your administrative burden by specifically identifying only those documents required for inspection preparation. The additional information will be needed in the Region II office by Friday, September 8, 2017, to support the inspection teams preparation week. During the information-gathering trip, Mr. Riley will also discuss the following inspection support administrative details: (1) availability of knowledgeable plant engineering and licensing personnel to serve as points of contact during the inspection; (2) method of tracking inspector requests during the inspection; (3) licensee computer access; (4) working space; (5)
Inspection Report 05000395/2017007, w/Attachment: Supplemental Information
arrangements for site access; and (6) other applicable information.


REGION II==
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.
Docket No.:


05000395
Thank you for your cooperation in this matter. If you have any questions, regarding the information requested or the inspection, please contact Mr. Riley at 404-997-4888 or contact me at 404-997-4607.


License No.:
Sincerely,
NPF-12
/RA/
 
Report No.:
 
05000395/2017007
 
Licensee:
 
South Carolina Electric & Gas (SCE&G) Company
 
Facility:
 
Virgil C. Summer Nuclear Station, Unit 1
 
Location:
 
Jenkinsville, SC 29065
 
Dates:


September 25 - October 20, 2017
Jonathan H. Bartley, Chief  
 
Inspectors:
 
M. Riley, Acting Senior Reactor Inspector (Lead)
 
C. Franklin, Reactor Inspector
 
N. Morgan, Reactor Inspector
 
D. Terry-Ward, Construction Inspector
 
W. Sherbin, Contractor
 
A. Della-Greca, Contractor
 
Approved by:
Shakur A. Walker, Chief  


Engineering Branch 1  
Engineering Branch 1  


Division of Reactor Safety
Division of Reactor Safety  
 
=SUMMARY=
Inspection Report (IR) 05000395/2017007; September 25 - October 20, 2017; Virgil C. Summer
 
Nuclear Station, Unit 1; Design Bases Assurance Inspection (Team).
 
The inspection activities described in this report were performed between September 25 through October 20, 2017, by a team of four U.S. Nuclear Regulatory Commission (NRC)inspectors and two contractors. The team identified three non-cited violations. The significance of inspection findings are indicated by their color (i.e., greater than Green, or Green, White,
Yellow, or Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, (SDP) dated April 29, 2015. Crosscutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated November 1, 2016. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,
Rev. 6.
 
A violation of very low safety significance that was identified by the licensee has been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. The violation and corrective action tracking numbers are listed in Section 4OA7 of this report.
 
===NRC-Identified and Self-Revealing Findings===
===Cornerstone: Mitigating Systems===
* Green: The NRC identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the emergency feedwater (EFW) pumps would be capable of taking suction from service water for an indefinite period of time as required by Updated Final Safety Analysis Report Section 10.4.9.2. The licensee entered this issue into their corrective action program (CAP) as condition report (CR) 17-05528 and performed an operability determination to verify the EFW pumps remained operable.
 
The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to evaluate worst-case design conditions resulted in a reasonable doubt that the EFW pumps could provide cooling water to the steam generators and perform their design basis function. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, and component (SSC), and the SSC maintained its operability. The team determined that no crosscutting aspect was applicable because the finding did not reflect current licensee performance (Section 1R21.2.b.1).
* Green: The NRC identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, involving two examples. Specifically, the licensee (1) failed to establish a testing program to assure the adequacy of the shutdown setpoint of the safety-related inverters, and (2) failed to establish a testing program to assure the adequacy of the time delay relay in the emergency feedwater/service water (EFW/SW) crosstie valve actuation circuitry. The licensee entered this issue into their CAP as CRs17-05534 and 17-05536, and performed an operability determination to verify that the safety-related components remained operable.
 
The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to establish a testing program for the low voltage DC setpoint of inverter XIT 5904 and for the time delay relay in the EFW/SW crosstie actuation circuitry could result in undetected degradation of the equipment to perform their intended safety functions. The team determined the finding to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability. The team determined that no cross-cutting aspect was applicable because the finding did not reflect current licensee performance (Section 1R21.2.b.2).
* Green: The NRC identified an NCV of 10 CFR Part 50, Appendix B, Criterion XVI,
Corrective Actions, for the licensees failure to identify that a deviation in equipment qualification of power shield relays in 480V switchgear XSW-1DB1 was a condition adverse to quality in their CAP. Specifically, the licensee failed to identify that Power Shield catalog #609903-T501N in purchase order NU-02SR750589 was not qualified to meet its original total integrated dose limit of 100,000 rads as stated in the Asea Brown Boveri 10 CFR Part 21 notification letter. The licensee entered this issue into their CAP as CR-17-05391 and performed an evaluation to determine there was reasonable assurance that the power shield relay in purchase order NU-02SR750589 could perform its intended safety function.
 
The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify that Power Shield catalog #609903-T501N in purchase order NU-02SR750589 was not qualified to the 1,350 rad TID specified in the equipment qualification database for zone AB-72 resulted in a reasonable doubt that the qualification requirements over the relays service life would be met. The team determined the finding to be of very low safety significance (Green)because the finding affected the design or qualification of a mitigating SSC and the SSC maintained its operability. The team determined that no crosscutting aspect was applicable because the finding did not reflect current licensee performance (Section 1R21.2.b.3).
 
