ML20303A178
| ML20303A178 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 09/29/2020 |
| From: | Gayheart C Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20303A202 | List: |
| References | |
| NL-20-1084 | |
| Download: ML20303A178 (10) | |
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A Southern Nuclear Cheryl A. Gayheart Regulatory Affairs Director 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 cagayhea@southemco.com September 29, 2020 Docket Nos.: 50-321 NL-20-1084 50-366 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Edwin I. Hatch Nuclear Plant - Units 1 & 2 Revision 38 to the Updated Final Safety Analysis Report, Fire Hazard Analysis Changes, Technical Specification Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54.37(b) Changes, 10 CFR 50.59 Summary Report, and Revised NRG Commitments Report Ladies and Gentlemen:
In accordance with 10 CFR 50.4(b) and 50.71(e), Southern Nuclear Operating Company (SNC) hereby submits Revision 38 to the Edwin I. Hatch Nuclear Plant Units 1 and 2 (HNP) Updated Final Safety Analysis Report (UFSAR). The revised HNP Units 1 and 2 UFSAR pages, indicated as Revision 38, reflect changes through August 31, 2020.
The HNP Units 1 and 2 Technical Specifications, Section 5.5.11, "Technical Specifications (TS)
Bases Control Program," provides for changes to the Bases without prior NRG approval. In addition, TS Section 5.5.11 requires that Bases changes made without prior NRG approval be provided to the NRG on .a frequency consistent with 10 CFR 50.71 (e). Pursuant to TS 5.5.11, SNC hereby submits a complete copy of the HNP TS Bases. The revised HNP TS Bases pages, indicated as Revision 112 for Unit 1 and Revision 124 for Unit 2, reflect changes to the TS Bases through August 31, 2020.
In accordance with Regulatory Issue Summary (RIS) 2001-05, "Guidance on Submitting Documents to the NRG by Electronic Information Exchange or on CD-ROM," all of the current pages of the HNP UFSAR, the HNP UFSAR reference drawings, the TS Bases, the Technical Requirements Manual (TRM), and the Fire Hazard Analysis (FHA) are being submitted on CD-ROM in portable document format (PDF). The revised HNP TRM pages, indicated as Revision 117 for Unit 1 and Revision 123 for Unit 2, reflect changes to the TRM through August 31, 2020. The revised HNP FHA, indicated as Revision 38, also reflects changes through August 31, 2020.
U.S. Nuclear Regulatory Commission NL-20-1084 Page2 In accordance with 10 CFR 50.59(d)(2), SNC hereby submits the 10 CFR 50.59 Summary Report containing a brief description of any changes, tests, or experiments, including a summary of the safety evaluation of each.
In accordance with NEI 99-04, "Guidelines for Managing NRC Commitment Changes,"
Revision 0, SNC reviewed its Commitment Database and identified no commitment changes for the applicable reporting period (September 1, 2019 to August 31, 2020).
SNC conducted a review of HNP plant changes for 10 CFR 54.37(b) applicability and identified no components that were determined to meet the criteria for newly identified components as clarified by RIS 2007-16, Revision 1, "Implementation of the Requirements of 10 CFR 54.37(b) for Holders of Renewed Licenses." provides a table of contents with associated file names for the set of two CD-ROMs (Enclosure 2). Enclosure 3 provides the 10 CFR 50.59 Summary Report.
This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at (205) 992-6611.
I declare under penalty of perjury that the foregoing is true and correct. Executed on this 29th day of September 2020.
