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Category:Letter
MONTHYEARML24213A3352024-07-31031 July 2024 License Amendment Request to Revise Technical Specification 3.2.1, Heat Flux Hot Channel Factor (Fq(Z)) (Fq Methodology), to Implement the Methodology from WCAP-17661-P-A, Revision 1. ML24206A1252024-07-24024 July 2024 Revision of Three Procedures and Two Forms That Implement the Radiological Emergency Response Plan (RERP) IR 05000482/20240022024-07-18018 July 2024 Integrated Inspection Report 05000482/2024002 IR 05000482/20244012024-07-0202 July 2024 Security Baseline Inspection Report 05000482/2024401 ML24178A3672024-06-26026 June 2024 Correction to 2023 Annual Radioactive Effluent Release Report – Report 47 ML24178A4142024-06-26026 June 2024 Revision of One Procedure and One Form That Implement the Radiological Emergency Response Plan (RERP) ML24162A1632024-06-11011 June 2024 Operating Corporation – Notification of Biennial Problem Identification and Resolution Inspection and Request for Information (05000482/2024010) ML24150A0562024-05-29029 May 2024 Foreign Ownership, Control or Influence (FOCI) Information – Change to Lists of Owners, Officers, Directors and Executive Personnel - Form 405F Amendment ML24089A2622024-04-29029 April 2024 Financial Protection Levels ML24118A0022024-04-27027 April 2024 Wolf Generating Nuclear Station - 2023 Annual Radiological Environmental Operating Report ML24118A0032024-04-27027 April 2024 2023 Annual Radioactive Effluent Release Report - Report 47 ML24113A1882024-04-19019 April 2024 Foreign Ownership, Control or Influence Information - Change to Lists of Owners, Officers, Directors and Executive Personnel - Form 405F Amendment ML24109A1842024-04-18018 April 2024 Cycle 27 Core Operating Limits Report ML24109A1212024-04-18018 April 2024 (WCGS) Form 5 Exposure Report for Calendar Year 2023 IR 05000482/20240012024-04-17017 April 2024 Integrated Inspection Report 05000482/2024001 ML24106A1482024-04-15015 April 2024 Notification of Inspection (NRC Inspection Report 05000482/2024003) and Request for Information ML24114A1442024-04-15015 April 2024 Redacted Updated Safety Analysis Report (WCGS Usar), Revision 37 ML24098A0052024-04-0707 April 2024 2023 Annual Environmental Operating Report ML24089A1352024-03-29029 March 2024 10 CFR 50.46 Annual Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes ML24089A0972024-03-29029 March 2024 Response to NRC Regulatory Issue Summary 2024-01, “Preparation and Scheduling of Operator Licensing Examinations” ML24025A0992024-01-25025 January 2024 Withdrawal of Request for Exemption from Specific 10 CFR Part 73 Requirements ML24018A0792024-01-22022 January 2024 Individual Notice of Consideration of Issuance of Amendment to Renewed Facility Operating License, Proposed Nshcd and Opportunity for a Hearing (EPID L-2024-LLA-0007) (Letter) ML24018A2482024-01-18018 January 2024 License Amendment Request to Modify the Implementation Date of License Amendment No. 238 ML24018A1382024-01-18018 January 2024 Inservice Inspection Request for Information IR 05000482/20230042024-01-11011 January 2024 Integrated Inspection Report 05000482/2023004 ML23356A0722024-01-0404 January 2024 Supplemental Information Needed for Acceptance of Requested Licensing Actions Exemption from Specific 10 CFR Part 73N Requirements (EPID L-2023-LLE-0048) (Redacted Version) WO 23-0002, Operating Corp., Summary of Actions Implemented for EA-18-165, Confirmatory Order, NRC Inspection Report 05000482/2019010 and NRC Investigation Report 4-2018-0082023-12-26026 December 2023 Operating Corp., Summary of Actions Implemented for EA-18-165, Confirmatory Order, NRC Inspection Report 05000482/2019010 and NRC Investigation Report 4-2018-008 ML23348A3662023-12-18018 December 2023 Notification of an NRC Fire Protection Baseline Inspection (NRC Inspection Report 05000482 2024012) and Request for Information ML23334A2502023-11-30030 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23299A2662023-11-29029 November 2023 Issuance of Amendment No. 238 Modified Implementation Date of License Amendment No. 237 ML23331A4972023-11-27027 November 2023 Supplement to License Amendment Request to Modify the 90-Day Implementation of License Amendment No. 237 ML23325A2112023-11-20020 November 2023 