ML24215A184
| ML24215A184 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 08/02/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24215A000 | List:
|
| References | |
| LO-169995 | |
| Download: ML24215A184 (1) | |
Text
Response to SDAA Audit Question Question Number: A-16.5.5.04-1 Receipt Date: 07/17/2023 Question:
Section 16.1.1 of the Standard Design Approval Application (SDAA) [1] states, These revised GTS [Generic Technical Specifications] were developed consistent with the Improved Standard Technical Specifications [and Bases] (ISTS) format and content typified in NUREG-1431, Revision 5, and NUREG 1432, Revision 5. These ISTS NUREGs are referred to below as WSTS for Westinghouse ISTS [2] and CE STS for Combustion Engineering ISTS [3],
respectively.
The U.S. Nuclear Regulatory Commission (NRC) staff noted the following differences between the NuScale US460 GTS and Bases [4] and the ISTS (with emphasis added by the NRC staff):
(1) W-STS Subsection 5.5.8, Steam Generator (SG) Program, states in paragraph 5.5.8.d.2, for steam generators (SGs) with Alloy 690 thermally treated tubing, After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 96 effective full power months, which defines the inspection period. However, corresponding GTS Subsection 5.5.4, Steam Generator (SG) Program, states in paragraph 5.5.4.d.2, After the first refueling outage following SG installation, inspect each SG at least every 96 effective full power months, which defines the inspection period.
(2) GTS Subsection 5.6.5, Steam Generator Tube Inspection Report, does not match the reporting requirements in corresponding W-STS Subsection 5.6.7, Steam Generator Tube Inspection Report.
(3) The Applicable Safety Analyses section in W-STS Bases Subsection B 3.4.20, Steam Generator (SG) Tube Integrity, states, The accident analysis for a SGTR [SG tube rupture]
assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser. However, the Applicable Safety NuScale Nonproprietary NuScale Nonproprietary
Analyses section in corresponding GTS Bases Subsection B 3.4.9, Steam Generator (SG)
Tube Integrity, states, The accident analysis for a SGTF [SG tube failure] assumes the contaminated secondary fluid is released to the atmosphere.
Please provide sufficient technical information to justify the above differences between the NuScale US460 GTS and Bases and the W-STS and Bases.
References:
[ ] NuScale Power, LLC Submittal of the NuScale Standard Design Approval Application Part 2
- Final Safety Analysis Report, Chapter 16, Technical Specifications, Revision 0 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22363A194).
[2] NUREG 1431, Standard Technical Specifications - Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases, Revision 5, dated September 2021 (ML21259A155 and ML21259A159, respectively).
[3] NUREG 1432, Standard Technical Specifications - Combustion Engineering Plants, Volume 1, Specifications, and Volume 2, Bases, Revision 5, dated September 2021 (ML21258A421 and ML21258A424, respectively).
[4] NuScale Power, LLC Submittal of the NuScale Standard Design Approval Application - Part 4, "US460 Generic Technical Specifications," Volumes 1 and 2, Revision 0 (ML23001A000).
Feedback discussed during 10/18/23 NRC call:
In the last sentence of the second paragraph of the APPLICABLE SAFETY ANALYSES section of Generic Technical Specification (GTS) Bases Subsection B 3.4.9, the reference to 10 CFR 50.34 (Ref. 4) was changed to 10 CFR 50.34a (Ref. 4). The staff notes that the response to A-16.3.4-1 made a conforming change in the REFERENCES section of GTS Bases Subsection B 3.4.9.
10 CFR 50.34 is Contents of applications; technical information, and 10 CFR 50.34a is Design Objective for equipment to control releases of radioactive material in effluents-nuclear power reactors. Please discuss the basis for changing 10 CFR 50.34 (Ref. 4) to 10 CFR 50.34a (Ref. 4).