=REPORT DETAILS=
 
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity {{a|1R21}}
 
==1R21 Design Bases Assurance Inspection (Team)==
{{IP sample|IP=IP 71111.21M}}
 
===.1 Inspection Sample Selection Process===
The team selected risk-significant samples and related operator actions for review using information contained in the licensees probabilistic risk assessment. In general, this included risk significant structures, systems, and components (SSCs) that had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1E-6. The sample included four components selected based on risk significance, one component associated with containment large early release frequency (LERF), five modifications to mitigation SSCs, and three operating experience (OE) items.
 
The team performed a margin assessment and a detailed review of the selected risk-significant components and associated operator actions to verify that the design bases had been correctly implemented and maintained. Where possible, this margin was determined by the review of the design basis and Updated Final Safety Analysis Report (UFSAR). This margin assessment also considered original design issues, margin reductions due to modifications, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for a detailed review. These reliability issues included items related to failed performance test results, significant corrective action, repeated maintenance, maintenance rule status, Inspection Manual Chapter 0326 conditions, NRC Resident Inspector input regarding problem equipment, system health reports, industry OE, and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, OE, and the available defense-in-depth margins. An overall summary of the reviews performed and the specific inspection findings identified is included in the following sections of the report.
 
===.2 Component Reviews===
====a. Inspection Scope====
Components Selected Based on Risk Significance
* 120VAC Inverter 5904 (XIT5904)
* ESF 480V Bus 1DA2 (XSW1DA2)
* Condensate Storage Tank (XTK0008)
* RHR Pump B (XPP0031B)
 
Components with LERF Implications
* Main Steam Isolation Valve C - (XVM02801C-MS)
 
Modifications to Mitigation SSCs
* ECR71221 - ESF Undervoltage Logic and Setting Calculations
* ECR50915 - Paralleling an Emergency Diesel Generator with the Alternate AC Line
* ECR50695E - Major Revision - EFW Sys Flow Control Enhancement
* ECR71571 - EDG Fuel Oil Transfer Suction Strainer D/P Pressure Switch Setpoints
* ECR71740 - ECCS HHSI Flow Balance Margin Enhancement
 
For the five components listed above, the team reviewed the plant technical specifications (TS), UFSAR, design bases documents, and drawings to establish an overall understanding of the design bases of the components. Design calculations and procedures were reviewed to verify that the design and licensing bases had been appropriately translated into these documents and that the most limiting parameters and equipment line-ups were used. Logic and wiring diagrams were also reviewed to verify that operation of electrical components conformed to design requirements. Test procedures and recent test results were reviewed against design bases documents to verify the adequacy of test methods and that acceptance criterion for tested parameters were supported by calculations or other engineering documents, and that individual tests and analyses served to validate component operation under accident conditions.
 
Maintenance procedures were reviewed to ensure components were appropriately included in the licensees preventive maintenance program, that components or sub-components were being replaced before the end of their intended service life, and that the licensee has appropriate controls in place for components that are beyond vendor recommended life. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action program documents were reviewed (as applicable) in order to verify that the performance capability of the component was not negatively impacted, and that potential degradation was monitored or prevented.
 
Maintenance Rule information was reviewed to verify that the component was properly scoped, and that appropriate preventive maintenance was being performed to justify current Maintenance Rule status. Component walkdowns and interviews were conducted to verify that the installed configurations would support their design and licensing bases functions under accident conditions, and had been maintained to be consistent with design assumptions.
 
For the five modifications listed above, the team reviewed design bases, licensing bases, and performance capability of components to ensure they have not been degraded through modifications. In addition, post-modification testing was reviewed to ensure operability was established by verifying unintended system interactions will not occur, SSC performance characteristic continue to meet the design bases, modification design assumptions are appropriate, and modification test acceptance criteria have been met. The team also verified design basis documentation was updated consistent with the design change, verified other design basis features were not adversely impacted, verified procedures and training plans affected by the modification were updated, and verified that affected test documentation was updated or initiated as required by applicable test programs. Walkdowns and interviews were conducted as necessary to verify that the modifications were adequately implemented. Documents reviewed are listed in the Attachment.


