ML24326A134
| ML24326A134 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 11/21/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24326A092 | List: |
| References | |
| LO-175900 | |
| Download: ML24326A134 (1) | |
Text
Response to SDAA Audit Question Question Number: A-16.2.1.2-2 Receipt Date: 02/19/2024 Question:
Bases Subsection B 2.1.2, ASA section, second paragraph about the role of the reactor safety valves (RSVs)
(b) has an editorial error in second sentence because just one transient is listed (turbine trip at full power without bypass capability), so the sentence should begin The transientis instead of The transientsare; (c) in third sentence, the listed instrument Function is high main steam line pressure, which is Module Protection System (MPS) Function 17.b, High Main Steam Pressure actuation of Decay Heat Removal System (DHRS) (labeled 16.b in US600). However, US600 B 2.1.2 mentioned pressurizer pressure, which is presumed to be MPS Function 7.b, High Pressurizer Pressure actuation of DHRS. Please confirm that Function 17.b is correct.
The NRC provided the following request as a follow-up to the response to A-16.2.1.2-2:
"Request call with NuScale to point to description of RSV sizing calculation assumed limiting event in an SDAA Chapter 5 topical report. SDAA Part 2 Section 5.2.2.2.1 in part says: A turbine trip at full power without bypass capability is the most severe AOO and is the bounding event used in the determination of RSV capacity and the RPV overpressure analyses. Sizing of the RCS and the PZR steam space avoids an RSV lift during normal operational transients that produce the highest RPV pressure at full power conditions, with system and core parameters within normal operating range. In the event of a safety valve lift, the size of the PZR steam space is sufficient to preclude liquid discharge. The analytical model used for the analysis of the overpressure protection system and the basis for its validity is in the NuScale Topical Reports "Non-Loss-of-Coolant Accident Analysis Methodology," TR-0516-49416-P, Revision 4 (Reference 5.2-1) and "Loss-of-Coolant Accident Evaluation Model," TR-0516-49422-P, Revision 3 (Reference 5.2-2).
NuScale Nonproprietary NuScale Nonproprietary
Also, request to discuss including FSAR Section 5.2.2.2.1 and the above topical report(s) in SDAA Part 4 Subsection B 3.4.4, References section."
Response
Part (b)
NuScale revises the second sentence of the second paragraph of the Applicable Safety Analyses Bases for GTS Section B 2.1.2 to read:
The transient selected to establish the required relief capacity, and hence valve size requirements and lift settings, is a turbine trip at full power without bypass capability.
Part (c)
The reactor safety valve capacity calculation for the NuScale US460 design assumes the bounding design-basis transient for reactor safety valve capacity is the turbine trip without bypass. Consistent with Acceptance Criterion 3B(iii) of the standard review plan NUREG-0800, Revision 3, Section 5.2.2, the second safety-grade signal initiates the reactor scram. Therefore, the high pressurizer pressure reactor trip is defeated in the capacity calculation. The reference to high main steam pressure (Function 17.b) in the third sentence of the second paragraph of the Applicable Safety Analyses Bases for GTS Section B 2.1.2 is correct. NuScale revises the reactor safety valve sizing discussion in the Bases for GTS 2.1.2 to include a reference to Final Safety Analysis Report Chapter 5.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
RCS Pressure SL B 2.1.2 NuScale US460 B 2.1.2-2 Draft Revision 2 BASES APPLICABLE SAFETY ANALYSES (continued)
The RSVs are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in Section III of the ASME Code (Ref. 2). The transients selected to establish the required relief capacity, and hence valve size requirements and lift settings, isare a turbine trip at full power without bypass capability (Ref. 6). During the transient, no control actions are assumed except that the Decay Heat Removal System valves on the secondary plant are assumed to open when the high main steam line pressure reaches the Decay Heat Removal System actuation setpoint.
The Module Protection System (MPS) setpoints provide pressure protection for normal operation and AOOs. The MPS high pressurizer pressure trip setpoint is set to provide protection against overpressurization (Ref. 4). The safety analyses for both the high pressurizer pressure trip and the RSVs are performed using conservative assumptions relative to pressure control devices.
More specifically, no credit is taken for operation of the following:
ad. Turbine Bypass System;
- b. Reactor Control System;
- c. Pressurizer Level Control System; andor
- d. Pressurizer spray.
SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel, piping, valves, and fittings under the ASME Code,Section III, is 110% of design pressure; therefore, the maximum allowable pressurizer pressure is 2420 psia.
APPLICABILITY SL 2.1.2 applies in MODES 1, 2, and 3 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODES 4 and 5 since the reactor vessel is vented to the containment until the upper reactor vessel assembly is removed, following which, the reactor vessel is vented directly to the ultimate heat sink; thus, making it unlikely that the RCS can be pressurized.
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.
- 6. FSAR Chapter 5.