The basis for this change is unclear to the staff because it appears that 10 CFR 50.34a is limited to normal reactor operations, including expected operational occurrences, and not accidents like a steam generator tube failure. Criteria for radiological consequences of accidents are included in 10 CFR 50.34. Therefore, please provide sufficient NuScale Nonproprietary NuScale Nonproprietary
technical information to justify changing 10 CFR 50.34 (Ref. 4) to 10 CFR 50.34a (Ref. 4).
- Paragraph 5.6.5.g in GTS Subsection 5.6.5 is in brackets. Please discuss the basis for including paragraph 5.6.5.g in brackets, including whether its anticipated that a plant referencing the NuScale US460 standard design will have plant-specific reporting requirements related to the steam generators. If its anticipated, please discuss what type of information would be reported in accordance with paragraph 5.6.5.g.
Response
NuScale revises Technical Specification Section 5.5.4.d.2 to include inspection of 100% of the tubes in each steam generator.
NuScale revises Technical Specification 5.6.5, Steam Generator Tube Inspection Report, to align with the guidance provided in Section 5.6.7 of the Westinghouse Standard Technical Specifications, NUREG-1431, Revision 5.
The NuScale evaluation of the radiological consequences of a steam generator tube failure assumes primary coolant flows into the secondary coolant through the failed steam generator resulting in a time-dependent release of activity into the steam lines and condenser and then to the atmosphere. NuScale revises the Technical Specification Bases Section B 3.4.9 discussion of steam generator tube failure leakage to reflect these assumptions.
Supplemental Response:
Markups of references to 10 CFR 50.34 were included on Technical Specification pages associated with the original response to Audit Question A-16.5.5.04-1 Those changes were part of an earlier change and are unrelated to the response to A-16.5.5.04-1. After further review, NuScale has determined that the change of 10 CFR 50.34 to 10 CFR 50.34a was misinformed.
As a result, NuScale revises the following Technical Specification Sections to replace 10 CFR 50.34a with 10 CFR 50.34.
NuScale Nonproprietary NuScale Nonproprietary
B 2.1.2 B 3.3.1 B 3.4.9 B 3.7.1 B 3.7.2 The comment associated with GTS Paragraph 5.6.5.g was resolved during a 10/18/2023 clarification call with the NRC.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
SG Tube Integrity B 3.4.9 NuScale US460 B 3.4.9-2 Draft Revision 1 BASES APPLICABLE The steam generator tube failure (SGTF) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTF is the basis for this ANALYSES Specification. The analysis of a SGTF event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.5, RCS Operational LEAKAGE, plus the leakage rate associated with a double-ended failure of a single tube. The accident analysis for a SGTF assumes the contaminated secondary fluid is released primary coolant flows into the secondary coolant through the failed steam generator resulting in a time-dependent release of activity into the steam lines and condenser and then to the atmosphere.
The analysis for design basis accidents and transients other than a SGTF assume the SG tubes retain their structural integrity (i.e., they are assumed not to fail). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.8, RCS Specific Activity, limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2) which NuScale implements as Principal Design Criterion 19 described in FSAR section 3.1 (Ref. 3),
10 CFR 50.34a (Ref. 4) or the NRC approved licensing basis (e.g., a small fraction of these limits).
Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging criteria be plugged in accordance with the Steam Generator Program.
During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging.
If a tube was determined to satisfy the plugging criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.
The tube-to-tubesheet weld is not considered part of the tube.
A SG tube has tube integrity when it satisfies the SG performance criteria.
The SG performance criteria are defined in Specification 5.5.4, Steam Generator Program, and describe acceptable SG tube performance.
Programs and Manuals 5.5 NuScale US460 5.5-5 Draft Revision 1 5.5 Programs and Manuals 5.5.4 Steam Generator (SG) Program (continued) c.
Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
1.
Inspect 100% of the tubes in each SG during the first refueling outage following initial startup or SG replacement.
2.
After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.
3.
If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected unit SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e.
Provisions for monitoring operational primary to secondary LEAKAGE.
Reporting Requirements 5.6 NuScale US460 5.6-5 Draft Revision 1 5.6 Reporting Requirements 5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a.
RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
3.3.1, Module Protection System (MPS) Instrumentation; 3.3.3, Engineered Safety Features Actuation System (ESFAS)
Logic and Actuation; 3.3.4, Manual Actuation Functions; 3.4.3, RCS Pressure and Temperature (P/T) Limits; and 3.4.4, Reactor Safety Valves (RSVs).
- b.
The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:
TR-130877-P, "Pressure and Temperature Limits Methodology,"
[Revision 2, December 2022.]
- c.
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluency period and for any revision or supplement thereto.
5.6.5 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 3 following completion of an inspection performed in accordance with the Specification 5.5.4, "Steam Generator (SG) Program." The report shall include:
- a.
The scope of inspections performed on each SG;.
- b.
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;Degradation mechanisms found.
- c.
For each degradation mechanism found:Nondestructive examination techniques utilized for each degradation mechanism.
- 1. The nondestructive examination techniques utilized;
Reporting Requirements 5.6 NuScale US460 5.6-6 Draft Revision 1 5.6 Reporting Requirements 5.6.5 Steam Generator Tube Inspection Report (continued)
- 2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
- 3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and
- 4. The number of tubes plugged during the inspection outage.
- d.
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results; Location orientation (if linear),
and measured sizes (if available) of service induced indications.
- e.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;Number of tubes plugged during the inspection outage for each degradation mechanism.
- f.
The results of any SG secondary side inspections; andThe number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator.
[g.
Insert any plant-specific reporting requirements, if applicable.]The results of condition monitoring, including the results of tube pulls and in situ testing.
RCS Pressure SL B 2.1.2 NuScale US460 B 2.1.2-3 Draft Revision 2 BASES SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in MODE 1 VIOLATIONS the requirement is to restore compliance and be in MODE 2 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for abnormal radioactive releases (Ref. 5).
The allowable Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.
If the RCS pressure SL is exceeded in MODE 2 or 3, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 2 or 3 may be more severe than exceeding this SL in MODE 1 since the reactor vessel temperature is lower and the vessel material, consequently, less ductile. As such, pressurizer pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.
REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.
- 2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000,
[2017 edition].
- 3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWA-5000, [2017 edition].
- 4. FSAR, Chapter 7.
- 5. 10 CFR 50.34a.
MPS Instrumentation B 3.3.1 NuScale US460 B 3.3.1-3 Draft Revision 2 BASES BACKGROUND (continued) reactor module must enter the Condition for the particular MPS Functions affected. The channel as-found condition will be entered into the Corrective Action Program for further evaluation and to determine the required maintenance to return the channel to OPERABLE.
During AOOs, which are those events expected to occur one or more times during the plant life, the acceptable limits are:
The critical heat flux ratio (CHFR) shall be maintained above the SL value to prevent critical heat flux (CHF);
Fuel centerline melting shall not occur; and Pressurizer pressure SL of 2420 psia shall not be exceeded.
Maintaining the variables within the above values ensures that the offsite dose will be within the 10 CFR 50 (Ref. 2) and 10 CFR 50.34a (Ref. 3) criteria during AOOs.
Accidents are events that are analyzed even though they are not expected to occur during the plant life. The acceptable limit during accidents is that the offsite dose shall be maintained within an acceptable fraction of 10 CFR 50.34a (Ref. 3) limits. Different accident categories allow a different fraction of these limits based on probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.
The MPS includes devices and circuitry that generate the following signals when monitored variables reach levels that are indicative of conditions requiring protective action:
- 1. Reactor Trip System (RTS) actuation;
- 2. Emergency Core Cooling System (ECCS) actuation;
- 3. Decay Heat Removal System (DHRS) actuation;
- 4. Containment Isolation System (CIS) actuation;
- 5. Secondary System Isolation (SSI);
- 6. Chemical and Volume Control System Isolation (CVCSI) actuation;
- 7. Demineralized Water Supply Isolation (DWSI) actuation;
MPS Instrumentation B 3.3.1 NuScale US460 B 3.3.1-60 Draft Revision 2 BASES SURVEILLANCE REQUIREMENTS (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.5 SR 3.3.1.5 is the performance of a CHANNEL CALIBRATION of the Class 1E isolation devices, as described in SR 3.3.1.4.
