ML17052A125
| ML17052A125 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/09/2017 |
| From: | Klos L Plant Licensing Branch IV |
| To: | Reddemann M Energy Northwest |
| Klos L, NRR/DORL/LPLIV, 415-5136 | |
| References | |
| CAC MF7699 | |
| Download: ML17052A125 (16) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 9, 2017 Mr. Mark E. Reddemann Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)
Richland, WA 99352-0968
SUBJECT:
COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT RE: TO MODIFY TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS 3.4.3.1AND3.4.4.1 SAFETY/RELIEF VALVE SETPOINT LOWER TOLERANCE (CAC NO. MF7699)
Dear Mr. Reddemann:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 240 to Renewed Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the technical specifications (TSs) in response to your application dated May 10, 2016, as supplemented by letters dated May 18, 2016, and January 31, 2017.
The amendment revises the safety function lift and lower setpoint tolerances of the safety/relief valves that are listed in Surveillance Requirements 3.4.3.1 and 3.4.4.1 of the TSs.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Since2:P ~
L. John Klos, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosures:
- 1. Amendment No. 240 to NPF-21
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 240 License No. NPF-21
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Energy Northwest (licensee), dated May 10 2016, as supplemented by letters dated May 18, 2016, and January 31, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Enclosure 1
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-21 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 240 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License No. NPF-21 and Technical Specifications Date of Issuance: March 9 , 2O1 7
ATTACHMENT TO LICENSE AMENDMENT NO. 240 COLUMBIA GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of the Renewed Facility Operating License No. NPF-21 and Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Facility Operating License REMOVE INSERT Technical Specification REMOVE INSERT 3.4.3-1 3.4.3-1 3.4.3-2 3.4.4-2 3.4.4-2
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 240 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- a. For Surveillance Requirements (SRs) not previously performed by existing SRs or other plant tests, the requirement will be considered met on the implementation date and the next required test will be at the interval specified in the Technical Specifications as revised in Amendment No. 149.
(3) Deleted.
(4) Deleted.
(5) Deleted.
(6) Deleted.
(7) Deleted.
(8) Deleted.
(9) Deleted.
(10) Deleted.
(11) Shield Wall Deferral (Section 12.3.2. SSER #4. License Amendment #7)
The licensee shall complete construction of the deferred shield walls and window as identified in Attachment 3, as amended by this license amendment.
(12) Deleted.
(13) Deleted.
- The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Renewed License No. NPF-21 Amendment No. 240
SRVs - ~ 25% RTP 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (SRVs) - ~ 25% RTP LCO 3.4.3 The safety function of 12 SRVs shall be OPERABLE, with two SRVs in the lowest two lift setpoint groups OPERABLE.
APPLICABILITY: THERMAL POWER;;::: 25% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SRVs inoperable. POWER to < 25% RTP.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the required In accordance SRVs are as follows: with the lnservice Testing Program Number of Set point SRVs _{Qfilg}
2 1165 + 34.9
- 58.2 4 1175 + 35.2
- 58.7 4 1185 + 35.5
- 59.2 4 1195 + 35.8
- 59.7 4 1205 + 36.1
-60.2 Following testing, lift settings shall be within +/-3%.
Columbia Generating Station 3.4.3-1 Amendment No. 449,409 225 238, 240
SRVs - ~ 25% RTP 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.2 Verify each required SRV opens when manually In accordance actuated. with the Surveillance Frequency Control Program Columbia Generating Station 3.4.3-2 Amendment No. 240
SRVs - < 25% RTP 3.4.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify the safety function lift setpoints of the required In accordance SRVs are as follows: with the lnservice Testing Program Number of Setpoint SRVs JQ§i9l 2 1165 + 34.9
- 58.2 4 1175 + 35.2
- 58.7 4 1185 + 35.5
- 59.2 4 1195 + 35.8
- 59.7 4 1205 + 36.1
- 60.2 Following testing, lift settings shall be within +/-3%.
SR 3.4.4.2 -------------------------------N 0 TE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify each required SRV opens when manually In accordance actuated. with the Surveillance Frequency Control Program Columbia Generating Station 3.4.4-2 Amendment No. 449,4-W,',236,238, 240
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By application dated May 10, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16131A891 ), as supplemented by letters dated May 18, 2016, and January 31, 2017 (ADAMS Accession Nos. ML16139A161 and ML17031A450, respectively),
Energy Northwest (the licensee) requested changes to the Technical Specifications (TS; Appendix A to Renewed Facility Operating License No. NPF-21) for the Columbia Generating Station (CGS).