Submittal of Request for Exemption from Specific Provisions in 10 CFR 73.55 ML23320A2772023-11-16016 November 2023 License Amendment Request to Revise Ventilation Filter Testing Program Criteria in Technical Specification 5.5.11.b and Administrative Correction of Absorber in Technical Specification 5.5.11 ML23311A2082023-11-0909 November 2023 Reassignment of U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch IV ML23305A3472023-11-0101 November 2023 Revision of One Form That Implements the Radiological Emergency Response Plan (RERP) ML23292A3572023-10-19019 October 2023 Operating Corporation, Request for Correction in Response to Issuance of Amendment No. 237 and Corresponding Safety Evaluation ML23292A3592023-10-19019 October 2023 License Amendment Request to Modify the 90-Day Implementation of License Amendment No. 237 ML23284A2392023-10-11011 October 2023 NRC Initial Operator Licensing Examination Approval 05000482/2023301 ML23277A2772023-10-11011 October 2023 Regulatory Audit Summary Concerning Review of Request Number CI3R-01 for Proposed Alternative Inspection Frequency for Containment Unbonded Post Tensioning System Components ML23276B4552023-10-0303 October 2023 Technical Specification 5.6.8 Post Accident Monitoring (PAM) Report ML23264A8682023-09-21021 September 2023 Independent Spent Fuel Storage Installation Registration of Dry Storage Canister WC-EOS-37PTH-008-B4L, WC-EOS-37PTH-007-B4L, and WC-EOS-37PTH-002-B4L ML23256A2062023-09-14014 September 2023 Withdrawal of Requested Licensing Action Exemption from Specific 10 CFR Part 73 Requirements ML23165A2502023-08-31031 August 2023 Issuance of Amendment No. 237 Request for Deviation from Fire Protection Requirements ML23243A0002023-08-31031 August 2023 Wolf Generating Station - Withdrawal of Request for Exemption from Specific 10 CFR Part 73 Requirements IR 05000482/20230052023-08-23023 August 2023 Updated Inspection Plan for Wolf Creek Nuclear Operating Corporation, Unit 1 (Report 05000482/2023005) - Mid-Cycle Letter 2023 2024-07-31
[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML22361A0052022-12-27027 December 2022 NRR E-mail Capture - Request for Additional Information Wolf Creek Request for Deviation from Fire Protection Program Requirements ML22321A2662022-11-16016 November 2022 January 2023 Emergency Preparedness Program Inspection - Request for Information ML22307A1372022-11-0202 November 2022 August 2022 Emergency Preparedness Exercise Inspection Unresolved Item - Request for Information ML22249A2552022-06-29029 June 2022 Request for Information for an NRC Post-Approval Site Inspection for License Renewal 05000482/2022012 ML22165A0882022-06-14014 June 2022 Information Request, Security IR 2022403 ML22166A3222022-06-14014 June 2022 August 2022 Emergency Preparedness Exercise Inspection - Request for Information ML22055A1142022-02-23023 February 2022 NRR E-mail Capture - Request for Additional Information - Wolf Creek Request to Revise Diesel Generator Completion Time ML22045A4512022-02-14014 February 2022 NRR E-mail Capture - Request for Additional Information - Wolf Creek Steam Generator Inspection Report 24th Refueling ML22025A3612022-01-21021 January 2022 PI&R RFI Final ML21327A2602021-11-23023 November 2021 NRR E-mail Capture - Request for Additional Information - Wolf Creek Revision of Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation ML21271A1952021-09-28028 September 2021 E-mail 9-28-21 RFI for Wc EP Inspection Nov 2021 ML21253A0902021-09-23023 September 2021 Supplemental Information Needed for Acceptance of Requested Licensing Actions License Amendment and Regulatory Exemption for a Risk-Informed Approach to Address GSI-191 (EPIDs L-2021-LLA- 0152 and L-2021-LLE-0039) ML21221A1282021-08-0505 August 2021 NRR E-mail Capture - Urgent: Draft Request for Additional Information - Wolf Creek one-time Request for Exemption from the Biennial Emergency Preparedness Exercise ML21132A1032021-05-12012 May 2021 Document Request List, Paperwork Reduction Act Statement ML20302A4782020-10-28028 October 2020 OPC R1 Request for Information ML20262A7512020-09-17017 September 2020 NRR E-mail Capture - Requests for Additional Information: Wolf Creek Amendment Request to Modify Technical Specification Surveillance Frequency Consistent with