ML22032A243
| ML22032A243 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/07/2022 |
| From: | Stephanie Devlin-Gill Plant Licensing Branch II |
| To: | Gayheart C Southern Nuclear Operating Co |
| Devlin-Gill S, 415-5301 | |
| References | |
| EPID L-2021-LLA-0136 | |
| Download: ML22032A243 (18) | |
Text
March 7, 2022 Ms. Cheryl A. Gayheart Regulatory Affairs Director Southern Nuclear Operating Company 3535 Colonnade Parkway Birmingham, AL 35243
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 241 AND 238 TO ELIMINATE THE ENCAPSULATION VESSELS AROUND THE FIRST CONTAINMENT SPRAY AND RESIDUAL HEAT REMOVAL / LOW HEAD SAFETY INJECTION RECIRCULATION SUCTION ISOLATION VALVES (EPID L-2021-LLA-0136)
Dear Ms. Gayheart:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 241 to Renewed Facility Operating License No. NPF-2 and Amendment No. 238 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively. The amendments are in response to your application dated July 29, 2021, as supplemented by letter dated December 13, 2021.
The amendments modify the Updated Final Safety Analysis Report to allow the removal of the encapsulation vessels and associated guard piping around the first Containment Spray and Residual Heat Removal / Low Head Safety Injection recirculation suction isolation valves.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions Federal Register notice.
Sincerely,
/RA/
Stephanie Devlin-Gill, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosures:
- 1. Amendment No. 241 to NPF-2
- 2. Amendment No. 238 to NPF-8
- 3. Safety Evaluation cc: Listserv
SOUTHERN NUCLEAR OPERATING COMPANY ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 241 Renewed License No. NPF-2
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 1 (the facility), Renewed Facility Operating License No. NPF-2 (the license) filed by Southern Nuclear Operating Company (the licensee), dated July 29, 2021, as supplemented by letter dated December 13, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, by Amendment No. 241, Renewed Facility Operating License No. NPF-2 is hereby amended to authorize the change to the Updated Final Safety Analysis Report (UFSAR) as requested by letter dated July 29, 2021, as supplemented by letter dated December 13, 2021, and evaluated in the NRC staff safety evaluation dated March 7, 2022. The licensee shall submit the update of the UFSAR authorized by this amendment in accordance with 10 CFR 50.71(e) and the licensees exemptions from certain requirements of 10 CFR 50.71(e)(4) (Agencywide Documents Access and Management System Accession Nos. ML013130216 and ML21179A183).
- 3.
This license amendment is effective as of its date of issuance and shall be implemented by December 31, 2024.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: March 7, 2022 Michael T.
Markley Digitally signed by Michael T. Markley Date: 2022.03.07 07:42:40 -05'00'
SOUTHERN NUCLEAR OPERATING COMPANY ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 238 Renewed License No. NPF-8
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 2 (the facility), Renewed Facility Operating License No. NPF-8 (the license) filed by Southern Nuclear Operating Company (the licensee), dated July 29, 2021, as supplemented by letter dated December 13, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, by Amendment No. 238, Renewed Facility Operating License No. NPF-8 is hereby amended to authorize the change to the Updated Final Safety Analysis Report (UFSAR) as requested by letter dated July 29, 2021, as supplemented by letter dated December 13, 2021, and evaluated in the NRC staff safety evaluation dated March 7, 2022. The licensee shall submit the update of the UFSAR authorized by this amendment in accordance with 10 CFR 50.71(e) and the licensees exemptions from certain requirements of 10 CFR 50.71(e)(4) (Agencywide Documents Access and Management System Accession Nos. ML013130216 and ML21179A183).
- 3.
This license amendment is effective as of its date of issuance and shall be implemented by December 31, 2024.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: March 7, 2022 Michael T.
