NL-24-0014, Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical.
| ML24030A895 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/30/2024 |
| From: | Coleman J Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-24-0014 | |
| Download: ML24030A895 (1) | |
Text
3535 Colonnade Parkway Birmingham, AL 35243 205 992 5000 tel
January 30, 2024
Docket Nos.: 50-321 NL-24-0014 50-366 10 CFR 50.90
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001
Edwin I. Hatch Nuclear Plant - Units 1 and 2 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, MODES
Ladies and Gentlemen:
On August 19, 2022, pursuant to the provisions of 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, Southern Nuclear Operating Company (SNC) requested a license amendment to the Technical Specifications (TS) for Edwin I. Hatch Nuclear Plant (HNP) Units 1 and 2 Renewed Facility Oper ating Licenses (RFOLs) DPR-57 and NPF-5, respectively [ML22231B055]. The proposed amendment would revise TS Table 1.1-1, MODES, to relax the required number of fully tensioned reactor pressure vessel (RPV) closure studs.
On September 20, 2023, NRC staff sent a second round of requests for additional information (RAIs) [ML23250A047]. SNC provided a response to the second round of RAIs on October 20, 2023 [ML23293A235], and the responses were discussed during public meetings on November 28, 2023 [ML23338A259] and January 8, 2024 [ML23361A124]. SNC has rewritten the October 20, 2023 RAI responses to address NRC staff feedback received during those public meetings.
The enclosed RAI response and attachments to this letter supersede those provided in SNCs letter dated October 20, 2023 [ML23293A235]. The response to RAI-03 resulted in changes to the responses to Standards 1 and 2 of the No Significant Hazards Consideration Determination Analysis. The changes to the responses to Standards 1 and 2 do not affect the conclusions of the No Significant Hazards Consideration Determination Analysis. Attachment 1 to the enclosure provides proposed markups of HNP Unit 1 and HNP Unit 2 TS Table 1.1-1 which supersede the HNP Unit 1 and Unit 2 TS markups provided in the original application [ML22231B055]. to the enclosure provides revised (clean typed) TS pages which supersede the HNP Unit 1 and HNP Unit 2 revised TS pages in the original application [ML22231B055]. Attachment 3 to the enclosure provides a revised response to No Significant Hazards Consideration Determination Analysis Standards 1 and 2 and supersedes pages E1-11 and E1-12 of to the original application [ML22231B055]. Attachment 4 provides marked-up pages of the HNP Unit 1 RFOL DPR-57 and HNP Unit 2 RFOL NPF-5 showing the new license U.S. Nuclear Regulatory Commission NL-24-0014 Page 2
conditions. Attachment 5 provides clean typed versions of the HNP Unit 1 and HNP Unit 2 RFOL pages with the new license conditions.
Additionally, SNC hereby supers edes the implementation period communicated in its letter dated August 19, 2022. The amendment, once approved, will be implemented within 60 days of issuance.
In accordance with 10 CFR 50.91, Notice for public comment; State consultation, paragraph (b)(1), a copy of this application, with enclosure and attachments, is being provided to the designated Georgia Officials.
This letter contains no regulatory commitments. If you have any questions regarding this submittal, please contact Ryan Joyce at 205.992.6468.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 30 th day of January 2024.
Respectfully submitted,
Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Company
JMC/agq/cg
Enclosure:
Revised Response to Request for Additional Information
Attachments: 1) HNP Units 1 and 2 Technical Specifications Marked-up Pages
- 2) HNP Units 1 and 2 Technical Specifications Revised Pages (clean typed)
- 3) Revised No Significant Hazards Consideration Determination Analysis
cc: NRC Regional Administrator, Region II NRC NRR Project Manager - Hatch NRC Senior Resident Inspector - Hatch Director, Environmental Protection Division - State of Georgia RType: CHA02.004 Edwin I. Hatch Nuclear Plant - Units 1 and 2 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, MODES
Enclosure
Response to Request for Additional Information
NL-24-0014 Enclosure Revised Response to Request for Additional Information
INTRODUCTION
On August 19, 2022, pursuant to the provisions of 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, Southern Nuclear Operating Company (SNC) requested a license amendment (Reference 1) to the Technical Specifications (TS) for Edwin I.