TSTF-425 ML20195A2712020-07-13013 July 2020 Generation Station - Notification of an NRC Fire Protection Baseline Inspection (NRC Inspection Report 05000482/2020012) and Request for Information ML20133J8972020-05-12012 May 2020 12 May 2020 E-mail - RFI for NRC In-Office Inspection of Recent Wcngs Eplan_Eal Changes ML20009E4802020-01-0909 January 2020 Relief Request to Utilize Code Case N-666-1 for Wolf Creek Generating Station - Request for Additional Information ML19346E6052019-11-0404 November 2019 Target Set Request for Information for Week of November 4, 2019 ML19224A5242019-08-12012 August 2019 NRR E-mail Capture - Request for Additional Information - License Amendment Request to Revise Wolf Creek Generating Station Technical Specification 3.3.5 ML18304A1052018-11-0505 November 2018 Request for Additional Information License Amendment Request for Transition to Westinghouse Core Design and Safety Analyses and Alternative Source Term (CAC No. MF9307; EPID L-2017-LLA-0211) ML18282A6402018-10-0909 October 2018 NRR E-mail Capture - Request for Additional Information - Wolf Creek Generating Station License Amendment Request for Revision to the Emergency Plan ML18270A0942018-10-0404 October 2018 Request for Additional Information License Amendment Request for Transition to Westinghouse Methodology for Selected Accident and Transient Analyses ML18207A4332018-07-24024 July 2018 NRR E-mail Capture - (External_Sender) Draft Request for Additional Information Relief Request from ASME Code N-666-1 Alternate Repair of Essential Service Water Piping Wolf Creek Generating Station, Unit 1 EPID No.:L-2018-LLR-0101 ML18116A6132018-04-25025 April 2018 Operating Corporation - Notification of Inspection (NRC Inspection Report 05000482/2018002) and Request for Information ML17331A1782017-12-0404 December 2017 Request for Additional Information License Amendment Request for Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternate Source Term (CAC No. MF9307; EPID L-2017-LLA-0211) ML17291A7102017-10-17017 October 2017 NRR E-mail Capture - Request for Additional Information - License Amendment Request (LAR) for Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term ML17265A0142017-09-21021 September 2017 NRR E-mail Capture - Request for Additional Information - License Amendment Request for Transition to Westinghouse Core Design and Safety Analyses ML17170A3152017-06-19019 June 2017 Notification of an NRC Triennial Fire Protection Baseline Inspection (NRC Inspection Report 05000482/2017008) and Request for Information ML17166A0382017-06-14014 June 2017 NRR E-mail Capture - Request for Additional Information - License Amendment Request for Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term ML17052A0282017-02-15015 February 2017 NRR E-mail Capture - Wolf Creek Generating Station - Official EAL RAIs ML16319A4282016-11-14014 November 2016 NRR E-mail Capture - Request for Additional Information (RAI) - Relief Requests 13R-14 and 13R-15 ML16300A2662016-10-21021 October 2016 Request for Additional Information - Relief Request I3R-13 Regarding Weld Examination Coverage IR 05000482/20160022016-08-0303 August 2016 NRC Integrated Inspection Report 05000482/2016002 ML16110A3722016-04-25025 April 2016 Request for Additional Information, License Amendment Request to Revise Technical Specification (TS) 4.2.1 and TS 5.6.5 to Allow Use of Optimized Zirlo as Approved Fuel Rod Cladding ML16020A1392016-01-15015 January 2016 Notification of NRC Component Design Bases Inspection (05000482/2016007) and Initial Request for Information ML15135A4862015-05-14014 May 2015 Wc 2015007 17T RFI ML15082A0052015-03-27027 March 2015 Request for Additional Information, Round 2, Request to Revise Technical Specifications to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation, Using the Consolidated Line Item Improvement Process ML15079A0762015-03-23023 March 2015 Request for Additional Information, Round 2, Flood Hazard Reevaluation Report for Recommendation 2.1 of the Near-Term Task Force Review of the Insights from the Fukushima Dai-Ichi Accident ML15040A6252015-02-10010 February 2015 Request for Additional Information, License Amendment Request to Revise Technical Specifications to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation, Using