Markley Digitally signed by Michael T. Markley Date: 2022.03.07 07:44:10 -05'00'
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 241 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 238 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364
1.0 INTRODUCTION
By letter dated July 29, 2021, as supplemented by letter dated December 13, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML21210A242 and ML21347A978, respectively), Southern Nuclear Operating Company (SNC, the licensee) requested in a license amendment request (LAR) to make a change, as described in the Updated Final Safety Analysis Report (UFSAR), to allow removal of the encapsulation vessels and associated guard piping around the first isolation valves in the recirculation suction lines for the Containment Spray (CS) and Residual Heat Removal (RHR) / Low Head Safety Injection (LHSI) systems for the Joseph M. Farley Nuclear Plant, Units 1 and 2 (Farley, or FNP). The change proposed by the licensee would remove four encapsulation vessels and associated guard piping at Unit 1 and four encapsulation vessels and associated guard piping at Unit 2.
Each of the eight encapsulation vessels is accessible in pump rooms and are installed on the CS and RHR/LSHI systems first isolation motor-operated gate valves (MOVs) and guard piping.
The supplement dated December 13, 2021, 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) staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on October 5, 2021, 86 FR 55013.
On June 24, 2021, the NRC staff held a public, pre-licensing meeting with the licensee regarding SNCs proposed LAR. A summary of the meeting is located at ADAMS Accession No. ML21176A171.
2.0 REGULATORY EVALUATION
2.1
System Description
There are four containment penetrations per unit that provide a supply of containment sump water to each train of both the RHR/LHSI and CS systems in each unit. Outside the containment boundary, each suction line contains two containment isolation valves in series.
The first of these two valves for each penetration have an encapsulation. These isolation valves are normally closed, motor-operated gate valves that are remotely opened following a loss-of-coolant accident (LOCA) with adequate level in the sump to initiate recirculation.
In Section 2.1 of the Enclosure to its letter dated July 29, 2021, SNC described the encapsulation vessels, in part, as follows:
There are four encapsulation vessels per unit at FNP, one for each of the first isolation valves in the containment sump post-loss-of-coolant accident (LOCA) recirculation path outside of containment. Both trains of the CS and RHR/LHSI systems are designed to function for post-LOCA recirculation to provide containment and core cooling, respectively. The CS and RHR/LHSI equipment are located in the lower levels of the auxiliary building. Each train of these systems is physically separated from the other by watertight compartments. Each system within a train is further separated into watertight compartments to protect from common mode flooding.
The design function of the encapsulation vessels, in combination with guard piping, is to provide a pressure boundary to capture and limit leakage from a passive single failure in the containment emergency sump suction line from underside of the containment liner, up to and including the first sump suction isolation valve in each CS and RHR/LHSI system.
The existing encapsulations are carbon steel vessels connected, through welded expansion joints, to guard piping on the upstream side of the first isolation valve and downstream to the suction piping between the first and second isolation valve. Each encapsulation vessel forms a leak-tight compartment around the first isolation valve.
The encapsulation vessels that the licensee proposes to remove enclose MOVs in each train of the CS and RHR/LHSI systems. These valves perform an active safety function in the open position to align the CS and RHR/LHSI pump suctions to the containment sump for the post-accident recirculation phase. Specifically, the encapsulation vessels surround the following valves:
Unit 1:
o Q1E13MOV8826A o Q1E13MOV8826B o Q1E11MOV8811A o Q1E11MOV8811B Unit 2 o Q2E13MOV8826A o Q2E13MOV8826B o Q2E11MOV8811A o Q2E11MOV8811B Farley UFSAR Section 6.3.2.2.5, Valves, provides the following description of MOVs in the emergency core cooling system (ECCS), including the isolation valves contained within the encapsulation vessels:
The seating design of motor-operated gate valves can be a flexible or solid wedge design. Gate valve design releases the mechanical holding force during the first increment of travel so that the motor operator works only against the frictional component of the hydraulic unbalance on the disc and the packing box friction. The disc is guided throughout the full disc travel to prevent chattering and to provide ease of gate movement. The seating surfaces are hard faced to prevent galling and to reduce wear.
2.2 Regulations The NRC staff applied the following NRC regulations during the review of SNCs proposed amendments in its letter dated July 29, 2021, as supplemented by letter dated December 13, 2021.
Title 10 of the Code of Federal Regulations (10 CFR) Section 50.34, Contents of applications; technical information, paragraph (b), Final safety analysis report, which describes, in part, the information each final safety analysis report shall include that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole.