Hatch Nuclear Plant (HNP) Units 1 and 2 Renewed Facility Operating Licenses DPR-57 and NPF-5, respectively. The proposed amendment w ould revise TS Table 1.1-1, MODES, to relax the required number of fully tensioned reactor pressure vessel (RPV) closure studs.
On January 20, 2023, SNC provided a response to a request for additional information (RAI)
(Reference 2). On March 28, 2023 and May 2, 2023, public meetings were held to discuss a number of topics regarding the license amendment request. On September 20, 2023, NRC staff sent a second round of RAIs. The RAIs are copied below, and SNCs response is provided following each RAI.
TS Table 1.1-1 Footnotes (a) and (b) specify the reactor vessel head closure bolt requirements.
For the purposes of this RAI response, reactor vessel head closure bolts as specified in the TS, are equivalent to reactor pressure vessel head closure studs or studs.
REQUEST FOR ADDITIONAL INFORMATION (RAI)
Title 10 of the Code of Federal Regulations (10 CFR) Part 50.36, Technical Specifications, specifies the content required to be included in TSs.
Regulations in 10 CFR Part 50, Domestic Lic ensing of Production and Utilization Facilities, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, includes the following GDCs applicable to the licensees LAR:
Criterion 1 (GDC 1), Quality standards and records, requires, in part, that structures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed.
Criterion 14 (GDC 14), Reactor coolant pressure boundary, requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary so as to have an extremely low probability of abnormal leakage, or rapidly propagating failure, and of gross rupture.
Criterion 30 (GDC 30), Quality of reactor coolant pressure boundary, which requires in part, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical.
Hatch Unit 1 was licensed to the applicable Atomic Energy Commission (AEC) preliminary general design criteria identified in the Federal Register 32 FR 10213, published July 11, 1967 (ML043310029). The Hatch Unit 1 Updated Final Safety Analysis Report (UFSAR), Appendix F, Conformance to the AEC Criteria (ML20303A202), contains the comparison of the applicable AEC criteria to the 10 CFR 50, Appendix A, GDCs.
2 of 7 NL-24-0014 Enclosure Revised Response to Request for Additional Information
RAI-01
The Hatch submittal uses the term out of serv ice extensively in describing the proposed TS changes for RPV bolts.
- 1) Please provide an explanation of SNCs proposed use of out of service in the TS as compared to the current TS OPERABLE/OPERABILITY definition.
- 2) If a bolt is less than fully tensioned, is it considered OPERABLE, and if so, please justify?
SNC Response to RAI-01
- 1) SNC does not propose to use the term out of service in the Technical Specifications (TS) or the renewed facility operating licenses (RFOLs). However, SNC used the term out of service throughout the August 19, 2022 license amendment request for a number of reasons including:
For consistency with the evaluation performed by Dominion Engineering, Inc.
(DEI), which provided the technical justification for the change and used the term out of service extensively.
The precedent license amendment requests and safety evaluations cited in SNCs submittal dated August 19, 2022 (Reference 1) did not use the terms operable, operability, inoperable, or inoperability. Also, note that following SNCs submittal, an amendment for a change to Watts Bar TS Table 1.1-1 was issued (Reference 4), and the associated safety evaluation used the term out of service.
The terms operable and operability are not defined in the regulations related to commercial nuclear power. The terms are defined in the TS for each plant and, therefore, only have meaning as used relative to that document. The scope of structures, systems, and components (SSCs) considered within the operability determination process is limited to those SSCs that are required to be operable by a TS limiting condition for operation (LCO). An SSC that is not directly addressed by a TS LCO may perform a necessary support function for an SSC that is directly required by a TS LCO.
The SSC providing the support function is itself not subject to operability determination process, but its capabilities are a consideration in determining operability of the supported SSC that is directly required by a TS LCO (Reference 3).
