IR 05000387/2015007
| ML15288A426 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/15/2015 |
| From: | Paul Krohn Engineering Region 1 Branch 2 |
| To: | Rausch T Susquehanna |
| References | |
| IR 2015007 | |
| Download: ML15288A426 (22) | |
Text
ber 15, 2015
SUBJECT:
SUSQUEHANNA STEAM ELECTRIC STATION - EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000387/2015007 AND 05000388/2015007
Dear Mr. Rausch:
On September 18, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Susquehanna Steam Electric Station, Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on September 18, 2015, with Mr. Jon Franke, Site Vice President, and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
In conducting the inspection, the team reviewed selected procedures, calculations and records, observed activities, and interviewed station personnel.
Based on the results of this inspection, no findings were identified.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system, Agencywide Documents Access and Management System (ADAMS).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-387 and 50-388 License Nos. NPF-14 and NPF-22
ML15288A426 Non-Sensitive Publicly Available SUNSI Review Sensitive Non-Publicly Available OFFICE RI/DRS RI/DRP RI/DRS NAME Spindale via email DSchroeder/SB PKrohn DATE 10/8/15 10/14/15 10/15/15
Enclosure:
Inspection Report 05000387/2015007 and 05000388/2015007 w/Attachment: Supplemental Information
REGION I==
Docket Nos.: 50-387; 50-388 License Nos.: NPF-14; NPF-22 Report No.: 05000387/2015007 and 05000388/2015007 Licensee: Susquehanna Nuclear, LLC (Susquehanna)
Facility: Susquehanna Steam Electric Station, Units 1 and 2 Location: Berwick, Pennsylvania Inspection Period: August 31 through September 18, 2015 Inspectors: S. Pindale, Senior Reactor Inspector, Team Leader J. Patel, Reactor Inspector A. Turilin, Project Engineer Approved By: Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety i Enclosure
SUMMARY OF FINDINGS
IR 05000387/2015007 and 05000388/2015007; 08/31/15 - 09/18/15; Susquehanna Steam
Electric Station, Units 1 and 2; Engineering Specialist Plant Modifications Inspection.
This report covers a 2-week inspection of the evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by two region based engineering inspectors and one project engineer. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,
Revision 4, dated December 2006.
No findings were identified.
ii
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications
.1 Evaluations of Changes, Tests, or Experiments (26 samples)
a. Inspection Scope
The team reviewed five safety evaluations to evaluate whether the changes to the facility or procedures, as described in the Updated Final Safety Analysis Report (UFSAR), had been reviewed and documented in accordance with Title 10 of the Code of Federal Regulations (CFR) Part 50.59 requirements. In addition, the team evaluated whether Susquehanna Nuclear, LLC (Susquehanna) had been required to obtain U.S. Nuclear Regulatory Commission (NRC) approval prior to implementing the changes. The team interviewed plant staff and reviewed supporting information including calculations, analyses, design change documentation, procedures, the UFSAR, the Technical Specifications (TS), and plant drawings to assess the adequacy of the safety evaluations. The team compared the safety evaluations and supporting documents to the guidance and methods provided in Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Evaluations, as endorsed by NRC Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, to determine the adequacy of the safety evaluations.
The team also reviewed a sample of twenty-one 10 CFR 50.59 screenings for which Susquehanna had concluded that a safety evaluation was not required to be performed.
These reviews were performed to assess whether Susquehanna's threshold for performing safety evaluations was consistent with 10 CFR 50.59. The sample included design changes, calculations, and procedure changes.
The team reviewed the safety evaluations that Susquehanna had performed and approved during the time period covered by this inspection (i.e., since the last plant modifications inspection) not previously reviewed by NRC inspectors. The screenings and applicability determinations were selected based on the safety significance, risk significance, and complexity of the change to the facility.
In addition, the team compared Susquehannas administrative procedures used to control the screening, preparation, review, and approval of safety evaluations to the guidance in NEI 96-07 to evaluate whether those procedures adequately implemented the requirements of 10 CFR 50.59. The reviewed safety evaluations and screenings are listed in the Attachment.
b. Findings
No findings were identified.