Title 10 CFR Section 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," requires in paragraph (b)(5), Long-term cooling, that after any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
Title 10 CFR Section 50.67, Accident source term, paragraphs (b)(2) states that the NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:
(i)
An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(ii)
An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(iii)
Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.
Title 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants (GDC) states, in part:
Criterion 56Primary containment isolation. Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
(1)
One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2)
One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3)
One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4)
One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.
Title 10 CFR Part 50.90, Application for amendment of license, construction permit, or early site permit, states that [w]henever a holder of a license, including a construction permit and operating license under this part, and an early site permit, combined license, and manufacturing license under part 52 of this chapter, desires to amend the license or permit, application for an amendment must be filed with the Commission, as specified in §§ 50.4 or 52.3 of this chapter, as applicable, fully describing the changes desired, and following as far as applicable, the form prescribed for original applications.
2.3 Regulatory Guidance The NRC staff applied the following guidance during the review of SNCs proposed amendments in its letter dated July 29, 2021.
NRC Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, July 2000 (ADAMS Accession No. ML003716792).
NRC Standard Review Plan (SRP) Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, Revision 1, July 1981 (ADAMS Accession No. ML052340555).
NRC SRP Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, Revision 3, December 2016 (ADAMS Accession No. ML16088A041).
NRC SRP Section 6.2.4, Containment Isolation System, Revision 3, March 2007 (ADAMS Accession No. ML070380197).
NRC SRP, Branch Technical Position (BTP) 3-3, Protection Against Postulated Piping Failures in Fluid System Outside Containment, Revision 3, March 2007 (ADAMS Accession No. ML070800027).
NRC SRP, BTP 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, Revision 3, December 2016 (ADAMS Accession No. ML16085A315).
2.4 Licensing Basis Documents The NRC staff considered the following license amendments during the review of SNCs proposed amendments in its letter dated July 29, 2021.
License Amendment Numbers 216 and 213, dated December 20, 2017, Joseph M. Farley Nuclear Plant, Units 1 and 2 - Issuance of Amendments adopting Alternative Source Term, TSTF-448, and TSTF-312 (CAC Nos. MF8861, MF8862, MF8916, MF8917, MF8918, MF8919; EPIDS Nos. L-2016-LLA-0017,L-2016-LLA-0018, L-2016-LLA-0019), (ADAMS Accession No. ML17271A265), approved to support a full-scope implementation of the Alternative Source Term radiological analysis methodology in accordance with 10 CFR Section 50.67 to perform the radiological consequences analyses of design basis accidents as described in RG 1.183, Revision 0 (July 2000).
3.0 TECHNICAL EVALUATION
3.1
NRC Staff Evaluation
The CS and RHR/LHSI systems are designed to function for post-LOCA recirculation to provide containment and core cooling, respectively. The encapsulated valves are containment isolation valves, located outside of containment, and are subject to GDC 56, Primary Containment Isolation.
The GDC 56 requires that the containment penetration line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves, one inside containment and one outside containment, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis." In Section 2.2 of the Enclosure to its letter dated July 29, 2021, the licensee indicates that the current licensing basis for Farley takes credit for the encapsulation and associated guard piping to meet GDC 56 on the other defined basis. The encapsulation and guard piping provides a pressure boundary to capture and limit leakage from the containment emergency sump suction line.
The proposed design modification and change of the UFSAR is to allow removal of the encapsulation vessels around the first isolation valves in the recirculation suction lines for the CS and RHR/LHSI systems and the associated guard piping. The proposed change includes the installation of watertight seals at the wall penetration openings created by the guard piping removal from the pump rooms. Because the licensee has taken credit for the encapsulation around the first isolation valves for meeting GDC 56 on the other defined basis in the current licensing basis, the NRC staff reviewed the proposed change against GDC 56.