====b. Findings====
Docket Nos.: 50-395 License Nos.: NPF-12
===.1 Failure to Verify the Adequacy of Design for the Emergency Feedwater System When===
Supplied by Service Water


=====Introduction:=====
Enclosure:
The NRC identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the emergency feedwater (EFW) pumps would be capable of taking suction from service water (SW) for an indefinite period of time as required by UFSAR Section 10.4.9.2. Specifically, the licensee failed to verify that cooling water flow to the EFW pumps bearings and seal flush lines would not be blocked due to debris in the SW system, rendering the pumps incapable of taking suction from SW for an indefinite period of time.
Notification of Virgil C. Summer Nuclear Station, Design Bases Assurance Inspection (Teams)  


=====Description:=====
cc: Distribution via Listserv
The EFW system at Virgil C. Summer Nuclear Plant consists of two electric motor-driven EFW pumps; one turbine-driven EFW pump; the condensate storage tank (CST); and the necessary piping, valves, instrumentation, and controls. UFSAR Section 10.4.9.1 states that the design basis of the EFW system is to automatically deliver a minimum total flow of 380 gallons per minute (gpm) to at least two steam generators pressurized to 1211 psig and that the system is required to deliver sufficient feedwater to the steam generators for cooldown upon loss of the normal feedwater supply. UFSAR Section 10.4.9.1 also states that the EFW system must operate until the residual heat removal (RHR) system can be placed in operation at a reactor coolant pressure and temperature of approximately 400 psig and 350°F, respectively. UFSAR Section 10.4.9.2 states that the CST, which contains clean water, provides the preferred suction source for the EFW pumps and that the SW system is a safety related backup source that the EFW pumps can take suction from for an indefinite period of time.


The SW system takes suction from a safety class impoundment adjacent to Lake Monticello. The team noted that the orifices in the motor-driven EFW and turbine-driven EFW pumps bearing coolers and seal flush lines had cross-sectional openings sized between 0.094 and 0.125 inch, whereas the SW rotating screens openings had cross-sectional openings of approximately 0.25 inch square. The team also noted that there were no other screens or strainers in the system. The team concluded that the design of the EFW system and cross-sectional opening of the SW rotating screens could allow debris large enough to pass through the screens and clog the bearing and oil cooler lines in the EFW pumps, preventing the use of SW as an indefinite suction source for EFW as described in UFSAR 10.4.9.2. The team was particularly concerned during design basis tornado accidents. UFSAR Section 3.5 states that the CST is not protected from tornado missiles and that the loss of the CST during a tornado would not affect the capability to shut down the reactor and maintain it in a safe shutdown condition (hot standby) because SW would be available as an alternate source of makeup. The licensee provided historical data to the team, which indicated the lake water normally was virtually free of debris. However, the team was concerned that if a tornado passes through the site, which is when the SW lineup to EFW pumps suction is needed due to tornado missile striking the CST, there would be substantially more debris in the lake than when measured with clean lake water.
____ __________


The team noted calculation DC05220-048, Determination of the Minimum Required Volume in CST for EFW, Rev. 6, stated it would take six hours of EFW operation to bring the plant to hot shutdown, assuming the plant is maintained at hot standby for two hours and then cooled down to hot shutdown in four hours. To address the teams concern, the licensee entered the issue into their corrective action program (CAP) as condition report (CR) 17-05528 on October 19, 2017, and performed an operability determination to verify the EFW system remained operable. The licensee determined that the motor-driven EFW pumps could operate for about 11 hours without cooling water to the bearings or seals, and the turbine-driven EFW pump could operate for up to six hours without cooling water to the bearings or seals, thus ensuring EFW could operate long enough to bring the plant to hot shutdown when RHR would be available.
SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRS RII:DRS


=====Analysis:=====
SIGNATURE MAR1 JHB1
The team determined that the licensees failure to verify that cooling water flow to the EFW pumps bearings and seal flush lines would not be blocked due to debris to ensure the EFW system could take suction from SW for an indefinite period of time as described in UFSAR Section 10.4.9.2 was a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, and a performance deficiency (PD). The PD was determined to be more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to evaluate worst-case design conditions resulted in a reasonable doubt that the EFW pumps could provide cooling water to the steam generators and perform their design basis function.


The team used inspection manual chapter (IMC) 0609, Att. 4, Initial Characterization of Findings, issued December 7, 2016, for mitigating systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating structure, system, and component (SSC), and the SSC maintained its operability. Since the underlying cause of the issue occurred during original design, the team determined that no crosscutting aspect was applicable because the finding was not indicative of current licensee performance.
NAME M. RILEY J. BARTLEY


=====Enforcement:=====
DATE 5/ 30/2017 5/30/2017
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculation methods, or by the performance of a suitable testing program.


UFSAR Section 10.4.9.2 stated that the EFW system could take suction from the SW system for an indefinite period of time. Contrary to the above, since original design and construction, the licensee failed to verify the EFW pumps could take suction from SW for an indefinite period of time, due to potential blockage of cooling water flow to the EFW pumps bearing and seal flush lines when aligned to SW. Loss of cooling to EFW pumps bearing and seal flush lines would adversely affect the ability of the EFW pumps to perform their design basis function. The licensee entered this issue into their CAP as CR 17-05528 and performed an operability determination to verify the EFW pumps remained operable. This violation is being treated as an NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy: (NCV 05000395/2017007-01, Failure to Verify the Adequacy of Design for the EFW system when supplied by Service Water).
E-MAIL COPY?
YES NO YES NO


===.2 Failure to Establish a Testing Program for Inverter XIT5904 and for the Time Delay===
Enclosure INFORMATION REQUEST FOR VIRGIL C. SUMMER NUCLEAR STATION DESIGN BASES ASSURANCE INSPECTION (TEAMS)
Relay in the Emergency Feedwater/Service Water Crosstie Valve Actuation Circuitry


=====Introduction:=====
Please provide the information electronically in.pdf files, Excel, or other searchable format on CDROM (or FTP site, SharePoint, etc.). The CDROM (or website) should be indexed and hyperlinked to facilitate ease of use. The requested items below, identified with an asterisk (*),
The NRC identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, involving two examples. Specifically, the licensee
should have a date range from June 1, 2014, until present.
: (1) failed to establish a testing program to assure the adequacy of the shutdown setpoint of the safety-related inverters, and
: (2) failed to establish a testing program to assure the adequacy of the time delay relay in the EFW/SW crosstie valve actuation circuitry.