.2 Permanent Plant Modifications (13 samples)
.2.1 Appendix R Multiple Spurious Operations #5K Unit 2 Residual Heat Removal F007A and
F024A Valves
a. Inspection Scope
The team reviewed modification EC-1305833 that evaluated and modified the control circuits for residual heat removal (RHR) valves F007A and F024A to resolve multiple spurious operations (MSO) caused by fire-induced circuit failures during a postulated fire in the main control room (MCR). NRC Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire, discussed a susceptibility to fire-induced circuit failures of motor-operated valves (MOVs), where a hot-short (i.e., an electrical short circuit between control wiring and power source) in the MCR could bypass the torque and limit switches such that the motor actuator would not de-energize and could cause valve damage. This modification reconfigured the control circuit wiring of the two RHR valves so that torque and limit switches are electrically located below the MCR control switches (i.e., located electrically between the control room and the motor control center). In this configuration, the torque and limit switches would not be bypassed during a postulated fault describe above.
The team assessed whether the modification was consistent with requirements and in the design and licensing bases. The team conducted interviews with the responsible design engineer and performed a walkdown of the motor control center affected by this modification. The team reviewed the modification to determine whether the design function of these valves had been adversely affected by the change. The team also reviewed post-modification testing of the equipment to verify proper operation of the valves; and to verify the valves met the stroke time requirements. The team also evaluated whether affected plant design drawings were properly updated. Finally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the
.
b. Findings
No findings were identified.
.2.2 Residual Heat Removal Pump Motor Replacement
a. Inspection Scope
The team reviewed modification EC-1315883 that was implemented to replace the motor for Unit 1 RHR pump 1P202B. The motor was replaced because the original motor was approaching its qualified end of life. The replacement motor was a spare motor previously installed in Unit 2 RHR pump 2P202B and had been reconditioned by an offsite vendor for stator rewind and rotor replacement.
The team reviewed the modification to verify that the design and licensing bases of the RHR system had not been degraded by the motor change. The team conducted interviews with the engineering staff and reviewed the engineering change to ensure that the replacement motor had similar electrical characteristics to the original motor. In particular, the team verified that the impact of the change was addressed for power consumption requirements, cable protection requirements, voltage drop requirements, overload conditions, and short-circuit protection requirements. The team also verified that the affected plant design drawings and calculations were properly updated; and reviewed post-modification testing of the motor to verify proper operation. Finally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.3 Replacement of Drywell Cooler Motor 1VM418A
a. Inspection Scope
The team reviewed EC-1473145 that replaced a drywell cooler motor 1VM418A. The replacement motor was equivalent to the existing motor, which became obsolete. The replacement motor had the same electrical characteristics, but was heavier than the original motor. The team evaluated the increased weight relative to its seismic qualification and the support structure calculation.
The team reviewed the modification to verify that the design and licensing bases had not been degraded by the modification. The team interviewed design engineering staff; and reviewed the seismic qualification report and the revised structure support calculation to evaluate whether the new motor would perform consistent with the previously installed motor design. The team reviewed drawings and post-modification testing to ensure that the motor was properly installed and tested. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.4 Correcting of the Degraded Grid Protection Scheme for 0B565
a. Inspection Scope
The team reviewed modification EC-738302 that was implemented to install a new under-voltage relay (27-3) across the A and C phases of the potential transformer secondary associated with the E emergency diesel generator (EDG) 480 Vac (Volts, alternating current) motor control center 0B565 bus. The new under-voltage relay design is to detect a single-phase fault condition caused by a blown C phase primary fuse in either of the 13.8 kV transformers, 0X555 or 0X556. The team evaluated whether the change was consistent with industry standards and whether it incorporated learning from related industry operating experiences.
The team assessed whether the modification was consistent with requirements in the design and licensing bases. The team reviewed the under-voltage protection scheme on the E EDG and evaluated whether the addition of the new relay between the A and C phases degraded the current protection scheme. The team assessed whether there was adequate electrical separation between the new Class 1E relay and non-Class 1E circuit in the cabinet where this relay was installed; and whether the seismic qualification of the equipment was affected by this modification. The team reviewed the setpoint calculation to evaluate whether the correct setting was established for the new under-voltage relay.