The NRC staff finds that the acceptance criteria under conditions described in NRC SRP Section 6.2.4, Revision 3 are applicable to the licensees proposed change. Specifically, under Section II, Acceptance Criteria, under SRP Acceptance Criteria, No. 5 states, in part:
Containment isolation provisions for lines in engineered safety feature or engineered safety feature-related systems normally consist of two isolation valves in series. For this type of isolation valve arrangement the valve is located outside containment, and the piping between the containment and the valve should be enclosed in leak-tight or controlled-leakage housing. If, in lieu of housing, piping and valve are designed conservatively to preclude a breach of piping integrity, the design should comply with SRP Section 3.6.2 requirements.
Design of the valve or the piping compartment should provide the capability to detect and terminate leakage from the valve shaft or bonnet seals.
The NRC staff finds that one means to apply the other defined basis of GDC 56 for the containment isolation provision, is to satisfy the two conditions, as described in SRP 6.2.4. In Section 3 of the Enclosure to its letter dated July 29, 2021, the licensee addressed these two conditions as follows:
- 1. The design should comply with SRP Section 3.6.2 requirements to preclude a breach of piping integrity, and
- 2. The design of the valve or the piping compartment should provide the capability to detect and terminate leakage from the valve shaft or bonnet seals.
The NRC staff reviewed these two conditions as they relate to the licensees proposed change, as described in the next two sections.
3.1.1 Condition 1. The design should comply with SRP Section 3.6.2 requirements to preclude a breach of piping integrity.
In Section 3 of the Enclosure to its letter dated July 29, 2021, the licensee described how the piping stress analyses was performed to demonstrate conformance of the containment sump suction piping design with pertinent NRC staff guidelines as described in SRP 3.6.2 and BTP 3-4 for precluding the breach of piping integrity.
The licensee stated that the containment sump piping is designed and installed to seismic Category I requirements and that the containment sump suction piping design pressure is 80 pound per square inch gauge (psig) and the design temperature is 300 degrees Fahrenheit
(°F). The licensee also stated that the normal operating [pressure and temperature] conditions for the containment sump suction piping are significantly lower as they are isolated with borated water under stagnant conditions. The licensee stated that it considered the containment sump suction piping as moderate-energy piping in its stress analysis to ensure the piping integrity.
In its letter dated November 16, 2021 (ADAMS Accession No. ML21321A377), the NRC staff requested that the licensee clarify the normal operating conditions (operating pressure and operating temperature) for the containment sump suction piping, and the definition of the moderate-energy piping used in the proposed amendment described in SNCs letter dated July 29, 2021. In response, the licensee submitted supplemental information via letter dated December 13, 2021, and stated that the sump suction piping lines for both the RHR/LHSI and CS systems up to the first isolation valve outside of containment are isolated during operational plant conditions with a standing water leg exposed to containment atmosphere.
The expected operational plant conditions for these isolated lines would not exceed a pressure of 50 psig and a temperature of 120°F based on exposure to the containment environment and the head of water in the isolated lines. Therefore, due to the normal operating conditions (operating pressure and operating temperature) the piping is considered as moderate-energy piping. The licensee also explained that the design conditions of 80 psig and 300°F, as reflected in its letter dated July 29, 2021, were used to bound post-accident conditions and provide the limiting condition inputs to the stress analysis calculations.
The NRC staff reviewed the licensees submittal and supplemental information and determined that the licensees consideration of the containment sump suction piping as moderate-energy piping is consistent with the pertinent NRC staff guidance (i.e., operational pressure is 275 psig and operational temperature is 200°F) as delineated in Appendix A, Definitions, of BTP 3-3, Revision 3, and, therefore, is acceptable.
In its letter dated November 16, 2021, the NRC staff also requested that the licensee clarify the criteria used in the proposed amendments for precluding leakage cracks (i.e., maintaining piping integrity) in the containment sump suction piping. In its response, dated December 13, 2021, the licensee stated that it did not use the criterion of maximum stresses < 0.4(1.8 Sh +SA) required by BTP Mechanical Engineering Branch (MEB) 3-4, Revision 3. The licensee stated that instead it used the criterion of maximum stresses < 0.4(1.2 Sh +SA), which is based on MEB 3-1, Revision 1, provided in SRP 3.6.2, Revision 1. The licensee also stated that this criterion is more conservative than the criterion of maximum stresses < 0.4(1.8 Sh +SA) delineated in BTP 3-4, Revision 3, and was chosen in order to maintain consistency with the existing stress calculations. Based on its review of the licensees response as described above, the NRC staff determined that the licensees selected maximum stress criterion of < 0.4(1.2 Sh
+SA) to preclude leakage cracks for the containment sump suction piping is acceptable because the criterion used is conservative.