=====Description:=====
1.
UFSAR Section 8.1 states that the licensee is committed to IEEE 308-1971. Section 6.3 of IEEE 308-1971 states that periodic tests shall be performed to (1)detect the deterioration of the system toward an unacceptable condition and (2)demonstrate that standby power equipment and other components that are not exercised during normal operation of the station are operable. The team identified the following two examples of the licensees failure to test safety-related equipment in accordance with IEEE 308-1971.
* List and brief description of permanent and field work completed plant modifications including permanent plant changes, design changes, set point changes, procedure changes, equivalency evaluations, suitability analyses, calculations, and commercial grade dedications. Include an index of systems (system numbers/designators and corresponding names), the safety classification for each modification, and type of modification.


Example 1 - Failure to Test Inverter XIT5904 Shutdown Setpoint:
2.


Calculation DC08320-010, Class 1E 125 Volt DC System Voltages and Voltage Drop, Rev. 18, evaluated the minimum voltage available at DC components under worst-case conditions and, in particular, at the end of the postulated four-hour mission time of the batteries. For safety-related 120 VAC Inverter XIT5904, the calculation determined that, with 58 battery cells in service, the minimum DC voltage at the inverter terminals would be 104.62 VDC. The calculation also stated that the minimum recommended operating voltage is 104 VDC and that The inverter low dc voltage sensor will turn off the inverter at 104 VDC. Because of the small margin between the calculated minimum DC voltage and the stated shutdown setpoint of the inverters, the team asked the licensee about the inverters low voltage DC trip setting, and the testing being done to verify that the inverters would not trip before the four hour mission time because of the trip setting drifting high. The team identified that the low voltage DC trip set point was not in the testing program and was not being tested periodically. Although the team recognized the inverters low voltage trip setting was set at the factory, the team noted that the setting was adjustable and that installation/maintenance activities, component aging, and/or environmental conditions could affect the factory setting and result in an inadvertent loss of the inverter during design basis conditions. Also as a result of the teams questions, the licensee received communication from the vendor indicating that the trip setting was actually 103 VDC.
From your most recent probabilistic safety analysis (PSA) excluding external events and fires:


On October 19, 2017, the licensee entered this issue into their CAP as CR-17-05534 and verified there was reasonable assurance that the inverter remained operable. The licensee determined that both safety-related batteries, XBA1A and XBA1B, passed their most recent capacity tests with a final voltage at four hours of 111.8VDC and 114.0VDC, respectively, well above the cutout setpoint of 103VDC.
a. Two risk rankings of components from your site-specific PSA: one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance


Example 2-Failure to Test Time Delay Relay in the Emergency Feedwater/Service Water Crosstie Valve Actuation Circuitry:
b. A list of the top 500 cut-sets


The automatic actuation circuitry of the EFW/SW crosstie valve consists of a CST level switch, a time delay relay, and a valve actuation circuit. The purpose of the time delay is to prevent spurious trips of the EFW pumps due to pressure spikes when starting and stopping the EFW pumps. Specifically, when the CST level switch actuates because of a loss of CST inventory, the signal goes to the time delay relay. When the relay times-out after 4.5 seconds, the crosstie valve gets a signal to open. The team noted that the time delay relay was set in 1982 and that the time delay function of the relay had not been tested since its installation date. The failure to perform periodic testing of the time delay relay could result in an undetected drift of the time delay relay, which could result in the loss of the EFW pump. On October 19, 2017, the licensee entered this issue into their CAP as CR 17-05536 and determined there was reasonable assurance that the relay could perform its intended safety function.
c. A list of the top 500 LERF contributors


=====Analysis:=====
3.
The team determined the licensees failure to establish a testing program to periodically test
: (1) the low voltage DC setpoint of inverter XIT5904 and
: (2) the time delay relay for the EFW/SW crosstie valve actuation circuitry in accordance with IEEE 308-1971, was a violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, and a PD. The PD was determined to be more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the inverters to respond to initiating events to prevent undesirable consequences.