HNP Unit 1 and Unit 2 TS Section 1.1, Definitions, includes a definition for MODE. The definition of MODE states: A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. HNP Unit 1 and Unit 2 TS Sections 3.1 through 3.10 identify the LCOs that must be met. The RPV closure studs are not SSCs explicitly addressed by an LCO in Sections 3.1 through 3.10. However, the RPV closure studs do provide a support function in that they provide compressive force on the flanges and the two concentric metal seal rings that maintain
3 of 7 NL-24-0014 Enclosure Revised Response to Request for Additional Information
the leak tightness of the RPV flange. Leak tightness of the RPV flange is part of the overall reactor coolant pressure boundary (RCPB), which is necessary for other LCOs to be met. TS LCO 3.4.1, Recirculation Loops Operating and TS LCO 3.4.4, RCS Operational Leakage are examples of LCOs which rely upon an intact RCPB.
To provide assurance that the above support function is performed and supported LCOs are met, the RPV closure studs are examined for structural integrity in accordance ASME Section XI requirements, installed, and fully tensioned. The proposed license conditions (see Attachments 4 and 5) would provide additional assurance that any RPV closure stud credited as being fully tensioned meets the applicable ASME Section XI requirements, irrespective of any exis ting applicable ASME Code relief (e.g., Unit 2 stud #33). In addition, the proposed TS Table 1.1-1 changes (see Attachments 1 and 2) would mandate the number of studs required to be installed and fully tensioned.
Together, the TS Table 1.1-1 and license condition provide assurance that the assumptions of the supporting analysis pr ovided in the license amendment request dated August 19, 2022 [ML22231B055] are preserved and the RPV closure flange leak tightness is preserved.
- 2) If an RPV closure stud does not meet the proposed license condition, it cannot be credited as fully tensioned. This woul d be analogous to considering the stud INOPERABLE if a TS LCO applied to the studs. However, if the proposed required number of RPV closure studs are fully tensioned and meet the proposed license condition, the overall ability of the required RPV closure studs to maintain the RPV flange leak tight by providing compressive force on the flanges and seal rings is preserved and there would be no adverse effect on meeting the supported LCOs.
RAI-02
In its January 20, 2023, RAI response submittal (ML23020A902), the licensee states that it fully intends to employ sound engineering principles to maintain all RPV bolts in service.
If a flaw is found during a Section XI inspection in an RPV bolt that will not be repaired or replaced, what are the licensees requirements to determine the maximum allowable torque that will be applied during tensioning of that bolt prior to plant startup?
SNC Response to RAI-02
The tensioning of the stud would be performed per normal reactor vessel reassembly procedure knowing that evaluations demonstrate the vessel stresses and leak tightness would be acceptable if that stud were to become untensioned while in service. Additionally, please note that the studs are tensioned to a set of elongation criteria, not torque. Flaws identified through volumetric examination methods cannot be characterized since there are limitations in determining the length and depth of the flaw within the stud; thus, alternative stud tensioning techniques based on known flaw properties cannot be developed and deployed. Lastly, note that with the proposed license condition, even if a stud that does not meet the applicable ASME Section XI acceptance criteria were fully tensioned, that stud would not count towards meeting the number of studs required by TS Table 1.1-1.
4 of 7 NL-24-0014 Enclosure Revised Response to Request for Additional Information
RAI-03
If approved, the latest proposed TS changes provided by the licensee during the public meeting on May 2, 2023 (Summary - ML23136B294), would allow one bolt from each unit to either not be installed or installed and not fully torqued. Please explain how this does not constitute a material alteration of the facility as originally designed and licensed.
Based on the above, please discuss whether the assertion in the No Significant Hazards Consideration that no hardware changes are proposed accurately reflects the proposed change, which would allow operation with up to two RPV head closure bolts not installed or torqued.
SNC Response to RAI-03
In Reference 1, Enclosure 1, Section 4.3, the response to Standard 2 states in part: There are no hardware changes nor are there any changes in the method by which any safety related plant system performs its safety function. Thi s statement was intended to imply that no new or different equipment is being installed as result of the proposed change. Operation with a stud less than fully tensioned could be considered a configuration change, and this configuration change has been analyzed. Additionally, this statement is assessing the RPV overall as an ASME Section III pressure vessel and its function as part of the reactor coolant pressure boundary (i.e., maintain coolable geometry and prevent leakage). As summarized in the response to Standard 2 of the no significant hazards consideration (NSHC) determination, there would be no leakage because adequate compression would remain, and stresses remain within the ASME Code allowable limits. These statement s are backed up by the calculations provided in Reference 1, which analyzed the RPV vessels and closure studs when a stud (Unit 1) is, or studs (Unit 2) are, out of service under design basis conditions. In Attachment 3, SNC provides revised responses to NSHC Standards 1 and 2 as part of this RAI response; the pages provided in Attachment 3 supersede those respective pages provided in the license amendment request dated August 19, 2022. A marked-up version of the responses is also provided below.