the Consolidated Line Item Improvement Process ML14323A5742014-12-0404 December 2014 Request for Additional Information, License Amendment Request to Revise Fire Protection Program Related to Alternative Shutdown Capability as Described in USAR ML14230A7572014-08-21021 August 2014 Request for Additional Information, Relief Request I3R-10, Alternative from Pressure Test Requirements of ASME Code Section XI, IWC-5220, Third 10-Year Inservice Inspection Interval ML14206A0122014-08-0101 August 2014 Request for Additional Information, Relief Requests I3R-08, RPV Interior Attachments and I3R-09, RPV Pressure-Retaining Welds, Exam Interval Extensions, Third 10-year Inservice Inspection Interval ML14197A3362014-07-25025 July 2014 Request for Additional Information, Relief Request I3R-11, Alternative from Pressure Test Requirements of ASME Code Section XI IWC-5220, Third 10-Year Inservice Inspection Interval ML14148A3872014-06-0202 June 2014 Request for Additional Information Related to Flood Hazard Reevaluation Report for Recommendation 2.1 of the Near-Term Task Force Review of the Insights from the Fukushima Dai-Ichi Accident ML14111A1002014-04-30030 April 2014 Request for Additional Information, Nrr/Dss/Srxb, License Amendment Request to Approve Transition to Westinghouse Core Design and Safety Analysis and Adoption of Full Scope Alternate Source Term ML14083A4002014-04-0303 April 2014 Request for Additional Information, Nrr/Dra/Arcb, License Amendment Request to Approve Transition to Westinghouse Core Design and Safety Analysis and Adoption of Full Scope Alternate Source Term ML14058A0882014-03-0505 March 2014 Request for Additional Information, Nrr/De/Eeeb, License Amendment Request to Approve Transition to Westinghouse Core Design and Safety Analysis and Adoption of Full Scope Alternate Source Term ML14027A1622014-01-28028 January 2014 Redacted, Request for Additional Information, Round 2, License Amendment Request to Approve Transition to Westinghouse Core Design and Safety Analysis and Adoption of Full Scope Alternate Source Term 2022-06-29
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 4, 2014 Mr. Adam C. Heflin President, Chief Executive Officer, and Chief Nuclear Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839
SUBJECT:
WOLF CREEK GENERATING STATION-REQUEST FOR ADDITIONAL INFORMATION RE: LICENSE AMENDMENT REQUEST TO REVISE THE FIRE PROTECTION PROGRAM RELATED TO ALTERNATIVE SHUTDOWN CAPABILITY (TAC NO. MF3112)
Dear Mr. Heflin:
By letter dated November 21, 2013, Wolf Creek Nuclear Operating Corporation (the licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC) a license amendment request to revise the fire protection program as described in the Updated Safety Analysis Report (USAR) for the Wolf Creek Generating Station. The licensee's proposed changes would incorporate a revised alternate shutdown methodology into the USAR. The NRC staff has reviewed the information provided in your application and determined that additional information is required in order to complete its formal review. The enclosed questions were provided to Mr. S. Wideman of your staff on December 4, 2014. Please provide a response to the enclosed questions within 45 days of the date of this letter. If you have any questions, please contact me at 301-415-2296 or via e-mail at fred.lyon@nrc.gov. Docket No. 50-482 Enclosure Request for Additional Information cc w/encl: Distribution via Listserv Sincerely, Carl F. Lyon, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST TO REVISE THE FIRE PROTECTION PROGRAM WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 By letter dated November 21, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13331A728), Wolf Creek Nuclear Operating Corporation (the licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC) a license amendment request to revise the fire protection program as described in the Updated Safety Analysis Report (USAR) for the Wolf Creek Generating Station. The licensee's proposed changes would incorporate a revised alternate shutdown methodology into the USAR. The NRC staff requests the following additional information in order to complete its review of the request. Reactor Systems Branch Questions: SRXB-RAI-1 In Section 3.7.5 of the submittal, it states that