In its letter dated November 16, 2021, the NRC staff also requested that the licensee clarify the use of the 1974 Edition of the ASME Boiler and Pressure Vessel Code (BPV Code)Section III for performing piping analyses rather than the licensees Code of Record, the 1971 Edition of the BPV Code, including Addenda up to and including 1971. In its response, dated December 13, 2021, the licensee explained that the software used for the stress analysis calculations is ME101, which was developed using the 1974 Edition of the ASME BPV Code,Section III, and is the analysis program of record as described in Farley UFSAR Appendix 3L, Section 3L.3.2.2.
The licensee also stated that a code reconciliation of the applicable differences between the two code years was performed and concluded that using input from the 1974 Edition BPV Code is either equivalent to or more conservative than the results from the 1971 Edition BPV Code and addenda.
The NRC staff reviewed the licensees submittal and supplemental information and determined that the licensees use of ASME BPV Code Section III, 1974 Edition for performing piping analyses is acceptable because it is consistent with the piping analysis method as described in Farley UFSAR Appendix 3L, Section 3L.3.2.2.
Based on the NRC staffs review of the information provided in Section 3 of the Enclosure to SNCs letter dated July 29, 2021, and the licensees supplemental information in the letter dated December 13, 2021, the NRC staff determined that the licensee has adequately demonstrated that the containment sump suction piping design is consistent with NRC staff guidance in SRP 3.6.2 and BTP 3-4 to preclude a breach of piping integrity and, therefore, is acceptable.
3.1.2 Condition 2. The design of the valve or the piping compartment should provide the capability to detect and terminate leakage from the valve shaft or bonnet seals.
With respect to leakage detection and termination, the licensee stated in Section 3 of the Enclosure to its letter dated July 29, 2021, that:
The limiting passive failure for the affected CS and RHR/LHSI isolation valves is a stem packing catastrophic failure of the RHR/LHSI valve at the initiation of the recirculation phase. Under the proposed change, passive leakage from the valve shaft or bonnet-to-body seal would drain into the associated pump room sump until the leakage is detected remotely in the control room and the valve is closed. The leakage rate from this scenario is similar to the limiting passive failure of the downstream (non-encapsulated) isolation valve and lower than the pump seal failure case, but the total mass loss prior to isolation will be less due to the lower flow rate.
Each of the CS and RHR/LHSI pump rooms is watertight and is protected by an individual sump. Each sump receives only drainage from the individual pump room that it is protecting. The watertight rooms protect each pump against flooding from outside the room. The safety-related sump pump running alarms will indicate leakage within an individual pump room and will provide the operator with a gross indication of the magnitude of the leak.
Action taken to isolate the leak by closing the sump suction isolation valve is performed by placing the appropriate MOV hand switch to the closed position. This action can be performed prior to the mass loss impairing the ability of the [ECCS] or the [CS] system to provide long-term cooling.
Farley UFSAR Section 6.3.2.11, System Reliability, addresses criteria applied to passive failures as follows:
In design of the ECCS, Westinghouse utilizes the following criteria:
A. During the long-term cooling period following a loss-of-coolant, the emergency core cooling flow paths are separable into two subsystems, either of which can provide minimum core cooling functions and return spilled water from the floor of the containment back to the RCS [reactor coolant system].
B. Either of the two subsystems can be isolated and removed from service in the event of a leak in that subsystem outside the containment. Maximum potential leakage from components is given in table 6.3-8. For leakage used in evaluating dose consequences from a LOCA, see table 15.4-14.
C. Should one of these two subsystems be isolated in this long-term period, the other subsystem remains operable.
D. Provisions are also made in the design to detect leakage from components outside the containment, to collect this leakage, and to provide for maintenance of the affected equipment.