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
Overall protection system performance will remain within the bounds of the accident analyses, since no hardware changes are proposed. Since the stresses remain within ASME Code allowables, the proposed change will not affect the probability of any event initiators nor will the proposed change affect the ability of any safety related equipment to perform its intended function. There will be no degradation in the performance of nor an increase in the number of challenges imposed on safety related equipment assumed to function during an accident situation.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
5 of 7 NL-24-0014 Enclosure Revised Response to Request for Additional Information
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Plant specific calculations indicate that the HNP Unit 1 reactor pressure vessel (RPV) will continue to meet ASME Code allowable stress criteria with one RPV closure stud less than fully tensioned, and the HNP Unit 2 RPV will continue to meet ASME Code allowable stress criteria with two RPV closure studs less than fully tensioned provided there are at least nine fully tensioned studs between the two. There are no hardware changes nor are there any changes in the method by which any safety related plant system performs its safety function. The method of plant operation is unaffected.
Leakage would be precluded since adequate compression remains. Analysis demonstrates that any gap opening remains less than the springback recovery of the inner closure o-ring. Since stresses remain within ASME Code allowables, no new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
RAI-04
During the NRCs May 2, 2023, public meeting, SNC proposed a revision to Unit 2 TS Table 1.1-1 footnotes along with a proposed Unit 2 license condition. Please provide any updated proposed revisions to TS Table 1.1-1 al ong with SNCs proposed license condition(s) for each unit.
SNC Response to RAI-04
SNC is proposing the revisions to Table 1.1-1 of the Unit 1 and Unit 2 TS provided in (markups) and Attachment 2 (clean typed). Additionally, SNC is proposing the new license conditions provided in Attachment 4 (markups) and Attachment 5 (clean typed).
RAI-05
During the NRCs May 2, 2023, public meeting, SNC proposed a revision to Unit 2 TS Table 1.1-1 footnotes along with a proposed Unit 2 license condition.
Please discuss licensee actions that would be taken if more than one reactor head closure bolt cannot be properly examined in accordance with IWB-3515.2(c) of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI.
SNC Response to RAI-05
The proposed Unit 2 TS changes provide a clear set of criteria to apply with respect to the TS Mode definitions, and the associated license conditions provide additional assurance that the
6 of 7 NL-24-0014 Enclosure Revised Response to Request for Additional Information
RPV closure studs will be tensioned within the bounds of the supporting calculation included with the license amendment request and will only be credited as being fully tensioned if they meet the applicable ASME Section XI requirements. Thus, if there were a second RPV closure stud that could not be properly examined in accordance with IWB-3515.2(c) of ASME Section XI, SNCs options would include:
- 1) Repair or replace this second RPV closure stud.
- 2) Assuming the license conditions are met, execute a design change to remove this second stud from the RPV design, thereby removing it from the scope of ASME Section XI.
REFERENCES
- 1. Letter from C. A. Gayheart (SNC) to the Document Control Desk (NRC), Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, M ODES, dated August 19, 2022 (NRC ADAMS Accession No. ML22231B055).
- 2. Letter from C. A. Gayheart (SNC) to the Document Control Desk (NRC), Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, M ODES, dated August 19, 2022 (NRC ADAMS Accession No. ML23020A902).
- 3. Nuclear Energy Institute (NEI) 18-03, Operability Determination, October 2019 (ADAMS Accession No. ML19284C872).
- 4. Letter from P. H. Buckberg to J. Barstow, Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment No. 161 Regarding a One-Time Use Change to Footnotes Applicable to Technical Specification Table 1.1-1 MODES (Emergency Circumstances), dated May 5, 2023 (NRC ADAMS Accession Number No. ML23125A220).