Specifically, failing to establish a testing program for the low voltage DC setpoint of inverter XIT 5904 and for the time delay relay in the EFW/SW crosstie actuation circuitry could result in undetected degradation of the equipment to perform their intended safety functions.
From your most recent PSA including external events and fires:


The team used IMC 0609, Att. 4, Initial Characterization of Findings, issued October 7, 2016, for mitigating systems, and IMC 0609, App. A, The SDP for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability. Since the underlying cause of the issues occurred since original plant installation, this finding was not assigned a crosscutting aspect because the issue did not reflect current licensee performance.
a. Two risk rankings of components from your site-specific PSA: one sorted by RAW, and the other sorted by Birnbaum Importance


=====Enforcement:=====
b. A list of the top 500 cut-sets
Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires in part, that A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. UFSAR Section 8.1 stated that the licensee was committed to IEEE 308-1971. Section 6.3 of IEEE 308-1971 required periodic tests be performed to
: (1) detect the deterioration of the system toward an unacceptable condition and
: (2) demonstrate that standby power equipment and other components that are not exercised during normal operation of the station are operable. Contrary to the above, since original plant installation, the licensee failed to establish a testing program to detect the deterioration of the shutdown setpoint of inverter XIT 5904 and the time delay relay in the EFW/SW crosstie valve actuation circuitry. Failing to establish a testing program for the low voltage DC setpoint of inverter XIT 5904 and for the time delay relay in the EFW/SW crosstie actuation circuitry could result in undetected degradation of the equipment to perform their intended safety functions. The licensee entered this issue into their CAP as CRs17-05534 and 17-05536, and performed an operability determination to verify that the safety-related components remained operable. This violation is being treated as an NCV consistent with section 2.3.2.a of the NRC enforcement policy (NCV 05000395/2017007-02, Failure to Establish a Testing Program for Inverter XIT5904 and Time Delay Relay in the EFW/SW Crosstie Valve Actuation Circuitry).


===.3 Failure to Identify a Condition Adverse to Quality for Power Shield Catalog #609903-===
4.
T501N in Purchase Order NU-02SR750589


=====Introduction:=====
Risk ranking of operator actions from your site-specific PSA sorted by RAW and human reliability worksheets for these items
The NRC identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to identify that a deviation in equipment qualification of power shield relays in 480V switchgear XSW-1DB1 was a condition adverse to quality in their corrective action program. Specifically, the licensee failed to identify that Power Shield catalog #609903-T501N in purchase order NU-02SR750589 was not qualified to meet the total integrated dose (TID) limit of 100,000 rads as stated in the Asea Brown Boveri (ABB) 10 CFR Part 21 notification letter.


=====Description:=====
5.
On September 27, 2013, ABB sent a 10 CFR Part 21 notification of deviation letter to the licensee stating the certificates of conformance for solid state circuit shield relays and Power Shield (K-Line) relays containing complementary metal oxide semiconductor technology incorrectly identified the relays were qualified to 100,000 rads and that the solid state relays and power shields were actually qualified to 1,000 rads. The notification letter identified 27 Power Shield catalog numbers that were affected. In 2013, the licensee entered this issue into their CAP as CRs 13-04047 and 13-03936 to address the purchase orders that contained the affected catalog numbers.


CR 13-03936 stated, Upon further review, the Part 21 letter noted that this issue arose since the reference documentation was never updated after a prior 1994 notice which specified that the affected K-line power shields are only qualified to 1000 rads rather than the 100000 rads.VCS response in 1994 was that the maximum any relays would see is less than or equal to 880 Rads, which is less than the 1,000 rad qualification. However, upon review of the relay's location, the power shield relays used in XSW-1DB1 will see a total integrated dose (40-year life) of 880 rad with a potential loss of coolant accident (LOCA) dose of 470 rad. This total dose of 1,350 rad is greater than the qualified 1000 rad limit. The location in question is in EQ zone AB-72 and shows a dose rate of 2.5 mR/hr (880 TID). However, discussion with health physics revealed that the room in question has been recorded at less than 0.5 mR/hr. The licensee stated that this equipment qualification (EQ) zone was mild per UFSAR Section 3.11.
List of time-critical operator actions with a brief description of each action


The licensee submitted purchase order NU-02SR750589 for review by the team, and the team identified that this purchase order contained Power Shield catalog #609903-T501-N, which was identified as one of the affected catalog numbers in the ABB 10 CFR Part 21 notification of deviation letter. The team noted that this purchase order was omitted from CRs 13-04047 and 13-03936, and that the licensee had not evaluated if this Power Shield relay was capable of performing its intended function. The relay was also located in safety-related 480VAC switchgear XSW-1DB1. Exceeding the TID limits of the safety-related power shield relay would not ensure that the protection system equipment in switchgear XSW-1DB1 could meet its performance requirements on a continuing basis during normal and design basis events as a result of the deviation.
6.
* List of components with low-design margins (i.e., pumps closest to the design limit for flow or pressure, diesel generator close to design-required output, heat exchangers close to rated design heat removal, and motor-operated valve risk-margin rankings, etc.) and associated evaluations or calculations