Interviews were conducted with the design engineering staff and a walkdown was conducted of affected components. The team reviewed the implementing work order to ensure that it was consistent with the modification package; and to ensure proper post-modification testing was performed. The team verified that the affected plant design documents and calculation were properly updated. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.5 Machining of the E Diesel Generator 3R Cylinder Water Inlet Flange
a. Inspection Scope
The team reviewed modification EC-1726799, which machined the 3R cylinder flange for the jacket water inlet header on the E EDG to re-establish the flange surface and replaced the original gasket with a thicker (1/8 inch) gasket. The flange surface was not square and, as a result, created a leak at the jacket water inlet to the EDG.
The team reviewed the modification to evaluate whether the design and licensing bases and performance capability of the EDG had been degraded by the modification. The team assessed Susquehannas technical evaluations and design details, including installation specifications, to evaluate whether the EDG would function in accordance with the technical evaluation assumptions, and with design and licensing requirements. Drawings were reviewed to evaluate whether they were properly updated. The team walked down the E EDG with the system engineer to verify the modification was performed as described in the modification package. Additionally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.6 Replace Battery 0D595 with a KCR-21 Battery
a. Inspection Scope
The team reviewed EC-1719118 that modified the E EDG Class 1E 125 Vdc (Volts, direct current) battery 0D595. Susquehanna determined that the original 19-plate design battery (KCR-19) was close to its qualified end of life and replaced it with a new 21-plate design (KCR-21) battery. The team noted that the other Class 1E 125 Vdc batteries at Susquehanna are KCR-21 models. The replacement battery consisted of the same number of cells (60 cells). Each cell has the same physical dimensions, but weighs 9 pounds more and has one additional positive and one additional negative plate, resulting in additional maximum current available. Susquehanna determined that changing the type of battery and increasing its capacity will not adversely affect the design function of the control power, that the battery charger size was adequate to recharge the battery to full capacity; and that the extra weight of the new battery did not adversely impact the seismic qualifications of the battery bank.
The team reviewed the modification to evaluate whether the design and licensing bases and performance capability of the E EDG 125 Vdc system had been degraded by the modification. The team assessed Susquehanna's technical evaluations and design details, including installation specifications and calculations; and interviewed engineering personnel to evaluate whether the 125 Vdc system would function in accordance with the modification's assumptions and design requirements. Drawings and procedures were reviewed to verify that they were properly updated to reflect the post-modification configuration. The team also reviewed completed work orders to assess whether installation activities were performed as specified by the modification. A review of condition reports (CR) was performed to evaluate whether there were any reliability or performance issues associated with the post-modification configuration. The team walked down the battery with the system engineer to verify the modification was performed as described in the modification package. Additionally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.7 High Pressure Coolant Injection HV255F066 and Reactor Core Isolation Cooling
HV249F059 Valve Re-Wiring
a. Inspection Scope
The team reviewed EC-1684748 that reconfigured the close and open control circuit configuration of the Unit 2 high pressure coolant injection (HPCI) system and reactor core isolation cooling (RCIC) system turbine exhaust to suppression pool valves (HV255F066 and HV249F059, respectively) to remove a shorting (shunting) switch contact in the valves circuity. This modification was implemented to address a vulnerability that had been introduced during a prior control circuit modification (EC-1323000) for both valves. This vulnerability would only occur coincident with specific postulated operator errors. Both valves are maintained in the open position for normal, safe shutdown, and postulated accident scenarios and are also required to support non-accident safe shutdown functions. Both valves also have a safety function to close and to remain closed for long-term containment isolation. The RCIC valve also has the capability of being operated from the remote shutdown panel for Appendix R considerations.
The team reviewed the modification to evaluate whether the design and licensing bases and performance capability of the HPCI and RCIC systems had been degraded by the modification. The team assessed Susquehanna's technical evaluations and design details, including installation specifications; and interviewed engineering personnel to evaluate whether the turbine exhaust to suppression pool valves would function in accordance with the modification assumptions and design requirements. Drawings were reviewed to verify whether they were properly updated to reflect the post-modification design and operation. The post-modification test results were reviewed to verify that the acceptance criteria had been met. A review of CRs was performed to evaluate whether there were any reliability or performance issues associated with the post-modification configuration. Additionally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.8 Ashcroft Model 1009 Pressure Gauge Socket Material from Carbon Steel to Stainless Steel
a. Inspection Scope
The team reviewed modification EC-1361170 that dedicated a commercial-grade pressure indicating gauge manufactured by Ashcroft. The commercial-grade dedication evaluated the use of Ashcroft pressure indicating gauge model 1009SL, replacing existing models 1320RL and 1009RL due to equipment obsolescence. The new gauge is to be used in the A through E EDG jacket water, turbocharger lubricating oil, fuel oil, and lubricating oil systems. The 1320RL and 1009RL model became obsolete and are no longer supplied by the vendor. The original models socket was made of carbon steel, while the replacement model socket is constructed from stainless steel. Susquehanna determined that the change from carbon steel to stainless steel is acceptable with respect to dynamic qualifications basis and the two materials have similar strength and weight properties.