7 of 7 Edwin I. Hatch Nuclear Plant - Units 1 and 2 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, MODES
Attachment 1
HNP Units 1 and 2 Technical Specifications Marked-up Pages Definitions 1.1
Table 1.1-1 (page 1 of 1)
MODES
AVERAGE REACTOR COOLANT MODE TITLE REACTOR MODE TEMPERATURE SWITCH POSITION (°F)
1 Power Operation Run NA
2 Startup Refuel(a) or Startup/Hot NA Standby
3 Hot Shutdown(a) Shutdown > 212
4 Cold Shutdown(a) Shutdown 212
5 Refueling(b) Shutdown or Refuel NA
(a) AllAt least 51 reactor vessel head closure bolts fully tensioned, subject to the limitations of license condition 2.(C)(12).
(b) One or more reactor vessel head closure bolts less Fewer than 51 reactor vessel head closure bolts fully tensioned.
HATCH UNIT 1 1.1-7 Amendment No.
Definitions 1.1
Table 1.1-1 (page 1 of 1)
MODES
AVERAGE REACTOR COOLANT MODE TITLE REACTOR MODE TEMPERATURE SWITCH POSITION (°F)
1 Power Operation Run NA
2 Startup Refuel(a) or Startup/Hot NA Standby
3 Hot Shutdown(a) Shutdown > 212
4 Cold Shutdown(a) Shutdown 212
5 Refueling(b) Shutdown or Refuel NA
(a) All At least 54 reactor vessel head closure bolts fully tensioned, subject to the limitations of license condition 2.(C)(3)(j).
(b) One or more reactor vessel head closure bolts less Fewer than 54 reactor vessel head closure bolts fully tensioned.
HATCH UNIT 2 1.1-8 Amendment No.
Edwin I. Hatch Nuclear Plant - Units 1 and 2 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, MODES
Attachment 2
HNP Units 1 and 2 Technical Specifications Revised Pages (clean typed)
Definitions 1.1
Table 1.1-1 (page 1 of 1)
MODES
AVERAGE REACTOR COOLANT MODE TITLE REACTOR MODE TEMPERATURE SWITCH POSITION (°F)
1 Power Operation Run NA
2 Startup Refuel(a) or Startup/Hot NA Standby
3 Hot Shutdown(a) Shutdown > 212
4 Cold Shutdown(a) Shutdown 212
5 Refueling(b) Shutdown or Refuel NA
(a) At least 51 reactor vessel head closure bolts fully tensioned, subject to the limitations of license condition 2.(C)(12).
(b) Fewer than 51 reactor vessel head closure bolts fully tensioned.
HATCH UNIT 1 1.1-7 Amendment No.
Definitions 1.1
Table 1.1-1 (page 1 of 1)
MODES
AVERAGE REACTOR COOLANT MODE TITLE REACTOR MODE TEMPERATURE SWITCH POSITION (°F)
1 Power Operation Run NA
2 Startup Refuel(a) or Startup/Hot NA Standby
3 Hot Shutdown(a) Shutdown > 212
4 Cold Shutdown(a) Shutdown 212
5 Refueling(b) Shutdown or Refuel NA
(a) At least 54 reactor vessel head closure bolts fully tensioned, subject to the limitations of license condition 2.(C)(3)(j).
(b) Fewer than 54 reactor vessel head closure bolts fully tensioned.
HATCH UNIT 2 1.1-8 Amendment No.
Edwin I. Hatch Nuclear Plant - Units 1 and 2 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, MODES
Attachment 3
Revised No Significant Hazards Consideration Determination Analysis to NL-23-0752 Revised No Significant Hazards Consideration Determination Analysis
4.3 No Significant Hazards Consideration Determination Analysis
Southern Nuclear Operating Company (SNC) is proposing to revise the Technical Specifications (TS) Table 1.1-1, MODES, for Hatch Nuclear Plant (HNP) Units 1 and 2.
The amendment would allow operation of HNP Units 1 and 2 with the required pressure vessel head closure studs fully tensioned. The required number of closure studs, which may be less than the total number, has been established by calculation that demonstrates operation of HNP Units 1 and 2 reactor pressure vessels with the required studs fully tensioned does not result in any component of the reactor pressure vessel closure flange exceeding the design basis ASME Code allowables. The calculation for HNP Unit 1 determined that any one closure stud may be less than fully tensioned. The required closure studs continue to meet ASME Code requirements for primary loads with one stud out of service. The calculation for HNP Unit 2 determined that any two closure studs may be less than fully tensioned as long as the two studs are separated by nine or more studs. The required closure studs continue to meet ASM E Code requirements for primary loads with two studs out of service.
SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
Overall protection system performance will remain within the bounds of the accident analyses. Since the stresses remain within ASME Code allowables, the proposed change will not affect the probability of any event initiators nor will the proposed change affect the ability of any safety related equipment to perform its intended function. There will be no degradation in the performance of nor an increase in the number of challenges imposed on safety related equipment assumed to function during an accident situation.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Plant specific calculations indicate that the HNP Unit 1 reactor pressure vessel (RPV) will continue to meet ASME Code allowable stress criteria with one RPV closure stud less than fully tensioned, and the HNP Unit 2 RPV will continue to meet ASME Code allowable stress criteria with two RPV closure studs less than fully tensioned provided there are at least nine fully tensioned studs between the two. The method of plant operation is unaffected. Leakage would be precluded since adequate compression
E1-11 to NL-23-0752 Revised No Significant Hazards Consideration Determination Analysis
remains. Analysis demonstrates that any gap opening remains less than the springback recovery of the inner closure o-ring. Since stresses remain within ASME Code allowables, no new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change does not affect any Safety Limits or controlling numerical values for a parameter established in the updated final safety analysis report or any specific values that define margin that are established in the plants licensing basis. ASME Section III stress limits for affected components are not exceeded. Plant specific evaluations indicate that the reactor vesse ls will continue to meet ASME Code allowable stress criteria with the required reactor pressure vessel closure studs fully tensioned.
The proposed change does not alter nor exceed the acceptance criteria for any analyzed event. There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
4.4 Conclusions
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions 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.
- 5. ENVIRONMENTAL CONSIDERATION
SNC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, Standards for protection against radiation, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in paragraph (c)(9) of 10 CFR 51.22, Criterion for categorical exclusion, identificati on of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring an environmental review. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact st atement or environmental assessment need be prepared in connection with the proposed amendment.
E1-12 Edwin I. Hatch Nuclear Plant - Units 1 and 2 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, MODES
Attachment 4
HNP Unit 1 and Unit 2 Proposed RFOL Condition Marked-up Pages
- c. The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, from the date of the most recent successful pressure measurement test.
(11) 10 CFR 50.69 Risk-Informed Categorization
Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. 305, dated June 26, 2020.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).
Prior to implementation of the Renewed License Amendment No. 305, dated June 26, 2020, Southern Nuclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as-built, as-operated, and as-maintained plant and shall ensure the risk acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met
(12) Reactor Vessel Head Closure Bolts
Hatch Nuclear Plant Unit 1 is approved to operate in Modes 1 - 4 with at least 51 reactor vessel head closure bolts fully tensioned. In addition, a reactor vessel head closure bolt cannot be considered fully tensioned unless all applicable ASME Section XI acceptance criteria are met (irrespective of any existing NRC approved alternative to, or relief from, the acceptance criteria). Upon implementation of Amendment No., Southern Nuclear Operating Company shall update the Reactor Vessel Reassembly procedure to include this requirement.
D. Southern Nuclear shall not market or broker power or energy from Edwin I. Hatch Nuclear Plant, Unit 1.
- 3. This renewed license is effective as of the date of issuance and shall expire at midnight, August 6, 2034.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION
Samuel J. Collins, Director Office of Nuclear Reactor Regulation Attachments:
Appendix A - Technical Specifications Appendix B - Environmental Protection Plan
Date of Issuance: January 15, 2002
Renewed License No. DPR-57 Amendment No.
(h) TSTF-448 Control Room Habitability
Upon implementation of the Amendments adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.4.4, in accordance with TS 5.5.14.c.(i), the assessment of CRE habitability as required by Specification 5.5.14.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.14.d, shall be considered met. following implementation:
i) The first performance of SR 3.7.4.4, in accordance with Specification 5.5.14.c.(i), shall be within the next 18 months.
ii) The first performance of the periodic assessment of CRE habitability, Specification 5.5.14.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, of the next successful tracer gas test.
iii) The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, from the date of the most recent successful pressure measurement test.