maximum operator response times were, when possible, set as less than or equal to 80 percent of the time-sensitive action required time. This section also claims that the 80 percent threshold partially accounts for instrumentation uncertainties. The effects of time-sensitive human actions are important to determining the acceptability of the presented sequences. Please provide: a. A list of time-sensitive actions in Evaluation SA-08-006 which were identified by procedure AL 21-017 to require between 80 percent and 100 percent of the four time-sensitive action required time to complete. b. A list of conservative assumptions purposefully included in evaluation SA-08-006 to account for uncertainties in plant response. In addition, c. Please explain if the 80 percent threshold also is meant to account for delays due to operator errors. If yes, explain how the 80 percent threshold was developed. The NRC staff normally accounts for operator error by postulating credible errors for specific tasks and adding the time margin required to recover from the worst case credible error. If a default value is to be used instead of a task-specific margin, the time margin should be 100 percent (i.e., the estimated time required should be doubled; see Appendix B of NUREG-1852, "Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire," October 2007 (ADAMS Accession No. ML073020676), for details). If the margin Enclosure SRXB-RAI-2 of 1 00 percent is used, the NRC staff would expect your threshold for further action to be set at 50 percent, not 80 percent. Please provide an explanation of the intended purpose of Figures 3 and 5 of the submittal; as presented, the differences between the plots of the individual sequences are indistinguishable. SRXB-RAI-3 The discussion of the applicability of the Chexai-Lellouche correlation is difficult to follow in its current form. Please provide a clear explanation of the following: a. The parameter ranges in Table 3 of Attachment 1 extend outside the Lellouche experimental database. Please explain what the plant conditions are in both the primary and secondary systems during which the Chexai-Lellouche correlation is applied to the RETRAN-30 calculation. b. The licensee stated that the Chexai-Lellouche correlation could be used to determine when and if binding occurs in the steam generator tubes, but the evaluations in SA-08-006 do not predict significant steam accumulation in the steam generator tubes. Please clarify when and where the Chexai-Lellouche correlation is or is not used in each of the 24 scenarios. c. The analysis of the submittal's sequences 1, 1A, and 1 C show that voids develop in the core regions and then move to the steam generators, where the voids collapse in the first steam generator tube volume. The lowest void fraction in the steam generator tubes is reported as 0.0 in Table 3 of Attachment 1. Please explain whether the void fraction reported in Table 3 is a rounded value, or whether the voids collapse in the steam generators so quickly that no voids exist at the bottom of the steam generator. d. Please explain the purpose of applying the Chexai-Lellouche to the primary system calculations, if no voids exist in the bottom of the steam generator tubes. SRXB-RAI-4 In Section 3.5 of the application, the licensee states, in part, that, "there is reasonable assurance that the fire will remain in the cabinet of origin and will not spread." This conclusion has important implications as to the acceptability of the selected sequences in the supporting analysis. If the fire were to spread beyond the cabinet of origin, the impact upon systems could be significantly different than that of a single cabinet fire with significant implications for the thermal-hydraulic reasons to the accident scenarios. Thus, validation of the submittal conclusion that a fire would not spread beyond the cabinet of origin is important. Please explain or illustrate whether or not the electrical cabinet separation requirements for adjacent cabinets, that resulted from the tests documented in NUREG/CR-4527, "An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets," April1987 (ADAMS Accession No. ML060960351 ), are consistent with the actual main control room cabinets at the site (e.g., the separation