Class 1E isolation devices ensure that electrical power to the associated MPS circuitry and logic will not adversely affect the ability of the system to perform its safety functions. The devices de-energize and isolate the MPS components if such a condition is detected. This surveillance verifies the setpoints and functions of the isolation devices including associated alarms and indications by performing a CHANNEL CALIBRATION of required Class 1E isolation devices. The overcurrent and undervoltage setpoints of the Class 1E isolation devices are established and controlled in accordance with the Setpoint Program. The calibration parameters associated with the CHANNEL CALIBRATION of these Class 1E isolation devices are established to assure component OPERABILITY of the device electrical protection and isolation functions. There are no LSSSs associated with the Class 1E devices such that the establishment of a limiting trip setpoint (LTSP) or nominal trip setpoint (NTSP) is not governed by the Setpoint Program. However, the performance of a CHANNEL CALIBRATION implements sections of the Setpoint Program and includes the channel OPERABILITY determination based on the As-Found and As-Left settings for the Class 1E device calibration parameters.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES
- 1. Regulatory Guide 1.105, Revision 4, February 2021.
- 3. 10 CFR 50.34a.
- 4. FSAR, Chapter 7.
- 5. FSAR, Chapter 15.
- 6. 10 CFR 50.49.
SG Tube Integrity B 3.4.9 NuScale US460 B 3.4.9-2 Draft Revision 2 BASES APPLICABLE The steam generator tube failure (SGTF) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTF is the basis for this ANALYSES Specification. The analysis of a SGTF event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.5, RCS Operational LEAKAGE, plus the leakage rate associated with a double-ended failure of a single tube. The accident analysis for a SGTF assumes primary coolant flows into the secondary coolant through the failed steam generator resulting in a time-dependent release of activity into the steam lines and condenser and then to the atmosphere.
The analysis for design basis accidents and transients other than a SGTF assume the SG tubes retain their structural integrity (i.e., assumed not to fail). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.8, RCS Specific Activity, limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2) which NuScale implements as Principal Design Criterion 19 described in FSAR section 3.1 (Ref. 3),
10 CFR 50.34a (Ref. 4) or the NRC approved licensing basis (e.g., a small fraction of these limits).
Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The LCO requires SG tube integrity be maintained. The LCO also requires all SG tubes that satisfy the plugging criteria be plugged in accordance with the Steam Generator Program.
During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging.
If a tube was determined to satisfy the plugging criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.
The tube-to-tubesheet weld is not considered part of the tube.
A SG tube has tube integrity when it satisfies the SG performance criteria.
The SG performance criteria are defined in Specification 5.5.4, Steam Generator Program, and describe acceptable SG tube performance.
SG Tube Integrity B 3.4.9 NuScale US460 B 3.4.9-7 Draft Revision 2 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency of prior to entering MODE 3 following a SG inspection ensures the Surveillance is complete and all tubes meeting the plugging criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.
REFERENCES
- 1. NEI 97-06, Rev. [3].
- 3. FSAR, Section 3.1.
- 4. 10 CFR 50.34a.
- 5. ASME, Boiler and Pressure Vessel Code,Section III, Subsection NB,
[2017 edition].
- 6. Draft Regulatory Guide 1.121, August 1976.
- 7. EPRI, Pressurized Water Reactor Steam Generator Examination Guidelines, Rev. [7].
MSIVs B 3.7.1 NuScale US460 B 3.7.1-2 Draft Revision 2 BASES APPLICABLE The MSIVs and MSIV Bypass Isolation Valves close to isolate the SAFETY SGs from the power conversion system. Isolation limits ANALYSES postulated releases of radioactive material from the SGs in the event of a SG tube failure (Ref. 4) and terminates flow from SGs for postulated steam line breaks outside containment (Ref. 5). This minimizes radiological contamination of the secondary plant systems and components, minimizes associated potential for activity releases to the environment, and preserves RCS inventory in the event of a SGTF.
