RAIO-0418-59506, LLC Response to NRC Request for Additional Information No. 368 (Erai No. 9242) on the NuScale Design Certification Application
| ML18101B407 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 04/11/2018 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| References | |
| RAIO-0418-59506 | |
| Download: ML18101B407 (17) | |
Text
RAIO-0418-59506 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com April 11, 2018 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Response to NRC Request for Additional Information No.
368 (eRAI No. 9242) on the NuScale Design Certification Application
REFERENCE:
U.S. Nuclear Regulatory Commission, "Request for Additional Information No.
368 (eRAI No. 9242)," dated February 14, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).
The Enclosure to this letter contains NuScale's response to the following RAI Questions from NRC eRAI No. 9242:
04.06-1 04.06-2 This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions on this response, please contact Paul Infanger at 541-452-7351 or at pinfanger@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution:
Samuel Lee, NRC, OWFN-8G9A Prosanta Chowdhury NRC, OWFN-8G9A Bruce Bavol, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9242 Zackary W. Rad Director Regulatory Affairs
RAIO-0418-59506 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com :
NuScale Response to NRC Request for Additional Information eRAI No. 9242
NuScale Nonproprietary Response to Request for Additional Information Docket No.52-048 eRAI No.: 9242 Date of RAI Issue: 02/14/2018 NRC Question No.: 04.06-1 10 CFR 50, Appendix A, General Design Criterion (GDC) 4 requires that structures, systems, and components important to safety be designed to accommodate the effects of, and to be compatible with, the environmental conditions during normal plant operation as well as during postulated accidents. 10 CFR 52.47 requires the information submitted for a design certification to include performance requirements and design information sufficiently detailed to permit procurement specifications and construction and installation specifications by an applicant. In addition, the Standard Review Plan (SRP), Revision 2, Section 4.6 provides guidance regarding the control rod drive cooling system (CRDS); specifically, Review Procedure 3 directs the reviewer to examine descriptions and drawings to confirm that the systems meet the design requirements, and specifies that the CRDS cooling system should be capable of maintaining the CRDS temperature below the applicants maximum temperature criterion.
Final Safety Analysis Report (FSAR) Tier 2, Section 4.6.1 states the electric coil operating conditions of the CRDS requires active cooling by water through a CRDS cooling water distribution header to cooling tubes in the drive coils of each control rod drive mechanism (CRDM). Section 4.6.1 adds that the cooling requirements for the CRDMs are provided by the reactor component cooling water system (RCCWS) in Section 9.2.2. The staff reviewed FSAR Tier 2, Section 4.6 and Section 9.2.2, but could not find the cooling requirements for the CRDS.
The applicant is requested to provide in the FSAR, maximum temperature criterion for the CRDM, and RCCWS cooling water temperature values and flow rates required to maintain adequate CRDM cooling for normal operation.
NuScale Response:
The Control Rod Drive Mechanism (CRDM) coils are designed using a Class N insulation system, which is rated to 392 degrees Fahrenheit. In order to provide margin, the Reactor Component Cooling Water System (RCCWS) is designed to limit coil temperatures consistent with those established for one insulation class lower, or 356 degrees Fahrenheit. Therefore, the maximum temperature design criterion for the CRDM is 356 degrees Fahrenheit.
NuScale Nonproprietary The CRDM RCCWS cooling water parameters were previously provided in the original (RAIO-1117-57110) and supplemental (RAIO-1217-57637) responses to RAI 09.02.02-4 (eRAI 9101). These letters were transmitted to the NRC on November 10, 2017 and December 12, 2017. The CRDM RCCWS parameters are as follows:
Normal Operation:
Flow Rate for Each CRDM = 2 gpm Heat Load for Each CRDM = 40,200 Btu/hr RCCWS Temperature in = 80 degrees Fahrenheit RCCWS Temperature out = 120.5 degrees Fahrenheit Sizing Basis:
Flow Rate for Each CRDM = 2 gpm Heat Load for Each CRDM = 40,200 BTU/hr RCCWS Temperature in = 100 degrees Fahrenheit RCCWS Temperature out = 140.5 degrees Fahrenheit These values are based on preliminary hardware designs and may change when site-specific conditions are incorporated and final detailed designs are complete. For this reason and also because the safety function of the CRDM, which is to insert upon a reactor trip, is not affected by the loss of CRDM cooling, these preliminary values for cooling parameters are not added to the FSAR.