Respectfully submitted, Cheryl Gay Regulatory Affairs Director CAG/tle
Enclosures:
- 1. CD-ROM Table of Contents
- 2. CD-ROMs (2 discs)
- 3. 10 CFR 50.59 Summary Report cc: Regional Administrator, Region II (w/o enclosures)
Senior NRR Project Manager - Hatch (w/o enclosures)
Senior Resident Inspector - Hatch (w/o enclosures)
RType: CHA02.004
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NL-20-1084 Edwin I. Hatch Nuclear Plant - Units 1 & 2 Revision 38 to the Updated Final Safety Analysis Report, Fire Hazard Analysis Changes, Technical Specification Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54.37(b) Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Enclosure 1 CD-ROM Table of Contents to NL-20-1084 CD-ROM Table of Contents I SEQ CONTENT FILENAME EXTENSION DISC 1 NRC File Nomenclature .doc 001 HATCH FSAR Ul .pdf UNIT 1 Active Page List Table of Contents Chapters 1 thru 14 Appendices A thru K, M, N & R 002 HATCH FSAR_U2_APL, TOC, CHl THRU CH4 .pdf UNIT2 Active Page List Table of Contents Chapters 1 thru 4 003 HATCH FSAR U2 CHS THRU CH7 .pdf UNIT 2 Chapters 5 thru 7 004 HATCH FSAR U2 CH8 THRU CH 18, APP A .pdf UNIT 2 Chapters 8 thru 18 Appendix A 005 HATCH BASES .pdf Units 1 and 2 Technical Specifications Bases 006 HATCH TRM UNIT 1 PT 1 .pdf 007 HATCH TRM UNIT 1 PT 2 .pdf 008 HATCH TRM UNIT 2 .pdf 009 HNP FHA .pdf to NL-20-1084 CD-ROM Table of Contents I FILENAME EXTENSION SEQ CONTENT DISC 2 010 HATCH FSAR REF DWGS PT 1 .pdf A-21603 - H-11606 011 HATCH FSAR REF DWGS PT 2 .pdf H-11607 - H-16002 012 HATCH FSAR REF DWGS PT 3 .pdf H-16003 - H-16174 013 HATCH FSAR REF DWGS PT 4 .pdf H-16176 - H-16339 014 HATCH FSAR REF DWGS PT 5 .pdf H-16512 - H-19941 015 HATCH FSAR REF DWGS PT 6 .pdf H-19942 - H-21114 016 HATCH FSAR REF DWGS PT 7 .pdf H-22250 - H-24748 017 HATCH FSAR REF DWGS PT 8 .pdf H-24749 - H-26036 018 HATCH FSAR REF DWGS PT 9 .pdf H-26037 - H-26102 019 HATCH FSAR REF DWGS PT 10 .pdf H-26103 - S-15290 020 HATCH FSAR REF DWGS PT 11 .pdf S-15304 - S-40969 021 HATCH FSAR REF DWGS PT 12 .pdf S-53448 - S-56429 Part 1 022 HATCH FSAR REF DWGS PT 13 .pdf S-56429 Part 2 - SX-28760
NL-20-1084 Edwin I. Hatch Nuclear Plant - Units 1 & 2 Revision 38 to the Updated Final Safety Analysis Report, Fire Hazard Analysis Changes, Technical Specification Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54.37(b) Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Enclosure 2 CD-ROMs (2 discs)
NL-20-1084 Edwin I. Hatch Nuclear Plant - Units 1 & 2 Revision 38 to the Updated Final Safety Analysis Report, Fire Hazard Analysis Changes, Technical Specification Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54.37(b) Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Enclosure 3 10 CFR 50.59 Summary Report to NL-20-1084 10 CFR 50.59 Summary Report 10 CFR 50.59 Summary Report Activity: SNC1073880TCC
Title:
2T48F081 Maintained Closed for Primary Containment Isolation 10 CFR 50.59 Evaluation Summary:
The proposed activity is the following compensatory measures put in place due to excessive leakage through the seat area of the 2T48F319 and 2T48F320 primary containment isolation valves (PCIVs):
- Closing and gagging closed valve 2T48F081 (which is normally open/ fail open)
- Removing power to valves 2T48F081, 2T48F083 and 2T48F082 As a result of these compensatory measures, the following aspects of the activity were evaluated in this 50.59:
- 1. Because the 2T48F082 penetration flow path releases directly to the main stack, the potential for additional secondary containment bypass leakage exists. Previously, leakage past the PCIVs was filtered through the standby gas treatment systems through the 2T48F081 penetration flow path. With this flow path being closed, there will be additional secondary containment bypass leakage since leakage past the PCIVs will not be filtered by standby gas through the 2T48F081 penetration flow path.
- 2. Primary containment pressure Will be experienced at valves 2T48F081, 2T48F083 and 2T48F082 and the associated upstream piping. In addition, non-safety instruments 2T48N018A, 2T48N019 and 2D11 N011 within this upstream piping will also experience primary containment pressure.
These compensatory measures were taken to address the degraded or inoperable PCIVs and primary containment, resulting in restoring them to an Operable But Degraded status. A review of the FSAR accidents and events reveals that these compensatory measures would not result in an increase of occurrence of any accident or event discussed in the FSAR or increase their consequences. The compensatory measures will not result in an increase in the likelihood of a malfunction of an SSC important to safety or increase their consequences since 2T48F082, 2T48F081 and 2T48F083 are closed and deactivated with no credible single-failure. Primary containment pressure and temperature will now be experienced at instruments 2T48N018A, 2T48N019 and 2D11 N011. Evaluations have determined that the inclusion of these instruments within primary containment pressures and temperatures would have no adverse impact on the extended boundary. All piping within the area exposed to primary containment pressure was designed and installed per ASME Section 111, Class II, Seismic I requirements, which meet the same requirements for containment isolation piping and valves. Implementation of this compensatory measure creates additional Secondary Containment bypass leakage source. For this reason, there is an impact to containment. However, it has been shown that no leakage limits are exceeded and dose remains within regulatory limits during postulated accidents.