To address the new issue identified by the team, the licensee entered this issue into their CAP on October 12, 2017, as CR-17-05391 and performed an evaluation to determine if there was reasonable assurance that the power shield relay in purchase order NU-02SR750589 could perform their intended safety function. As part of the evaluation, the licensee reviewed survey maps of EQ zone AB-72 dating back to 2003, which showed that the dose rates of the room was less than 0.5 mR/hr. The licensee determined that upon utilizing a conservative number of 0.5 mR/hr, the TID for that area would be less than 180 rad. With the added 470 rad from a LOCA event, the total dose the relay would see would be approximately 650 rad, which was within the 1,000 rad qualified life of the relay.
7.
* List and brief description of Root Cause Evaluations performed 8.
* List and brief description of common-cause component failures that have occurred


=====Analysis:=====
9.
The licensees failure to identify that a deviation in the equipment qualification of Power Shield catalog #609903-T501N in purchase order NU-02SR750589 was a condition adverse to quality in CRs 13-04047 and 13-03936, was a violation of Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and a PD. The PD was determined to be more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify that Power Shield catalog #609903-T501N in purchase order NU-02SR750589 was not qualified to the 1,350 rad TID specified in the equipment qualification database for zone AB-72 resulted in a reasonable doubt that the qualification requirements would be met over the relays service life.


The team used IMC 0609, Att.4, Initial Characterization of Findings, issued October 7, 2016, for mitigating systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding affected the design or qualification of a mitigating SSC and the SSC maintained its operability. Since the underlying cause of the issue occurred on September 27, 2013, the team determined that no crosscutting aspect was applicable because the finding was not indicative of current licensee performance.
List and brief description of equipment currently in degraded or nonconforming status as described in NRC Inspection Manual Chapter 0326, issued December 3, 2015 10. *List and brief description of Operability Determinations and Functionality Assessments


=====Enforcement:=====
11. *List and reason for equipment that has been classified in maintenance rule (a)(1) status
Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above, since September 27, 2013, the licensee failed to establish measures to assure that a deviation related to equipment qualification of Power Shield catalog
#609903-T501N in purchase order NU-02SR750589 was promptly identified. Exceeding the TID limits of the safety-related relay in 480VAC Switchgear XSW-1DB1 would not ensure that the protection system equipment could meet its performance requirements on a continuing basis. The licensee entered this issue into their CAP as CR-17-05391 and performed an operability determination to verify that the relay remained operable.


This violation is being treated as an NCV consistent with section 2.3.2.a of the Enforcement Policy (NCV 05000395/2017007-03, Failure to Identify a CAQ for Power Shield Catalog #609903-T501N in Purchase Order NU-02SR750589).
12. *List of equipment on the sites Station Equipment Reliability Issues List, including a description of the reason(s) why each component is on that list, and summaries (if available) of your plans to address the issue(s) along with dates added or removed from the issues list


===.3 Operating Experience===
13. List of current operator work arounds/burdens
====a. Inspection Scope====
The team reviewed three operating experience issues for applicability at the Virgil C.


Summer Nuclear Plant. The team performed an independent review for these issues and, where applicable, assessed the licensees evaluation and disposition of each item.
14. Copy of Updated Final Safety Analysis Report


The issues that received a detailed review by the team included:
15. Copy of Technical Specification(s)  
* Westinghouse NSAL 09-8: Presence of Vapor in Emergency Core Cooling System/Residual Heat Removal System in Modes 3/4 Loss-of-Coolant Accident Conditions
* NRC Information Notice 15-13: Main Steam Isolation Valve Failure Events
* NRC Information Notice 91-13: Inadequate Testing of Emergency Diesel Generators (EDGs)


====b. Findings====
16. Copy of Technical Specifications Bases
None


==OTHER ACTIVITIES==
17. Copy of Technical Requirements Manual(s)
{{a|4OA6}}


==4OA6 Meetings, Including Exit==
18. Copy of the Quality Assurance Program Manual
On October 20, 2017, the team presented the inspection results to Mr. Don Shue and other members of the licensees staff. On November 20, 2017, a re-exit meeting was conducted via teleconference to present the final inspection results to Mr. George Lippard III and other members of the licensees staff. Proprietary information that was reviewed during the inspection was returned to the licensee or destroyed in accordance with prescribed controls.


{{a|4OA7}}
19. Copy of Corrective Action Program Procedure(s)


==4OA7 Licensee-Identified Violations==
20. Copy of Operability Determination Procedure(s)  
The following licensee-identified violation of NRC requirements was determined to be of very low safety significance (Green) and met the NRC Enforcement Policy criteria for being dispositioned as a non-cited violation.
* Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.