Susquehanna also determined that the pressure range, form, fit, and mounting configuration of the original and replacement gauges are similar.
The team reviewed the modification to confirm that the design and licensing bases and performance capability of the EDG systems had not been degraded by installation of the new pressure indication gauges. The team interviewed design engineers and reviewed vendor data and evaluations to evaluate whether the replacement gauge satisfied the applicable design requirements; and that the gauges critical characteristics were properly translated into receipt acceptance criteria. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.9 Reactor Core Isolation Cooling Pump Turbine Skid (1S212) Foundation
a. Inspection Scope
The team reviewed modification EC-1616476 that evaluated the lack of grout beneath a portion of the Unit 1 RCIC pump turbine skid (1S212). Specifically, the condition was an approximate 1 1/2 inch open space between the bottom of the support chair of the RCIC turbine skid and the top of the equipment pedestal, partially exposing six anchor bolts. The basic structural function of the grout at the base of the RCIC pump turbine skid was to provide a structurally adequate and level support for the turbine in order to assure that the RCIC system will function as required and would not be adversely affected by a potential misalignment. A seismic horizontal load from the RCIC turbine puts a shear load onto the six anchor exposed bolts due to the lack of grout. As a result, a moment is placed on the shaft of the bolts, which were not designed for bending. Susquehanna re-evaluated the calculation that provided the original design justification for the RCIC pump turbine skid mounting anchors and determined that the mounting skid was not adversely affected due to the relatively low magnitude of the shear loads on the mounting bolts during design basis events.
The team reviewed the revised calculation (EC-012-2064) to verify that the design and licensing bases, and performance capability of the RCIC system had not been degraded.
The team interviewed design engineers and reviewed calculations and evaluations to determine if the capability of the RCIC system met design and licensing requirements.
Drawings were reviewed to verify that they were properly updated. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.10 Remove the Overspeed Test Controller from Unit 1 and 2 High Pressure Coolant Injection
System Governor Control Circuits
a. Inspection Scope
The team reviewed EC-1643755 that removed the overspeed test controllers from the Unit 1 and 2 HPCI system control circuits. This overspeed test controller has not been used at Susquehanna for several years due to prior operating experience issues, and the HPCI turbine overspeed test is accomplished via other means. Specifically, the overspeed testing is performed by procedure using a portable controller that is temporarily connected to the control circuit. This modification installed a 200 ohm resistor in place of the overspeed test controller to configure the as-left controller circuit in an electrically equivalent condition.
The team reviewed the modification to evaluate whether the design and licensing bases and performance capability of the HPCI system had been degraded by the modification.
The team assessed Susquehanna's technical evaluations and design details, including installation specifications; and interviewed engineering personnel to evaluate whether the HPCI control system would function in accordance with the modification's assumptions and design requirements. Drawings and procedures were reviewed to verify whether they were properly updated to reflect the post-modification configuration. The team also reviewed completed work orders to assess whether installation activities were performed as specified by the revised design. The post-modification results were reviewed to verify that the acceptance criteria had been met. In addition, the team walked down the HPCI system to independently evaluate material conditions and configuration control with the approved design. A review of CRs was performed to evaluate whether there were any reliability or performance issues associated with the post-modification configuration.
Additionally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.11 BTT 1506934, Add Piping to Elevate Residual Heat Removal/Reactor Core Isolation
Cooling/High Pressure Coolant Injection Steam Vents to Above Postulated Flood Level during an Induced Cooling Tower Basin Flood
a. Inspection Scope
The team reviewed modification BTT 1506934 that installed vertical extensions of the RHR/RCIC/HPCI steam vent penetrations, located on exterior of the Unit 1 and 2 reactor buildings. The steam vents provide for a pressure relief function for the rooms blowout panels. Susquehanna determined that vertical vent extensions were necessary to prevent adverse consequences in the event of a design basis rupture of the cooling tower basin. An elevated release point from the existing vents would ensure that water during the postulated flooding event would not enter the steam vents.