Impact on DCA:
FSAR Section 4.6.1 has been been revised as described in the response above and as shown in the markup provided in this response.
NuScale Final Safety Analysis Report Functional Design of Control Rod Drive System Tier 2 4.6-2 Draft Revision 2 electromagnetic coils and housings, including the pressure housings. The major components of the CRDM are annotated, and detailed in the subsequent figures. The power and cooling water connectors are located on top of the mast assembly and sensor coil for ease of access through the removable cover on top of the CNV (Figure 4.6-1).
Figure 4.6-3 illustrates the CRDM drive coil and embedded cooling coils shown on the right view without the coil stack housings and mast assembly. The electrical connector on top of the left view is located above the cooling water fittings for separation purposes.
Figure 4.6-4 shows the layout of the rod position indicator sensor coil assemblies which are located directly above the rod travel housing. Rod position indication is facilitated by means of electromagnetic induction in the sensor coils, as the top of the control rod drive shaft travels upwards or downwards within the pressure boundary. Figure 4.6-5 provides an overview of the latch mechanism assembly (LMA), with the remote disconnect latch shown separately for better illustration. The three magnetic poles, latches and grippers on the left represent an industry-standard LMA design that performs the rod withdrawal/
insertion/reactor trip functions, whereas the remote disconnect grippers (RDG) are relied upon during the remote disconnection/re-connection for NPM refueling only. Figure 4.6-6 illustrates the remote disconnection of the control rod drive shaft from the CRA that is not available in the operating NPM location, in order to preclude inadvertent CRA disengagement.
The CRDM assembly is a hermetically sealed electro-mechanical device, which moves the CRA in and out of the reactor core, and holds the CRA at any elevation within the range of CRA travel. If electrical power is interrupted to the CRDM, the control rod drive shaft is released, and the attached CRA drops into the reactor core.
RAI 04.06-1 The CRDMs are mounted on the RPV head, and the CRDM pressure housings are safety-related American Society of Mechanical Engineers (ASME) Class 1 pressure boundaries. The CRDS components internal to the reactor coolant pressure boundary are designed to function in borated primary coolant with up to 2000 ppm boron at primary coolant pressures and temperatures ranging from ambient conditions to 650 degrees F design temperature and 2,100 psia RPV design pressure. During normal operating conditions the upper portion of the RPV and the CRDM pressure housing are in contact with saturated steam on the inside at 625 degrees F and 1850 psia. The lower portion of the drive rod is submerged in the primary coolant at hot leg temperature flowing upward through the upper riser and CRA guide tubes. The electric coil operating conditions require active cooling by water through a CRDS cooling water distribution header to cooling tubes in the drive coils of each CRDM as shown in Figure 4.6-3. The cooling requirements for the CRDMs are provided by the reactor component cooling water system (RCCWS) in Section 9.2.2. The RCCWS is designed to maintain the CRDM winding temperature below the design maximum temperature of 356 degrees Fahrenheit.
The CRDS cooling line is branched into supply lines inside the containment vessel to each individual CRDM. After passing through the CRDM cooling tubes, the flexible return lines rejoin into a single return header leaving containment. A thermal relief valve is provided on the return header to provide overpressure protection for the CRDS cooling piping during a containment isolation event.
The structural materials of construction for the CRDS are discussed in detail in Section 4.5.1.
NuScale Nonproprietary Response to Request for Additional Information Docket No.52-048 eRAI No.: 9242 Date of RAI Issue: 02/14/2018 NRC Question No.: 04.06-2 GDC 26, Reactivity Control System Redundancy and Capability, requires two independent reactivity control systems of different design principles that are capable of reliably controlling reactivity changes during normal operation. GDC 29 requires that protection and reactivity control systems be designed to ensure an extremely high probability of functioning in the event of an anticipated operational occurrence. The CRDS is one of those systems, and the areas of review under SRP Section 4.6, Functional Design of Control Rod Drive System, include functional tests for the CRDS.