The isolation of steam lines is also required for the operation of the DHRS. Isolation valve closure precludes blowdown of more than one SG, preserving the heat transfer capability of an unaffected SG if a concurrent single failure occurs. The DHRS provides cooling for non-loss-of-coolant accident (non-LOCA) design basis events when normal secondary-side cooling is unavailable or otherwise not utilized. The DHRS removes post-reactor trip residual and core decay heat and allows transition of the reactor to safe shutdown conditions.
The safety-related and nonsafety-related MSIV and MSIV bypass valves satisify Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO This LCO requires four isolation valves on each SG steam line to be OPERABLE. This includes safety-related and nonsafety-related MSIVs and MSIV bypass valves in each steam line. The valves are considered OPERABLE when they will close on an isolation actuation signal, their isolation times are within limits, and valve leakage is within limits.
This LCO provides assurance that the safety-related and nonsafety-related MSIVs and MSIV bypass valves will be available to perform their design safety function to limit consequences of accidents that could result in offsite exposures comparable to the 10 CFR 50.34a limits or the NRC staff approved licensing basis. This LCO also provides assurance that the safety-related and non-safety related MSIVs and MSIV bypass valves will be available to perform their design safety function in support of the DHRS as described in LCO 3.5.2.
APPLICABILITY The safety-related and nonsafety-related MSIVs and MSIV Bypass Valves must be OPERABLE in MODE 1, 2, and MODE 3 when not PASSIVELY COOLED. Under these conditions, the isolation of the MSIVs ensures the DHRS can perform its design function and the valves provide a barrier to limit the release of radioactive material to the environment.
Closure of the MSIVs also preserves the RCS inventory in the event of a SGTF. Therefore, these valves must be OPERABLE or the flow path
Feedwater Isolation B 3.7.2 NuScale US460 B 3.7.2-2 Draft Revision 2 BASES APPLICABLE SAFETY ANALYSES (continued) of coolant accident (non-LOCA) design basis events when normal secondary side cooling is unavailable or otherwise not utilized. The DHRS removes post-reactor trip residual and core decay heat and allows transition of the reactor to safe shutdown conditions. The FWIV and FWRV have a specific leakage criteria to maintain DHRS inventory.
The FWIV and FWRV satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO This LCO requires the FWIVs and FWRV in each of the two feedwater lines to be OPERABLE. The valves are considered OPERABLE when their isolation times are within limits and they close on an isolation actuation signal and their leakage is within limits.
This LCO provides assurance that the FWIVs will perform their design safety function and the FWRVs their non-safety function to limit consequences of accidents that could result in offsite exposures comparable to the 10 CFR 50.34a limits or the NRC staff approved licensing basis.
APPLICABILITY The FWIVs and FWRVs must be OPERABLE whenever there is significant mass and energy in the Reactor Coolant System and the steam generators. This ensures that, in the event of a high energy line break, a single failure cannot result in the blowdown of more than one steam generator, an inoperability of the DHRS, or a containment bypass path in the event of a steam generator tube failure. In MODE 1 and 2 FWIVs and FWRVs are required to be OPERABLE to limit the amount of available fluid that could be added to containment in case of a secondary system pipe break inside containment. In MODE 3 and not PASSIVELY COOLED, the FWIVs and FWRV are required to be OPERABLE, to support DHRS operability.
In MODES 4 and 5 the steam generators energy is low. Therefore, the FWIVs and FWRVs are normally closed since FW system is not required.
ACTIONS The ACTIONS table is modified by two Notes. The first being that separate entry is allowed for each valve. This is acceptable because the ACTIONS table provide actions for individual component entry. The second indicating that FWIV flow path may be unisolated intermittently under administrative control.