The supplemental letter dated January 31, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on July 19, 2016 (81 FR 46961 ).
This change to the CGS TS Surveillance Requirements (SRs) 3.4.3.1 and 3.4.4.1 revises the safety function lift setpoint lower tolerance for the safety/relief valves (SRVs). The change revises the tolerances from plus/minus (+/-)3 percent to plus (+)3 percent, minus (-)5 percent.
This change to the SRs is limited to the lower tolerance and does not affect the upper tolerance.
This change only applies to the lower as-found tolerance and not to the as-left tolerance, which will remain unchanged at +/-3 percent of the safety lift setpoint. The as-found tolerances are used for determining operability and to increase sample sizes for SRV testing. There will be no revision to the actual setpoints of the valves installed in the plant due to this change.
The licensee stated that the change reduces an unnecessarily restrictive SR and that the change will not impact the reliability of the SRVs or adversely impact their ability to perform their safety function. The SRVs are required to meet the American Society of Mechanical Engineers (ASME) limits based on valve type and size to ensure acceptable valve performance. These limits are not being changed. This amendment will preclude the submittal of unnecessary licensee event reports to the NRC due to set point drift in the low (conservative) direction.
Enclosure 2
2.0 REGULATORY EVALUATION
The Commission's regulatory requirements related to the contents of TS, set forth in Title 1O of the Code of Federal Regulations (1 O CFR) 50.36, "Technical specifications," require that the TS limiting conditions for operation (LCOs) are consistent with assumed values of the initial conditions in the licensee's safety analyses. As stated in 10 CFR 50.36, the TSs will include SRs to assure that the LCOs (and associated remedial actions) are met. Specifically, 10 CFR 50.36(c)(2)(i) states, in part:
Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
The regulation in 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether an LCO is required to be included in TSs. These criteria require an LCO for: (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant system pressure boundary; (2) initial plant conditions that are assumed in a design-basis transient and accident analysis; (3) components or systems that are used for mitigating consequences of the design-basis transient and accident; and (4) components or systems, which probabilistic risk assessment has shown to be significant to public health and safety.
The SRVs are part of the primary success path and are assumed in the Final Safety Analysis Report in the accident and safety analyses to mitigate the effects of the licensing-basis accidents. In accordance with Criterion 3 of 10 CFR 50.36(c)(2)(ii) discussed above, TS LCOs are required for both the SRV safety and relief modes of operation. Since the setpoint tolerance and SRs for SRVs are proposed to be changed, the licensee is required to provide acceptable analyses to support the adequacy of the TS changes.
General Design Criterion 15 - Reactor Coolant System Design General Design Criterion (GDC) 15, "Reactor coolant system design," of Appendix A of 10 CFR 50 states:
The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
General Design Criterion 35 - Emergency Core Cooling GDC 35, "Emergency Core Cooling," of Appendix A of 10 CFR 50 states:
A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
3.0 TECHNICAL EVALUATION
3.1 System Description CGS has a total of 18 SRVs installed on the four main steam lines. Each SRV is a Crosby 6xRx10, Style HB-65-BP, originally built to ASME Section Ill, 1971 Edition with no Addenda. Per TS LCO 3.4.3, 12 SRVs shall be operable to provide the safety function, with 2 SRVs in the lowest two lift setpoint groups operable. The SRVs are actuated in either of two modes: 1) safety mode or 2) relief mode.
In the safety mode (or the spring mode of operation) the valves open when steam pressure at the valve inlet overcomes the spring force holding the valve closed. This mode satisfies the ASME Boiler and Pressure Vessel Code (ASME Code) requirements. It is this mode of operation for which the lower surveillance tolerances, or the as-found tolerances, for the safety function lift setpoints will be relaxed from -3 percent to -5 percent. The upper surveillance tolerances, or the as-left tolerances, will remain at +/-3 percent, as approved by the NRC via Amendment No. 137, dated May 2, 1995 (ADAMS Accession No. ML022120154).
In the relief mode, a pneumatic piston/cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force, even with the valve inlet pressure equal to 0 pounds per square inch gauge (psig). In this mode, valves may be opened manually or automatically at the selected preset pressure. Seven of the SRVs that provide the safety and relief function are part of the automatic depressurization system (ADS). The relief and automatic depressurization modes rely upon powered actuation to open the valves and are not affected by this proposed change.