Regulatory Guide 1.68, Revision 4, and Design Specific Review Standard (DSRS) Section 14.2 provide guidance for testing of the CRDS. DSRS Section 14.2 states that the applicant should provide test abstracts of SSCs and unique design features, including tests and acceptance criteria.
FSAR Tier 2, Section 14.2 provides the elements of the Initial Test Program (ITP). The information provided regarding these tests are not sufficient to ensure adequacy of the results of the ITP, specifically:
- a. FSAR Tier 2, Table 14.2-80 (Test#80), does not include specific acceptance criteria for CRDS performance. No numerical values are specified for ITP rod insertion and withdrawal speeds, the limit for control rod assembly (CRA) position indications within the associated group position or control rod demand position, and the CRA fully withdrawn position. The applicant is requested to provide the design limits within the acceptance criteria of Test #80 or reference a location in the FSAR that provides the values for the design limits.
- b. Acceptance criteria i' of FSAR Tier 2, Table 14.2-81 (Test #81), does not provide specific acceptance values for drop time. The applicant is requested to provide the drop time within the acceptance criteria of Test #81 or reference a location in the FSAR that provides the values for drop time. In addition, the test should clearly indicate that the drop test involves a full-height drop.
NuScale Nonproprietary
- c. Acceptance criteria ii' of FSAR Tier 2, Table 14.2-81 (Test #81), specifies the arithmetic average of all CRA drop times are within TS limits. However, RG 1.68 and technical specification (TS) surveillance requirement (SR) 3.1.4.3 do not permit the use of arithmetic averages to satisfy CRA drop time testing. RG 1.68 also adds, those control rods for which the scram times fall outside the two sigma limit of the scram time data for all control rods should be retested a sufficient number of times (e.g., three times) to reasonably ensure proper performance during subsequent plant operations. Therefore, the applicant is requested to revise Test #81 acceptance criteria ii to ensure each CRA drop time is within specified limits and all control rod drop times fall within the two sigma limit per RG 1.68. In addition, the applicant is requested to revise Test #81 acceptance criteria ii to provide the specific drop time values or surveillance requirement(s) needed to ensure proper operation of the CRDS.
- d. RG 1.68 specifies, to the extent practical, testing should demonstrate control rod scram times at both hot zero power and cold temperature conditions, with flow and no flow conditions in the reactor coolant system as required to bound conditions under which scram might be required.
i.
FSAR Tier 2, Table 14.2-81 (Test #81) specifies that the CRA drop time testing is performed when the reactor coolant system (RCS) is at hot zero power (HZP).
However, the staff could not find a test of CRA drop times during cold temperature conditions. Therefore, the applicant is requested to provide a CRA drop time test for cold temperature conditions, or provide justification for how testing during only HZP is bounding and demonstrates the control rods will drop within the required time under cold temperature plant conditions.
ii.
FSAR Tier 2, Table 14.2-81 (Test #81), specifies that the CRA drop time testing is performed when the RCS is at HZP. However, the staff could not find a test of CRA drop times during flow conditions. Therefore, the applicant is requested to add CRA drop time test acceptance criteria to the FSAR Tier 2, Table 14.2-107 (Test #107) and Table 14.2-104 (Test #104), which involve reactor trips at 10-20% reactor thermal power and 100% reactor thermal power, respectively.
RG 1.68 Appendix A, A-5 Power Ascension Tests, item (g), specifies, Check rod scram times from data recorded during scrams that occur during the startup test phase to determine that the scram times remain within allowable limits. However, the staff could not find this item anywhere in FSAR 14.2, ITP. The applicant is requested to include item (g) in either FSAR 14.2.4 or the Startup Administration Manual (COL Item 14.2-2). If it is to be addressed in the Startup Administrative Manual it can be specified separately. Alternatively, the COL information item could specify that the Startup Administrative Manual will meet RG 1.68 guidelines.