between S8038 and S8037 cannot be compared to the requirements established in NUREG/CR-4527). SRXB-RAI-5 Multiple spurious auctions are important in the progression of accident sequences. Please provide the following: a. A discussion regarding how single or multiple spurious actuations have been considered in each of the accident scenarios, consistent with the requirements in Regulatory Guide 1.189, Revision 2, "Fire Protection for Nuclear Power Plants," October 2009 (ADAMS Accession No. ML092580550), Sections 5.4.1 and 5.4.4, which specify that such actuations should be considered after control has been transferred from the control room to the alternative or dedicated shutdown system and after control of the plant has been achieved. b. A discussion regarding the consideration of spurious actuations that could defeat the alternate safe shutdown system (e.g., spurious actuations that could negate the successful isolation of the main feedwater system or the chemical injection system). This information will be used by the NRC staff to validate the input assumption used in the supporting analysis. SRXB-RAI-6 Please explain how fuel thermal conductivity degradation is accounted for in the analysis. SRXB-RAI-7 Please justify the core axial power shape that is used in the analysis. SRXB-RAI-8 Please explain if loss of feedwater without offsite power may cause the pressurizer to overfill. SRXB-RAI-9 Please list the operator actions that will be taken in the control room prior to evacuation due to a fire, and identify how these operator actions are incorporated into the analysis used in support of this request. Fire Protection Branch Questions: The amendment request included changes from the licensee's commitments to certain technical requirements of Title 10 of the Code of Federal Regulations Part 50, Appendix R,Section III.L, as documented in the WCGS Fire Hazards Analysis Report. The commitments are related to crediting the automatic feedwater isolation system for closure of the main feedwater isolation valves (MFIVs) and/or the main feedwater regulating valves (MFRVs) and the MFRV bypass valves to terminate main feedwater flow and prevent steam generator overfill in the event that a fire necessitates the evacuation of the control room. FP-RAI-01 Operator Manual Actions Table 7.1 within E-1F9915, "Design Control Document ForOFN RP-017, Control Room Evacuation," provides a detailed evaluation for each step in OFN RP-017. For the required actions from outside the control room, OFN RP-017 does not specify whether the operator manual actions have been evaluated for feasibility and reliability, e.g., per NUREG-1852 "Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire (NUREG-1852)," October 2007 (ADAMS Accession No. ML073020676). Provide a discussion of basis for the feasibility and reliability of operator manual actions performed outside of the control room. FP-RAI-02 Multiple Spurious Actuation Drawing E-1 F9915, Section 2.2, "Assumptions," item 2.2.3 states: Prior to transfer of control to the Auxiliary Shutdown System only a single spurious actuation is assumed to occur at a time, except in the case of two redundant valves in a high/low pressure interface line. All potential spurious actuations are mitigated/prevented using OFN RP-017 but timing is based on the spurious actuations occurring one at a time, or two at a time in the case of high/low pressure interface lines. Regulatory Guide 1.189 "Fire Protection for Nuclear Power Plants," Revision 2, Section 5.4.4 states in part: After control of the plant is achieved by the alternative or dedicated shutdown system, single or multiple spurious actuations that could occur in the fire-affected area should be considered, in accordance with the plant's approved FPP. Please justify the reasoning for assuming that after control of the plant is achieved from the alternative location that the timing of spurious actuations should be based on the spurious actuations occurring one at a time for non-high/low pressure interface lines.
- ML 14323A574 .. *email dated DORULPL4-DORLILPL4-DRA/AFPB/BC DSS/SRXB/8 DORULPL4-DORL/LPL4-OFFICE 1/PM 1/LA c 1/BC(A) 1/PM NAME Flyon JBurkhardt* AKiein* CJackson EOesterle Flyon DATE 12/04/2014 11/20/2014 11/19/2014 12/04/2014 12/04/2014 12/04/2014