3.2 NRC Staff Evaluation of Proposed SRV Setpoint Tolerance In the license amendment request the licensee stated that the CGS SRVs are Crosby Style 6xRx1 O, Style HB-65-BP. Industry experience shows that it is the nature of these valves to have a drift-variance in the negative direction. The valve manufacturer attributes these variances to no cause other than setpoint drift/variance. A review of as-found test data for the CGS SRVs from 2005 to 2015 (including 97 SRV tests) indicates a tendency for minor setpoint drift in both the positive and negative direction with an average drift of -0.88 percent with a median of -0.43 percent. There were 16 valves that failed below the -3 percent tolerance and only 1 valve which had a greater than -5 percent deviation from its setpoint.
According to Crosby Valve and Gage Company Procedure 1-11069, "Instruction Manual for Crosby Style 6xRx10 HB-65-BP Safety Relief Valve for Main Steam Service," the low limit of the setpoint tolerance may be extended to 1,067 psig (if all other valve functional requirements are met) and still assure normal valve response. The proposed lower setpoint tolerance of -5 percent is within the vendor's allowable limit. Since the high setpoint tolerance is unchanged, the capability of the SRVs to ensure ASME overpressure protection is maintained.
Therefore, the NRC staff finds the proposed TS change of lower setpoint tolerance from -3 percent to -5 percent to be within the acceptable limits of the valve manufacture.
The purpose of the lower setpoint tolerance is to ensure sufficient margin exists between the normal operating pressure of the system and the point at which the SRVs actuate in the safety mode. The nominal operating pressure of the reactor pressure vessel (RPV) at power is 1,020 psig. For the lowest as-found setpoint of -5 percent, the SRV set at the lowest nominal pressure (1, 165 psig) would lift at 1, 107 psig and provide an operating margin of 87 pounds per square inch differential (psid) between the nominal reactor operating pressure and the SRV lift pressure. This provides sufficient margin and is sufficient to prevent unwanted actuation of the SRVs postulated to occur during pressurization transients. In addition, the lowest relief mode opening pressure setpoint for the SRVs is 1,091 psig. This is 16 psid lower than the lowest opening pressure with the -5 percent setpoint tolerance of the SRVs in safety mode. The NRC staff therefore finds that there will be no additional inadvertent SRV lifts due to the lower -5 percent setpoint tolerance.
The licensee qualitatively evaluated the proposed safety setpoint lower tolerance change from -3 percent to -5 percent in regards to vessel overpressure protection, anticipated operational occurrence thermal limits and anticipated transients without scram (ATWS). For overpressure protection, higher opening pressures and/or later opening of the SRVs provides the most conservative results as the objective of overpressure events is to maximize vessel dome and thus vessel bottom pressures. The NRC staff finds that the proposed change to decrease the SRV safety mode setpoint lower tolerance from -3 percent to -5 percent could allow opening of the SRVs earlier, which results in a benefit to the ASME overpressure protection analyses. The NRC staff finds that the proposed change is in compliance with GDC 15.
The licensee stated that for the events that establish the operating limit minimum critical power ratio (OLMCPR) and the off-rated limits, later opening of the SRVs provides the most conservative results. The pressure feedback on the core is larger with later opening times such that the power feedback is worse, and as a result, the OLMCPR would be worse if the time of the minimum critical power ratio was after the SRV opening time. The NRC staff finds that the proposed change to the SRV safety mode setpoint lower tolerance from -3 percent to -5 percent could allow the SRVs to open sooner, which would result in either a benefit or no change to the OLMCPR and off-rated limits.
The licensee states that for the ATWS analyses, nominal values for the relief mode of the SRVs are used. Since the proposed change is only for the SRV safety mode setpoint tolerance and does not propose any change to the SRV relief mode setpoints, the NRC staff finds there would be no impact on the ATWS analyses.
The licensee evaluated the impact of the setpoint tolerance change on the performance characteristics of the high pressure core spray (HPCS), reactor core isolation cooling (RCIC),
and standby liquid control (SLC) systems. The HPCS system is a safety system and part of the emergency core cooling system (ECCS). HPCS provides high pressure water injection into the RPV in the event of a loss-of-coolant accident (LOCA). Additionally, HPCS is a backup to the RCIC system to provide makeup water in the event of a loss of feedwater (LOFW) flow transient. For the LOCA events, the licensee states that credit is taken for the safety mode of the SRV, and the analysis uses the opening pressures of the SRV safety mode setpoint upper tolerance of +3 percent. Since the proposed change is a decrease in SRV safety mode setpoint lower tolerance, the NRC staff finds there is no impact on the HPCS system.