NuScale Nonproprietary NuScale Response:
Part a. response Tier 2, Table 14.2-80 (Test #80) is a preoperational test abstract for the Control Rod Drive System (CRDS). As described in Sections 14.2.10 and 14.2.12, test abstracts provide the bases for detailed preoperational and startup test procedures. Detailed preoperational and startup test procedures are developed and submitted to the NRC by Combined License (COL) holders by no later than 60 days prior to the conduct of preoperational and startup testing. While test abstracts provide the acceptance criteria for satisfying the test objectives, the design detail at the time of test abstract development is insufficient to include numerical values for certain parameters.
Rod insertion and withdrawal rates meet design requirements as described in Section 3.9.4.1.
Verification that individual CRA position agrees with control rod demand position is not applicable. Therefore, the acceptance criterion to perform this verification was deleted (previously Test #80, Acceptance Criterion ii).The limit associated with individual CRA position indications within the associated group position is found in Technical Specifications as indicated by Test #80, Acceptance Criterion iii. The CRA fully withdrawn position is an unambiguous description that has no numerical value in its design details. References to specific numerical values to be used for the specific acceptance criteria will be provided by the COL holder during development of detailed test procedures.
Part b. response Tier 2, Table 14.2-81 (Test #81) is a preoperational test abstract for the CRA. As described in Sections 14.2.10 and 14.2.12, test abstracts provide the bases for detailed preoperational and startup test procedures. Detailed preoperational and startup test procedures are developed and submitted to the NRC by COL holders by no later than 60 days prior to the conduct of preoperational and startup testing. While test abstracts provide acceptance criteria for satisfying the test objectives, the design detail at the time of test abstract development is insufficient to include numerical values for certain parameters.
Rod drop times are consistent with bounding safety analysis (in Section 15.0) and in agreement with applicable Technical Specifications. Table 14.2-81 will be modified to include that the drop test involves a "full height" drop.
Part c. response Tier 2, Table 14.2-81 (Test #81) acceptance criterion ii. has been modified to remove the statement about "arithmetic average" and include the two sigma limit. Acceptance criterion ii.
states that "Each CRA drop time is within two sigma of the drop time data for all control rods, or
NuScale Nonproprietary has been verified within Technical Specification limits by a minimum of three additional performances of this test."
Part d. response Subpart i. response NuScale has added a new Tier 2 test abstract, Table 14.2-81A, Control Rod Assembly Ambient Temperature Full-Height Drop Time Test (Test #81A), to perform drop time testing at ambient temperature conditions, consistent with the Test Method and acceptance criteria of Test #81 at HZP.
Subpart ii. response As requested, NuScale has added an analysis of rod drop times to the Tier 2, Table 14.2-104, Reactor Trip from 100 Percent Power Test (Test #104). The analysis of rod drop times at 10-20% thermal power has been added to Tier 2, Table 14.2-107 Remote Shutdown Workstation Test (Test #107).
Last paragraph response During startup testing, NuScale has committed to perform rod drop time testing at ambient conditions (Test #81A), HZP (Test #81), 10-20% thermal power (Test #107), and 100% thermal power (Test #104). Technical specifications dictate the requirements and frequency for rod drop time testing. No additional wording was added to FSAR Section 14.2.4 or COL Item 14.2-2.
Impact on DCA:
New FSAR Table 14.2-81A has been added and Tables 14.2-75, 14.2-80, 14.2-81, 14.2-104 and 14.2-107 have been revised as described in the response above and as shown in the markup provided in this response.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-170 Draft Revision 2 RAI 04.06-2 Table 14.2-75: Initial Fuel Loading Precritical Test (Test #75)
Startup test is required to be performed for each NPM.
This test is performed after initial fuel loading but prior to initial criticality.
Test Objectives i.
Identify the sequence for precritical testing (after fuel load and prior to criticality).
ii.
The pre-critical tests are:
a.
Reactor Coolant System Flow Measurement Test (Test #77) b.
NuScale Power Module Temperatures Test (Test #78) c.
Primary and Secondary System Chemistry Test (Test #79) d.
Control Rod Drive System - Manual Operation, Rod Speed, and Rod Position Indication Test (Test #80) e.
Control Rod Assembly (CRA)Full-Height Drop Time Test (Test #81) f.
Control Rod Assembly Ambient Temperature Full-Height Drop Time Test (Test #81A) fg. Pressurizer Spray Bypass Flow Test (Test #82)
Prerequisites None Test Method i.