Additionally, these compensatory measures do not create the possibility of an accident beyond those already analyzed in the FSAR.
to NL-20-1084 10 CFR 50.59 Summary Report Activity: RER SNC991082-20
Title:
Hatch-1 Cycle 30 Reload 50.59 10 CFR 50.59 Evaluation Summary:
The Hatch-1 core was reloaded during the February 2020 outage into a new core configuration.
The core loading includes 5 Lead Test Assemblies (LTAs) previously irradiated in Cycle 29, four GNF2 ARMOR LTAs and one GNF2 lronClad LTA. One of the ARMOR LTAs is reconstituted during the refueling outage. The four LTRs in the reconstituted ARMOR LTA are replaced with solid (unfueled) Zircaloy rods. The purpose of both LTA programs is to develop specimens for post-irradiation examination (PIE). PIE will characterize material properties and performance, support alloy and coating optimization, and fuel design for potential reload applications.
The 50.59 evaluation was conservatively performed because of the potential adverse impact of the reconstitution of the ARMOR LTA, which reduced fuel inventory, on the fuel and reactor core design function. This evaluation relied upon the prior evaluation for Hatch-1 Cycle 29 for the loading of the ARMOR and Ironclad LTAs and was reviewed for applicability to the ARMOR reconstitution.
Reload Licensing Analyses specific to Hatch-1 Cycle 30, and specific evaluations of the ARMOR LTAs were sufficient to show that the licensing analyses performed with standard GNF2 assemblies remains valid and applicable with the loading of the reconstituted ARMOR LTA. Additionally, the loading of the reconstituted ARMOR LTA in a non-limiting core location and operation in compliance with the cycle-specific COLR is sufficient to ensure that margin to the Specified Acceptable Fuel Design Limits is maintained, as required by the reactor core design basis.
Therefore, the responses to Questions 1-7 all result in "NO" answers. Question 8 was not addressed because NRG-approved methods were used to ensure compliance with the SAFDLs.
Activity: DCP SNC489858
Title:
Degraded Grid - U1 Final Configuration 10 CFR 50.59 Evaluation Summary:
This activity modified the requirements of the TS Bases for LCO 3.8.1, "AC Sources -
Operating" with regard to the qualified offsite sources. The new minimum configuration to meet this LCO requirement is two SATs feeding one ESF bus, and the remaining two ESF buses each being fed by a different SAT. This will ensure that in the event of the loss of one offsite circuit, a minimum of two ESF buses will be available to perform the safety function. This is similar to the existing requirements, except that the 2F bus is no longer required to be supplied by two qualified offsite circuits, provided either the 2E bus or 2G bus is supplied by two qualified offsite circuits.
to NL-20-1084 10 CFR 50.59 Summary Report Activity: OCP SNC826498
Title:
XL3 Fire Panel Replacements 10 CFR 50.59 Evaluation Summary:
The activity replaces the existing XL3 slave panels, master panel, CXL panel, associated printers and workstation, audible notification appliances, and addressable field devices with new MXL slave panels, a OMS workstation and printer, audible notification appliances, and new addressable field devices, respectively.
The new Siemens system replicates the existing Pytrotronics system functions. The replacement fire protection system will be operated in the same manner as the existing system.
The existing fire protection system has automatic control functions discussed in the Updated FSAR regarding the actuation of solenoid valves for the pre-action and deluge systems. The activity does not change the existing automatic control functions for the existing fire protection system.
The activity was screened in and a 50.59 evaluation was performed because of the digital equipment being added to the fire protection system. No other potentially adverse impacts of the proposed activity were identified during the 50.59 screening.
The fire protection system is a non-safety related system that is described in the fire hazards analysis, which is referred to in the Updated FSAR. The activity does not change the functions of the fire protection system as described in the Updated FSAR.
A FMEA identified failure modes related to the digital upgrades that would cause degraded or partial performance of the MXL slave panels, OMS workstation, printer, TRI-B6R modules, and audible notification appliances. All common failures within the fire protection system result in trouble and/or supervisory alarms generated at both the MXL slave panel and the OMS workstation in Main Control Room, and in some cases, compensatory measures such as fire watches being performed until the system functionality was restored.
The results of the FMEA show the failures associated with the circuit supervision system are typical and that the system has few inherent risks associated with it. The class B devices conform to NFPA 72-2016, which requires sufficient self-diagnostics to minimize system failures and identify degraded components. The system is able to provide the necessary protection during a fire event even under abnormal conditions. The system will also alert the operator when there are degraded components or parts of the system so needed repairs can be initiated.
Given the FMEA results, it was concluded that the activity will not initiate accidents or malfunctions described in the Updated FSAR, 2) does not increase the consequences of an accident or malfunction previously evaluated in the Updated FSAR, 3) does not create the possibility for an accident of a different type or malfunction with a different result than previously evaluated in the Updated FSAR, and 4) does not impact the impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment. Therefore, activity does not require a license amendment to be submitted and approved by the NRG prior to implementation of the activity.