For LOFW events that do not isolate the RPV, the licensee states that the RPV pressure is maintained by the turbine bypass valves at or near the pressure regulator setpoint. With the RPV pressures near the pressure regulator setpoint, HPCS and RCIC operation are not affected by changes in the SRV performance. For LOFW events with a closure of main steam isolation valves, the licensee states that RPV pressure is dictated by the opening setpoint of the relief mode of the SRVs. Since the proposed change does not make any change to the SRV relief mode setpoint, the NRC staff finds there is no impact on the HPCS or RCIC systems for LOFW.
The SLC system is designed to shut down the reactor from power conditions to cold shutdown in the postulated situation of a failure of the control rods to insert. The licensee states that the SLC system is designed for injection at a maximum RPV pressure equal to the upper analytical limit for the lowest group of SRVs opening in the relief mode. Since the proposed change does not make any change to the SRV relief mode setpoint, the NRC staff finds there is no impact on the SLC system.
The ADS functions to reduce the reactor pressure so that flow from low-pressure coolant injection and core spray enters the reactor vessel in time to cool the core and prevent excessive fuel clad temperature. The ADS uses several of the SRVs to relieve the high pressure steam to the suppression pool. The NRC staff finds that the proposed change has no impact on the ECCS and is therefore in compliance with GDC 35.
One of the issues of concern to the NRC staff was of possible effects of increased valve seat leakage due to lower simmer margins. The expanded lower as-found tolerance could result in less testing and maintenance applied to the valves; therefore, this might result in an increased tendency for the SRVs to drift to lower setpoints and leak. With increased leakage there is an increased risk of a damaging water hammer occurrence following a loss of offsite power as described in NRC Information Notice (IN) 87-10, "Potential for Water Hammer During Restart of Residual Heat Removal Pumps," dated February 11, 1987, and IN 87-10, Supplement 1, dated May 15, 1997. However, since the licensee will continue to apply the +/-3 percent as-left tolerance for resetting the SRVs and since the valves are not expected to drift significantly greater amounts for longer operational service, the -5 percent as-found tolerance is not expected to result in increased leakage.
Based on its evaluation of the licensee's application, the NRC staff finds that the licensee has adequately justified the proposed changes to the TSs for CGS. Specifically, the NRC staff finds that the change to the lower SRV tolerance:
- Is not significant to SRV operation during overpressure transients.
- The nominal setpoints for the relief and safety modes are not changed. The nominal operating pressure of the RPV at power is 1,020 psig. For the lowest as-found setpoint of -5 percent, the SRV set at the lowest nominal pressure (1, 165 psig) would lift at 1, 107 psig providing an operating margin of 87 pounds per square inch between the nominal reactor operating pressure and the SRV lift pressure. This provides sufficient margin and is sufficient to prevent unwanted actuation of the SRVs postulated to occur during pressurization transients.
- The relief mode for two SRVs is set at 1,091 psig. The relief mode on these two SRVs would actuate before the above stated 1, 107 psig lowest allowable lift in safety mode.
The additional margin of 16 psig is sufficient to prevent unwanted actuation of the SRVs, since the relief mode setpoint of 1,091 psig ensures that pressure transients will still cause the valves to open in the relief mode prior to the safety mode. Therefore, there will be no additional inadvertent SRV lifts.
- According to Crosby Valve and Gage Company Procedure 1-11069, the low limit of the setpoint tolerance may be extended to 1,067 psig and still assure normal valve response. The proposed lower setpoint tolerance of -5 percent is within the vendor's allowable limit. Since the high setpoint tolerance is unchanged, the capability of the SRVs to ensure ASME overpressure protection is maintained.
- The NRC staff has determined that the proposed change will not affect plant overpressure protection.
- The NRC staff finds the change to be compliant with 10 CFR 50.36, GDC 15 and GDC 35.
Therefore, the NRC staff has concluded that the proposed changes to TS SR 3.4.3.1 and SR 3.4.4.1 are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Washington State official was notified of the proposed issuance of the amendment on February 22, 2017. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Registeron July 19, 2016 (81 FR 46961).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Robert Beaton Date: March 9, 2017
ML17052A125 *via email OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/SRXB/BC* NRR/DE/EICB/BC NRR/DSS/STSB/BC NAME JKlos PBlechman EOesterle MWaters AKlein DATE 2/22/17 2/22/2017 1/10/2017 3/2/2017 2/23/2017 OFFICE NRR/DE/EPNB/BC OGG- NLO NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME DAiiey BMizuno RPascarelli JKlos DATE 2/23/2017 3/8/2017 3/9/2017 3/9/2017