(i) 10 CFR 50.69 Risk-Informed Categorization
Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. 250 dated June 26, 2020.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).
Prior to implementation of the Renewed License Amendment No. 250 dated June 26, 2020, Southern Nuclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as-built, as-operated, and as-maintained plant and shall ensure the risk acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met.
(j) Reactor Vessel Head Closure Bolts
Hatch Nuclear Plant Unit 2 is approved to operate in Modes 1 - 4 with at least 54 reactor vessel head closure bolts fully tensioned. In addition, a reactor vessel head closure bolt cannot be considered fully tensioned unless all applicable ASME Section XI acceptance criteria are met (irrespective of any existing NRC approved alternative to, or relief from, the acceptance criteria). Any bolt that is less than fully tensioned shall have at least nine adjacent bolts on either side that are fully tensioned. Upon implementation of Amendment No., Southern Nuclear Operating Company shall update the Reactor Vessel Reassembly procedure to include this requirement.
Renewed License No. NPF-5 Amendment No.
Edwin I. Hatch Nuclear Plant - Units 1 and 2 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, MODES
Attachment 5
HNP Unit 1 and Unit 2 Proposed RFOL Condition Pages (clean typed)
- c. The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, from the date of the most recent successful pressure measurement test.
(11) 10 CFR 50.69 Risk-Informed Categorization
Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. 305, dated June 26, 2020.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).
Prior to implementation of the Renewed License Amendment No. 305, dated June 26, 2020, Southern Nuclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as-built, as-operated, and as-maintained plant and shall ensure the risk acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met
(12) Reactor Vessel Head Closure Bolts
Hatch Nuclear Plant Unit 1 is approved to operate in Modes 1 - 4 with at least 51 reactor vessel head closure bolts fully tensioned. In addition, a reactor vessel head closure bolt cannot be considered fully tensioned unless all applicable ASME Section XI acceptance criteria are met (irrespective of any existing NRC approved alternative to, or relief from, the acceptance criteria). Upon implementation of Amendment No., Southern Nuclear Operating Company shall update the Reactor Vessel Reassembly procedure to include this requirement.
D. Southern Nuclear shall not market or broker power or energy from Edwin I. Hatch Nuclear Plant, Unit 1.
- 3. This renewed license is effective as of the date of issuance and shall expire at midnight, August 6, 2034.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION
Samuel J. Collins, Director Office of Nuclear Reactor Regulation Attachments:
Appendix A - Technical Specifications Appendix B - Environmental Protection Plan
Date of Issuance: January 15, 2002
Renewed License No. DPR-57 Amendment No.
(h) TSTF-448 Control Room Habitability
Upon implementation of the Amendments adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.4.4, in accordance with TS 5.5.14.c.(i), the assessment of CRE habitability as required by Specification 5.5.14.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.14.d, shall be considered met. following implementation:
i) The first performance of SR 3.7.4.4, in accordance with Specification 5.5.14.c.(i), shall be within the next 18 months.
ii) The first performance of the periodic assessment of CRE habitability, Specification 5.5.14.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, of the next successful tracer gas test.
iii) The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, from the date of the most recent successful pressure measurement test.
(i) 10 CFR 50.69 Risk-Informed Categorization
Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. 250 dated June 26, 2020.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).
Prior to implementation of the Renewed License Amendment No. 250 dated June 26, 2020, Southern Nuclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as-built, as-operated, and as-maintained plant and shall ensure the risk acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met.
(j) Reactor Vessel Head Closure Bolts
Hatch Nuclear Plant Unit 2 is approved to operate in Modes 1 - 4 with at least 54 reactor vessel head closure bolts fully tensioned. In addition, a reactor vessel head closure bolt cannot be considered fully tensioned unless all applicable ASME Section XI acceptance criteria are met (irrespective of any existing NRC approved alternative to, or relief from, the acceptance criteria). Any bolt that is less than fully tensioned shall have at least nine adjacent bolts on either side that are fully tensioned. Upon implementation of Amendment No., Southern Nuclear Operating Company shall update the Reactor Vessel Reassembly procedure to include this requirement.
Renewed License No. NPF-5 Amendment No.