Identify the specific plant conditions required for each precritical test procedure to maintain technical specification operability.
ii.
Identify the prerequisites required for each precritical test procedure.
iii. Determine the test sequence for precritical testing based on technical specification requirements and test prerequisites.
Acceptance Criterion The sequence for precritical testing has been determined.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-175 Draft Revision 2 RAI 04.06-2 Table 14.2-80: Control Rod Drive System - Manual Operation, Rod Speed, and Rod Position Indication Test (Test #80)
Startup test is required to be performed for each NPM.
This test is performed after initial fuel loading but prior to initial criticality.
Test Objectives i.
Verify the ability to manually fully insert and fully withdraw individual control rod assemblies (CRAs) from the MCR.
ii.
Verify CRA rod position indications provide indication of rod movement.
iii. Verify individual CRA position indications are within the required number of steps of their associated group position.
iv. Verify the rod insertion and withdrawal speeds are within design limits.
Prerequisites i.
The core is installed.
ii.
The NPM is fully assembled.
iii. The RCS is at hot zero power (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).
iv. All RCS temperatures satisfy the minimum technical specification temperature for criticality.
v.
The nuclear instrumentation system is calibrated and operable.
vi. The SDM is within the limits specified in the core operating limits report.
Test Method i.
Individually withdraw and insert each shutdown bank and regulating bank from the MCR a sufficient number of steps to verify that the individual CRA positions are within the required number of steps of their group position as required by TS.
Only the tested bank will be withdrawn. All other banks are fully inserted. Repeat the test until all shutdown banks and regulating banks are tested.
ii.
With all shutdown and regulating banks fully inserted, fully withdraw and then fully insert one CRA. Repeat these steps until all CRAs are tested.
Acceptance Criteria i.
All CRAs can be individually fully withdrawn and fully inserted from the MCR.
ii.
Individual CRA positions agree with the control rod demand position within design limits for the full range of CRA travel.
iiiii. Individual CRA position indications are within the number of steps of their associated group position as required by TS.
iviii.The CRA insertion and withdrawal speeds are within the design limits identified in Section 3.9.4.1.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-176 Draft Revision 2 RAI 04.06-2 Table 14.2-81: Control Rod Assembly Full-Height Drop Time Test (Test #81)
Startup test is required to be performed for each NPM.
This test is performed after initial fuel loading but prior to initial criticality.
Test Objective Verify each CRA satisfies the CRA drop time acceptance criteria for RCS flow at 0% reactor thermal power.
Prerequisites i.
The core is installed.
ii.
The NPM is fully assembled.
iii. The RCS is at hot zero power (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).
iv. All RCS temperatures satisfy the minimum technical specification temperature for criticality.
v.
The nuclear instrumentation system is calibrated and operable.
vi. The SDM is within the limits specified in the core operating limits report.
vii. A CRA drop time acceptance criteria for 0% thermal reactor power has been developed and is in agreement with the technical specification CRA drop time surveillance requirement.
Test Method i.
Fully Wwithdraw each individual CRA.
ii.
Interrupt the electrical power to the associated CRDM.
iii. Measure the CRA drop time.
Acceptance Criteria i.
Each CRA drop time is less than or equal to the CRA drop time acceptance criteria for HZPwithin Technical Specification limits.
ii.
The arithmetic average of all CRA drop times is within TS limitsEach CRA drop time is within two sigma of the drop time data for all control rods, or has been verified within Technical Specification limits by a minimum of three additional performances of this test.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-177 Draft Revision 2 RAI 04.06-2 Table 14.2-81a: Control Rod Assembly Ambient Temperature Full-Height Drop Time Test (Test #81A)
Startup test is required to be performed for each NPM.
This test is performed after initial fuel loading but prior to initial criticality.
Test Objective Verify each CRA satisfies the CRA drop time acceptance criteria for RCS at ambient temperature.
Prerequisites i.
The core is installed.
ii.
The NPM is fully assembled.
iii. The RCS is at ambient temperature.
iv. The nuclear instrumentation system is calibrated and operable.
v.
The SDM is within the limits specified in the core operating limits report.
Test Method i.
Fully withdraw each individual CRA.
ii.