Thus, for the long-term emergency core cooling function, adequate core cooling capacity exists with one flow path removed from service whether isolated because of a leak, because of blocking of one flow path, or because failure of a line inside the containment results in a spill of the delivery of one subsystem.
Farley UFSAR Table 6.3-8, Maximum Potential Recirculation Loop Leakage External to Containment, lists long-term acceptable leakage rates from various components in the ECCS. For large valves, Table 6.3-8 lists Valves bonnet to body leakage rates of up to 2400 cubic centimeters (cm3) per hour (0.01 gallons per minute (gpm)) and stem packing leakage rates of 40 cm3/hour (0.0002 gpm) for valves with a 1-inch stem diameter.
Farley UFSAR Section 6.3.2.11 states that the ECCS will meet minimum core cooling requirements, and offsite doses resulting from the leak will be within 10 CFR 100 limits, for leaks up to a maximum of 50 gpm that are detected and isolated within 30 minutes.
Farley UFSAR Section 6.3.2.2.5 addresses the seating design of ECCS isolation valves. The ECCS isolation valves are motor-operated gate valves and can be a flexible or solid wedge design. When the valve is open, the bonnet of the valve is exposed to system pressure and stem packing or bonnet seal failures could result in substantial leakage. However, bonnet or stem leakage would be bounded by the 50 gpm for 30 minutes limit established for a passive pump shaft-seal failure, because of the smaller areas for flow. Both solid-wedge and flexible-wedge gate valve designs allow the valve motor-operator to firmly engage the seating surfaces of the wedge against the valve body seats when the valve is closed. This firm seating of the valve would resist containment pressure unseating the high-pressure side, thereby isolating the valve bonnet and stem from containment. Therefore, the NRC staff have reasonable assurance that the closure of the inboard isolation valve will substantially reduce the potential for leakage around the body-to-bonnet seal and along the valve stem. Any continuing leakage would be minimal, and total leakage would be bounded by the 50 gpm for 30 minutes limit. As described in the UFSAR, the remaining unaffected train would perform the recirculation function.
The NRC staff finds the maximum leakage resulting from the removal of encapsulation and associated guard piping is bounded by the current licensing basis of the pump seal failure, which is described in UFSAR Section 6.3.2.11 and Table 6.3-8, Maximum Potential Recirculation Loop Leakage External to Containment. UFSAR Section 6.3.2.11 states that piping leaks, valve packing leaks, or flange gasket leaks are less severe than the pump seal failure. Therefore, based upon the information provided by the licensee and the UFSAR, the NRC staff has determined that the licensee can adequately manage leakage following the proposed change to remove the valve encapsulation and associated guard piping. For example, the NRC staff considered the following:
the current licensing basis time for operator actions to detect and isolate a passive failure leak, leakage detection system sensitivity, leakage detection identification of the impacted train, the leakage detection alarm in control room, the drain path through the auxiliary building to the radwaste system, and the valve design that permits termination of the leakage by its closure.
The NRC staff concludes that the licensee has provided adequate justification that the design of the valve or the piping compartment provides the capability to detect and terminate leakage from the valve shaft or bonnet seals. The NRC staff finds that the proposed removal of encapsulation and associated guard piping is consistent with SRP Section 6.2.4, as based on the conformance of SRP Section 3.6.2 and adequacy of leakage detection and termination, and therefore, will continue to meet GDC 56 on the other defined basis.
3.1.3 Design Basis Accident Analysis Farley UFSAR Section 6.3.2.11 describes the current configuration of the systems as:
With respect to piping and mechanical equipment outside the containment, considering the provisions for visual inspection and leak detection, leaks will be detected before they propagate to major proportions. A review of the equipment in the system indicates that the largest sudden leak potential would be the sudden failure of a pump shaft seal. Evaluation of leak rate assuming only the presence of a seal retention ring around the pump shaft showed flows less than 50 [gpm] would result. Piping leaks, valve packing leaks, or flange gasket leaks have been of a nature to build up slowly with time and are considered less severe than the pump seal failure.