The team reviewed the modification to verify that the design and licensing bases of systems had not been degraded by installing extensions to the RHR, RCIC, and HPCI steam flood vent penetrations and to confirm that the components met the appropriate quality standards. The team also reviewed post-modification testing of the equipment to verify proper operation and interaction with the existing system controls. The team also conducted a walkdown of the affected components. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section
1R17 .1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.12 Reactor Core Isolation Cooling Pump Seal Leakage during Extended Loss of Alternating
Current Power Scenario Calculation
a. Inspection Scope
The team reviewed a new calculation EC-050-1034, RCIC Pump Seal Leakage during Extended Loss of AC Power Scenario, to assess RCIC room flooding impacts due to RCIC pump seal leakage during a postulated extended loss of power event. In the postulated scenario, RCIC would be used for reactor vessel injection since its capability closely matches inventory boil-off from decay heat. Susquehanna determined the magnitude of RCIC pump seal leakage in the event of a complete seal failure, and evaluated whether the seals would fail under the conditions that RCIC would be operated during the postulated scenario (at elevated seal temperatures).
The team reviewed the calculation to verify that the assumptions and calculation methodology adequately evaluated the postulated scenario. In particular, the team reviewed design and seal vendor documentation to evaluate maximum expected seal temperatures and the seals ability to withstand expected configurations. The team also reviewed applicable strategies and procedures intended to provide for sufficient RCIC pump seal cooling during a postulated extended loss of power event. The team walked down the RCIC pump, seal, and associated components to identify abnormal conditions.
The 10 CFR 50.59 screening determination associated with this calculation was also reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.13 Engineered Safeguard Service Water Pump Structure Heat Loss Calculation Revision
a. Inspection Scope
The team reviewed a revision to calculation EC-028-0009, Engineered Safeguard Service Water (ESSW) Pump Structure Heat Loss, to account for pump room doors being opened at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into a postulated event under extreme cold weather conditions.
The calculation evaluated heat transfer in the pump rooms and the impact to safety-related equipment. Based upon the evaluated conditions with specific doors being opened during an extended loss of power event, the calculation determined that pumphouse room heaters would need to be re-powered within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of opening the doors.
The team reviewed the calculation to verify that the assumptions and calculation methodology adequately evaluated the postulated scenario. The team also reviewed applicable strategies and associated procedures designed to respond to the extended loss of power event. The team evaluated whether Susquehanna had the appropriate plans, equipment, and procedures in place to respond to the event. The team walked down the ESSW building and associated components to identify abnormal conditions and to assess operators ability to complete the required actions. The 10 CFR 50.59 screening determination associated with this revised calculation was also reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the
.
b. Findings
No findings were identified.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems (IP 71152)
a. Inspection Scope
The team reviewed a sample of CRs associated with 10 CFR 50.59 and plant modification issues to evaluate whether Susquehanna was appropriately identifying, characterizing, and correcting problems associated with these areas, and whether the planned or completed corrective actions were appropriate. In addition, the team reviewed CRs written on issues identified during the inspection to verify adequate problem identification and incorporation of the issues into the corrective action system.
The CRs reviewed are listed in the Attachment.
b. Findings
No findings were identified.
4OA6 Meetings, including Exit
The team presented the preliminary inspection results to Mr. Jon Franke, Site Vice President, and other members of Susquehanna staff at a meeting on September 18, 2015. The team returned proprietary information reviewed during the inspection and verified that this report does not contain proprietary information.
ATTACHMENT
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- M. Chaiko, Senior Staff Engineer
- E. Griffiths, Station Engineer
- J. Oswald, Design Engineer
- P. Penny, Design Engineering Supervisor
- M. Radvansky, Design Engineer
- K. Scherer, Support Engineer
- J. Smith, Nuclear Fuels Supervisor
NRC Personnel
- T. Daun, Resident Inspector - Susquehanna
ITEMS OPENED, CLOSED AND DISCUSSED
None