Interrupt the electrical power to the associated CRDM.
iii. Measure the CRA drop time.
Acceptance Criteria i.
Each CRA drop time is within Technical Specification limits.
ii.
Each CRA drop time is within two sigma of the drop time data for all control rods, or has been verified within Technical Specification limits by a minimum of three additional performances of this test.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-200 Draft Revision 2 RAI 04.06-2 Table 14.2-104: Reactor Trip from 100 Percent Power Test (Test #104)
Startup test is required to be performed for each NPM.
This test is performed at 100 percent reactor thermal power.
Test Objectives i.
Verify the ability of the NPM to sustain a reactor trip from 100% reactor thermal power and automatically cool the RCS to mode 3 (all RCS temperatures < 420 °F).
ii.
Assess the dynamic response of the plant to the reactor trip.
iii. Verify each fully withdrawn CRA satisfies the CRA drop time acceptance criteria at full flow conditions.
Prerequisites i.
The NPM is operating in a steady-state condition at full reactor thermal power.
ii.
The plants electrical distribution system is aligned for normal operation.
Test Method i.
Manually trip the reactor from the MCR.
ii.
Measure the drop time for each fully withdrawn CRA.
iii. Allow the RCS to cool to mode 3.
Acceptance CriterionCriteria i.
The reactor trips.
ii.
The CIVs close.
iii. The decay heat removal valves open.
iv. The turbine generator bypass valve operates to prevent opening of the main steam safety valve.
v.
The turbine speed does not exceed overspeed design limits.
vi. The reactor vent valves do not open.
vii. Water hammer indications a.
Audible indications of water hammer are not observed b.
No damage to pipe supports or restraints c.
No damage to equipment d.
No equipment leakage as a result of the reactor trip viii. The RCS cools to a stable condition in mode 3 without operator intervention.
ix. Each fully withdrawn CRA drop time is within Technical Specification limits.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-203 Draft Revision 2 RAI 04.06-2 Table 14.2-107: Remote Shutdown Workstation Test (Test #107)
Startup test is required to be performed for each NPM.
This test is performed at approximately 10 - 20 percent reactor thermal power.
Test Objectives i.
Verify the NPM safety-related controls can be disabled at the remote shutdown station.
ii.
Verify the NPM nonsafety-related controls are functional at the remote shutdown station.
III. Verify each fully withdrawn CRA satisfies the CRA drop time acceptance criteria with the reactor operating at 10 - 20%
reactor thermal power.
Prerequisites i.
Communication exists between the MCR and the remote shutdown station.
ii.
The reactor is operating in a steady-state condition at 10 - 20% reactor thermal power.
Test Method i.
Using the appropriate operating procedure, the operator manually trips the reactor under test before leaving the MCR.
ii.
Measure the drop time for each fully withdrawn CRA.
iii. Using the appropriate operating procedure, the operator uses manual switches in the remote shutdown station to isolate the module protection system manual actuation switches, override switches, and the enable nonsafety control switches for each nuclear power modules module protection system in the MCR to prevent spurious actuation of equipment due to fire damage.
Acceptance Criteria i.
An operator verifies that the module protection switch controls in the MCR have been disabled.
The displays in the remote shutdown station verify the following NPM status:
ii.
The reactor is tripped.
iii. All CIVs are closed.
iv. The DHRS actuation valves are open.
v.
All RCS temperatures cool to less than 420°F (mode 3, safe shutdown) without operator action.
vi. Safety-related components cannot be operated from the remote shutdown station.
vii. The nonsafety-related controls in the remote shutdown station controls can be used to place the plant in a configuration specified by the appropriate operating procedure.