Contrary to the above, since 2010, the licensee failed to evaluate the loading of the emergency diesel generators at the maximum voltage and frequency allowed by TS 3/4.8.1 in Calculation DC08360-006, Diesel Generator 1A and 1B Loading, Rev.
21. List of motor operated valves and air operated valves in the valve program, and their associated design margin and risk ranking


12, and to evaluate battery terminal voltage at the maximum battery cell-to-cell resistance allowed by TS 3/4.8.2 in Calculation DC08320-010, Class 1E 125 Volt DC System Voltages and Voltage Drop, Rev. 18. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating SSC, and the SSC maintained its operability. The licensee entered these issues into their CAP as CRs 10-02395 and 10-02033.
22. Primary AC and DC calculations for safety-related buses


ATTACHMENT:
23. One-line diagram of electrical plant (Electronic only)


=SUPPLEMENTAL INFORMATION=
24. Index and legend for electrical plant one-line diagrams


==KEY POINTS OF CONTACT==
25. Piping and instrumentation diagrams (P&IDs) for safety-related systems (Electronic)  
===Licensee Personnel===
:
: [[contact::D. Shue]], Nuclear Operations Manager
: [[contact::R. Ray]], Maintenance Manager
: [[contact::T. Ledbetter]], Planning and Scheduling Manager
: [[contact::S. Zarandi]], Nuclear Support Services General Manager
: [[contact::B. Thompson]], Nuclear Licensing Manager
: [[contact::W. Stuart]], Engineering Services General Manager
: [[contact::R. Haselden]], Organization Effectiveness General Manager
: [[contact::L. Harris]], Quality Systems Manager
: [[contact::G. Douglass]], Nuclear Protection Services Manager
: [[contact::G. Williams]], Plant Support Engineering Program Supervisor
: [[contact::M. Carr]], Engineering
: [[contact::C. Calvert]], Design Engineering Manager
: [[contact::T. Bussey]], Nuclear Fuels & Analysis Supervisor)  
: [[contact::R. Perry]], Nuclear Licensing
: [[contact::J. Archie]], Chief Nuclear Officer
: [[contact::M. Verrilli]], Materials and Procurement
: [[contact::G. Lindamood]], Santee Cooper
: [[contact::C. Boozer]], Engineering Supervisor
: [[contact::W. Martin]], Nuclear Licensing
: [[contact::W. Kearney]], Operations
: [[contact::D. Edwards]], Operations
: [[contact::J. Ward]], Design Engineering
: [[contact::D. Dobson]], Design Engineering


===NRC Personnel===
26. Index and legend for P&IDs
:
: [[contact::J. Reece]], Senior Resident Inspector
: [[contact::E. Hilton]], Resident Inspector


==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
27. Index (procedure number, title, and current revision) of station Emergency Operating Procedures, Abnormal Operating Procedures, and Annunciator Response Procedures
Opened &


===Closed===
28. Copies of corrective action documents generated from previous CDBI
: 05000395/2017007-01 NCV Failure to Verify the Adequacy of Design for the EFW system when Supplied by SW (Section 1R21.b.1)
: 05000395/2017007-02 NCV Failure to Establish a Testing Program for Inverter XIT5904 and Time Delay Relay in the EFW/SW Crosstie Valve Actuation Circuitry (Section 1R21.b.2)
: 05000395/2017007-03 NCV Failure to Identify a CAQ for Power Shield Catalog #609903-T501N (Section 1R21.b.3)


==LIST OF DOCUMENTS REVIEWED==
29. Copy of any self-assessments performed, and corrective action documents generated, in preparation for current DBAI


30. Contact information for a person to discuss PSA information prior to and during the information-gathering trip (Name, title, phone number, and e-mail address)
}}
}}

Latest revision as of 20:25, 8 January 2025

Notification of Virgil C. Summer Design Bases Assurance Inspection - NRC Inspection Report 05000395/2017007
ML17150A455
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 05/30/2017
From: Bartley J
NRC/RGN-II/DRS/EB1
To: Lippard G
South Carolina Electric & Gas Co
References
IR 2017007
Download: ML17150A455 (5)
v • d • e


Text

May 30, 2017

SUBJECT:

NOTIFICATION OF VIRGIL C. SUMMER NUCLEAR STATION DESIGN BASES ASSURANCE INSPECTION - U.S. NUCLEAR REGULATORY COMMISSION INSPECTION REPORT 05000395/2017007

Dear Mr. Lippard:

The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC)

Region II staff will conduct a Design Bases Assurance Inspection (DBAI) at your Virgil C.

Summer Nuclear Station during the weeks of September 18 - 22, and October 2 - 6, 2017.

Mr. Marcus Riley, a reactor inspector from the NRCs Region II office, will lead the inspection team. The inspection will be conducted in accordance with Inspection Procedure 71111.21M, Design Bases Assurance Inspection (Teams), dated December 8, 2016 (ADAMS ML16238A320).

The inspection will evaluate the capability of components that have been modified and risk-significant/low-margin components to function as designed and to support proper system operation. The inspection will also include a review of selected operator actions, operating experience, and modifications.

During a telephone conversation on May 23, 2017, with Mr. Bruce Thompson and Mr. Renard Perry, we confirmed arrangements for an information-gathering site visit and the two-week onsite inspection. The schedule is as follows:

  • Information-gathering visit: Week of August 28 - September 1, 2017

Onsite weeks: Weeks of September 18 - 22, and October 2 - 6, 2017

The purpose of the information-gathering visit is to meet with members of your staff to identify components that have been modified, risk-significant components and operator actions.

Information and documentation needed to support the inspection will also be identified.