The NRC staff reviewed the applicable section in the Farley UFSAR and the licensees proposed amendment described in its letter dated July 29, 2021. The NRC staff found that the current configuration of the systems maintain a pair of automatic MOVs in series outside of containment for each sump suction line. The NRC staff found that each of the encapsulated valves has a redundant unencapsulated valve located in series, along with other valves, within the same watertight pump room. As a result, the NRC staff found that the removal of the encapsulation vessels and associated guard piping around the first isolation valves in the recirculation suction lines for the CS and RHR/LHSI systems will not result in a new or different failure mode type than is currently present with similar, unencapsulated valves. Therefore, the NRC staff found that the removal of the encapsulation and associated guard piping with each of these valves does not affect and is bounded by the current UFSAR analysis which is based upon handling leaks up to a maximum of 50 gpm, and the ability to detect and isolate such leaks in the emergency core cooling path within 30 min.
The NRC staff found that any portion of the postulated leakage after removal of the encapsulation and associated guard piping, which becomes airborne, will be processed by and remain within the design basis of the safety-related Penetration Room Filtration System as credited in the current licensing basis. The NRC staff found that the applicable sections of the UFSAR remain bounding when the encapsulation and associated guard piping is removed from the first isolation valves in the recirculation suction lines for the CS and RHR/LHSI system valves, because the removal of encapsulation and associated guard piping from the valves will not result in a new or different failure mode type than is currently present with similar, unencapsulated valves.
The NRC staff confirmed that no change to the previously approved Alternate Source Term (AST) dose analysis is required due to the changes described in the proposed amendment described in SNCs letter dated July 29, 2021, because the failure of the unencapsulated valve was previously factored into the AST analysis as described in UFSAR Section 15.4.1.7.
Therefore, there is no impact on the previously approved design basis accidents (DBA) or the DBAs associated with the licensees approved AST as a result of the licensees proposed removal of the encapsulation and associated guard piping. Accordingly, the Farley UFSAR continues to meet the requirements of 10 CFR Section 50.67 and to be consistent with RG 1.183.
The NRC staff reviewed the current licensing basis and the assumptions used by the licensee to assess the changes requested in the proposed amendment described in SNCs letter dated July 29, 2021. Based on its review, as discussed above, the NRC staff finds these changes do not affect any previously approved DBA dose consequence analysis, including those associated with the licensees approved AST and, therefore, continue to meet regulatory requirements set forth in 10 CFR 50.67. Therefore, the NRC staff concludes that the proposed license amendment does not impact the radiological dose consequence analysis of Farleys DBAs.
3.2 NRC Staff Conclusion
Based on its review, the NRC staff concludes that the proposed change to the UFSAR, as described in letter dated July 29, 2021, as supplemented by letter dated December 13, 2021, to allow removal of the encapsulation vessels and associated guard piping around the first isolation valves in the recirculation suction lines for the CS and RHR/LHSI systems for Farley is consistent with regulations 10 CFR 50.34(b), 10 CFR 50.46, 10 CFR 50.67, 10 CFR 50.90, and GDC 56, and NRC staff guidance in SRP Section 3.6.2, SRP Section 6.2.4, and BTP 3-4, Revision 3. Therefore, the proposed change is found to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Alabama State official was notified of the proposed issuance of the amendments on February 01, 2022. On February 02, 2022, the official stated that the State of Alabama had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration (86 FR 55013, October 5, 2021), and there has been no public comment on such finding. Accordingly, the amendments meet 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 amendments.
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 Commissions regulations; and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
Chang Li, NRR Yueh-Li Li, NRR Sean Meighan, NRR Steve Smith, NRR Date of Issuance: March 7, 2022
ML22032A243 OFFICE DORL/LPL2-1/PM DORL/LPL2-1/LA DSS/SCPB/BC(A)
DRA/ARCB/BC NAME SDevlin-Gill KGoldstein SJones KHsueh DATE 02/01/2022 02/03/2022 01/31/2022 01/28/2022 OFFICE DEX/EMIB/BC(A)
DSS/STSB/BC OGC/NLO DORL/LPL2-1/BC NAME ITseng VCusumano MFWoods MMarkley DATE 01/31/2022 02/02/2022 02/28/2022 03/07/2022 OFFICE DORL/LPL2-1/PM NAME SDevlin-Gill DATE 03/07/2022