viii. Each fully withdrawn CRA drop time is within Technical Specification limits.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-205 Draft Revision 2 RAI 04.06-2 Table 14.2-109: List of Test Abstracts Test Number System Abbreviation Test Abstract 1
SFPCS Spent Fuel Pool Cooling System 2
PCUS Pool Cleanup System 3
RPCS Reactor Pool Cooling System 4
PSCS Pool Surge Control System 5
PLDS Pool Leakage Detection System 7
RCCWS Reactor Component Cooling Water System 8
CHW Chilled Water System 9
ABS Auxiliary Boiler System 10 CWS Circulating Water System 11 SCW Site Cooling Water System 12 PWS Potable Water System 13 UWS Utility Water System 14 DWS Demineralized Water System 15 NDS Nitrogen Distribution System 16 SAS Service Air System 17 IAS Instrument Air System 18 CRHS Control Room Habitability System 19 CRVS Normal Control Room HVAC System 20 RBVS Reactor Building HVAC System 21 RWBVS Radioactive Waste Building HVAC System 22 TBVS Turbine Building Ventilation 23 RWDS Radioactive Waste Drain System 24 BPDS Balance-of-Plant Drains 25 FPS Fire Protection System 26 FDS Fire Detection 27 MSS Main Steam 28 CFWS Feedwater System 29 FWTS Feedwater Treatment 30 CPS Condensate Polisher Resin Regeneration System 31 HVD Heater Vents and Drains 32 CARS Condenser Air Removal System 33 TGS Turbine Generator 34 TLOS Turbine Lube Oil System 35 LRWS Liquid Radioactive Waste System 36 GRWS Gaseous Radioactive Waste System 37 SRWS Solid Radioactive Waste System 38 CVCS Chemical and Volume Control System 39 BAS Boron Addition System 40 MHS Module Heatup System 41 CES Containment Evacuation System 42 CFDS Containment Flooding and Drain System 43 CNTS Containment System 44 CRDS Control Rod Drive System Flow-Induced Vibration 45 RVI Reactor Vessel Internals Flow-Induced Vibration 46 RCS Reactor Coolant System 47 ECCS Emergency Core Cooling System 48 DHRS Decay Heat Removal System 49 ICIS In-core Instrumentation
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-206 Draft Revision 2 50 MAE Module Assembly Equipment 51 FHE Fuel Handling Equipment System 52 RBC Reactor Building Cranes 53 PSS Process Sampling System 54 EHVS 13.8 kV and Switchyard System 55 EMVS Medium Voltage AC Electrical Distribution System 56 ELVS Low Voltage AC Electrical Distribution System 57 EDSS Highly Reliable DC Power System 58 EDNS Normal DC Power System 59 BPSS Backup Power Supply 60 PLS Plant Lighting System 61 MCS Module Control System 62 PCS Plant Control System 63 MPS Module Protection System 64 PPS Plant Protection System 65 NMS Neutron Monitoring System 66 SDIS Safety Display and Indication 67 RMS Fixed Area Radiation Monitoring System 68 COMS Communication System 69 SMS Seismic Monitoring System 70 HFT Hot Functional Testing 71 MAEB Module Assembly Equipment Bolting 72 SG Steam Generator Flow-Induced Vibration 73 N/A Security Access Control 74 N/A Security Detection and Alarm 75 N/A Initial Fuel Loading Precritical 76 N/A Initial Fuel Load 77 N/A Reactor Coolant System Flow Measurement 78 N/A NuScale Power Module Temperatures 79 N/A Primary and Secondary System Chemistry 80 N/A Control Rod Drive System-Manual Operation, Rod Speed, and Rod Position Indication 81 N/A Control Rod Assembly Full-Height Drop Time 81A N/A Control Rod Assembly Ambient Temperature Full-Height Drop Time Test 82 N/A Pressurizer Spray Bypass Flow 83 N/A Initial Criticality 84 N/A Post-Critical Reactivity Computer Checkout 85 N/A Low Power Test Sequence 86 N/A Determination of Zero-Power Physics Testing Range 87 N/A All Rods Out Boron Endpoint Determination 88 N/A Isothermal Temperature Coefficient Measurement 89 N/A Bank Worth Measurement 90 N/A Power-Ascension 91 N/A Core Power Distribution Map 92 N/A Nuclear Monitoring System Power Range Flux Calibration 93 N/A Reactor Coolant System Temperature Instrument Calibration 94 N/A Reactor Coolant System Flow Calibration 95 N/A Radiation Shield Survey 96 N/A Reactor Building Ventilation System Capability 97 N/A Thermal Expansion 98 N/A Control Rod Assembly Misalignment Table 14.2-109: List of Test Abstracts (Continued)
Test Number System Abbreviation Test Abstract