Mr. George MacDonald, a Region II Senior Risk Analyst, will support Mr. Riley during the information-gathering visit to review probabilistic risk assessment data and identify components to be examined during the inspection. Additionally, during the onsite weeks, time will be needed on the plant-referenced simulator in order to facilitate the development of operator action-based scenarios. The enclosure lists documents that will be needed prior to the information-gathering visit.

Please provide the referenced information to the Region II Office by Monday, August 14, 2017.

Additional documents will be requested following the information-gathering visit. The inspectors will try to minimize your administrative burden by specifically identifying only those documents required for inspection preparation. The additional information will be needed in the Region II office by Friday, September 8, 2017, to support the inspection teams preparation week. During the information-gathering trip, Mr. Riley will also discuss the following inspection support administrative details: (1) availability of knowledgeable plant engineering and licensing personnel to serve as points of contact during the inspection; (2) method of tracking inspector requests during the inspection; (3) licensee computer access; (4) working space; (5)

arrangements for site access; and (6) other applicable information.

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.

Thank you for your cooperation in this matter. If you have any questions, regarding the information requested or the inspection, please contact Mr. Riley at 404-997-4888 or contact me at 404-997-4607.

Sincerely,

/RA/

Jonathan H. Bartley, Chief

Engineering Branch 1

Division of Reactor Safety

Docket Nos.: 50-395 License Nos.: NPF-12

Enclosure:

Notification of Virgil C. Summer Nuclear Station, Design Bases Assurance Inspection (Teams)

cc: Distribution via Listserv

____ __________

SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRS RII:DRS

SIGNATURE MAR1 JHB1

NAME M. RILEY J. BARTLEY

DATE 5/ 30/2017 5/30/2017

E-MAIL COPY?

YES NO YES NO

Enclosure INFORMATION REQUEST FOR VIRGIL C. SUMMER NUCLEAR STATION DESIGN BASES ASSURANCE INSPECTION (TEAMS)

Please provide the information electronically in.pdf files, Excel, or other searchable format on CDROM (or FTP site, SharePoint, etc.). The CDROM (or website) should be indexed and hyperlinked to facilitate ease of use. The requested items below, identified with an asterisk (*),

should have a date range from June 1, 2014, until present.

1.

  • List and brief description of permanent and field work completed plant modifications including permanent plant changes, design changes, set point changes, procedure changes, equivalency evaluations, suitability analyses, calculations, and commercial grade dedications. Include an index of systems (system numbers/designators and corresponding names), the safety classification for each modification, and type of modification.

2.

From your most recent probabilistic safety analysis (PSA) excluding external events and fires:

a. Two risk rankings of components from your site-specific PSA: one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance

b. A list of the top 500 cut-sets

c. A list of the top 500 LERF contributors

3.

From your most recent PSA including external events and fires:

a. Two risk rankings of components from your site-specific PSA: one sorted by RAW, and the other sorted by Birnbaum Importance

b. A list of the top 500 cut-sets

4.

Risk ranking of operator actions from your site-specific PSA sorted by RAW and human reliability worksheets for these items

5.

List of time-critical operator actions with a brief description of each action

6.

  • List of components with low-design margins (i.e., pumps closest to the design limit for flow or pressure, diesel generator close to design-required output, heat exchangers close to rated design heat removal, and motor-operated valve risk-margin rankings, etc.) and associated evaluations or calculations

7.

  • List and brief description of Root Cause Evaluations performed 8.
  • List and brief description of common-cause component failures that have occurred

9.

List and brief description of equipment currently in degraded or nonconforming status as described in NRC Inspection Manual Chapter 0326, issued December 3, 2015 10. *List and brief description of Operability Determinations and Functionality Assessments

11. *List and reason for equipment that has been classified in maintenance rule (a)(1) status

12. *List of equipment on the sites Station Equipment Reliability Issues List, including a description of the reason(s) why each component is on that list, and summaries (if available) of your plans to address the issue(s) along with dates added or removed from the issues list

13. List of current operator work arounds/burdens

14. Copy of Updated Final Safety Analysis Report 15. Copy of Technical Specification(s)

16. Copy of Technical Specifications Bases

17. Copy of Technical Requirements Manual(s)

18. Copy of the Quality Assurance Program Manual

19. Copy of Corrective Action Program Procedure(s)

20. Copy of Operability Determination Procedure(s)

21. List of motor operated valves and air operated valves in the valve program, and their associated design margin and risk ranking

22. Primary AC and DC calculations for safety-related buses

23. One-line diagram of electrical plant (Electronic only)

24. Index and legend for electrical plant one-line diagrams

25. Piping and instrumentation diagrams (P&IDs) for safety-related systems (Electronic)

26. Index and legend for P&IDs

27. Index (procedure number, title, and current revision) of station Emergency Operating Procedures, Abnormal Operating Procedures, and Annunciator Response Procedures

28. Copies of corrective action documents generated from previous CDBI

29. Copy of any self-assessments performed, and corrective action documents generated, in preparation for current DBAI

30. Contact information for a person to discuss PSA information prior to and during the information-gathering trip (Name, title, phone number, and e-mail address)

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