RAIO-0917-55893, LLC Response to NRC Request for Additional Information No. 97 (Erai No. 8878) on the NuScale Design Certification Application
| ML17262B208 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 09/19/2017 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| References | |
| RAIO-0917-55893 | |
| Download: ML17262B208 (9) | |
Text
RAIO-0917-55893 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com September 19, 2017 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.
97 (eRAI No. 8878) on the NuScale Design Certification Application
REFERENCE:
U.S. Nuclear Regulatory Commission, "Request for Additional Information No.
97 (eRAI No. 8878)," dated July 21, 2017 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. 8878:
10.04.05-1 10.04.05-2 10.04.05-3 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 Darrell Gardner at 980-349-4829 or at dgardner@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution:
Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Demetrius Murray, NRC, OWFN-8G9A (QFORVXUH 1X6FDOH5HVSRQVHWR15&5HTXHVWIRU$GGLWLRQDO,QIRUPDWLRQH5$,1R
Sincerely, Zackary W. Rad Di t
R l t Aff i
RAIO-0917-55893 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. 8878
NuScale Nonproprietary Response to Request for Additional Information Docket No.52-048 eRAI No.: 8878 Date of RAI Issue: 07/21/2017 NRC Question No.: 10.04.05-1 GDC 4 10 CFR 52.47(a)(2) requires that a standard design certification applicant provide a description and analysis of the structures, systems, and components (SSCs) of the facility, with emphasis upon performance requirements, the bases, with technical justification therefor, upon which these requirements have been established, and the evaluations required to show that safety functions will be accomplished.
10 CFR 52.47(c)(2) requires that a standard design certification of a nuclear power reactor design that uses simplified, inherent, passive, or other innovative means to accomplish its safety functions must provide an essentially complete nuclear power reactor design except for site-specific elements such as the service water intake structure and the ultimate heat sink, and must meet the requirements of 10 CFR 50.43(e).
GDC 4 requires, in part, that SSCs important to safety be appropriately protected against dynamic effects, including the effects of discharging fluids. According to SRP 10.4.5, the requirements of GDC 4 are met when the circulating water (CW) system design includes provisions to accommodate the effects of discharge water (e.g., flooding).
The staff was unable to determine whether the pipe rupture flooding analysis accounts for the external portion of the circulating water system (CWS) piping (e.g., piping running between the cooling tower(s) and turbine generator buildings (TGBs)). In addition, the staff was unable to verify whether the grade slope is sufficient to carry this CWS flooding water away from the buildings housing SSCs as indicated in FSAR Tier 2, Subsection 10.4.5.3.
The applicant is requested to provide additional CWS pipe rupture flooding information including any supporting figures and drawings. The FSAR is to be modified accordingly.
NuScale Response:
SRP 10.4.5,Section II.1 indicates that the requirements of GDC 4 are met when flooding of
NuScale Nonproprietary safety-related areas that result from a malfunction or a failure of a component or piping of the CWS does not have unacceptable adverse effects on the functional performance capabilities of safety-related systems or components.
FSAR Section 10.4.5.3 indicates that the turbine building (TB) does not contain safety-related systems or components. Even if the grade slope is not sufficient to carry CWS flooding water away from the TB, since the TB contains no safety-related systems or components there is not a possibility of damage to safety-related SSCs in the TB resulting from CWS flooding water.
The reactor building (RXB) and control building (CRB) are the only safety-related buildings on a NuScale site. The RXB contains SSCs important to safety and the CRB supports the module protection system by housing and providing structural support. FSAR Section 3.4.1.4 notes that flooding of the RXB or CRB caused by external sources does not occur and that failure of equipment outside the CRB and RXB cannot cause internal flooding. Flooding exterior to the RXB and CRB would not impact the ability of the plant to safely shutdown as the structures are sealed, penetrations below grade are minimized as described in FSAR Section 3.4.2.1, and the passive safety system design of the reactor module does not require any water systems outside of the ultimate heat sink pool to provide shutdown cooling. Although a specific external CWS pipe rupture analysis has not been performed, the COL applicant is required to demonstrate the site is properly graded to prevent localized flooding as discussed in FSAR Table 2.0-1, Hydrologic Engineering subsection.
Therefore, the FSAR content described above demonstrates that flooding from a CWS pipe rupture external to the turbine building will not impact a safety-related area. Since a safety-related area will not be affected, the functional performance capabilities of safety-related systems or components in that area cannot be affected, which satisfies the acceptance criteria of SRP 10.4.5,Section II.1.
Impact on DCA:
There are no impacts to the DCA as a result of this response.
NuScale Nonproprietary Response to Request for Additional Information Docket No.52-048 eRAI No.: 8878 Date of RAI Issue: 07/21/2017 NRC Question No.: 10.04.05-2 Hydraulic Transients & Expansion Joint 10 CFR 52.47(a)(9) governs the technical content of a design certification application and requires that, for applications for light-water cooled nuclear power plants, an evaluation of the standard plant design against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. It also states that, where a difference exists, the evaluation shall discuss how the proposed alternative provides an acceptable method of complying with the Commission's regulations, or portions thereof, that underlie the corresponding SRP acceptance criteria.
GDC 4 requires, in part, that SSCs important to safety be appropriately protected against dynamic effects, including the effects of discharging fluids. According to SRP 10.4.5, the requirements of GDC 4 are met when the circulating water (CW) system design includes provisions to accommodate the effects of discharge water. The SRP further states that means should be provided to detect leakage in the CW system in order not to adversely affect when there is failure of a component (e.g., expansion joint) or piping in the CW system.
SRP 10.4.5,Section III.1 states:
Although the circulating water system is not safety related, a failure of this system, or any of its components, may affect a safety-related component or system. Since large quantities of water flow through the circulating water system (CWS), a leak or break in a component or pipe or expansion joint failure could cause severe and unacceptable flooding of adjacent areas. The reviewer verifies that the design includes provisions to minimize hydraulic transients and their effect upon the functional capability and the integrity of system components. In evaluating the effects of the failure of an expansion joint, the reviewer assumes that the butterfly valve(s) are not available to isolate CWS flow out of the failed expansion joint unless the valve(s) have been designed to safety-grade requirements. The reviewer analyzes the descriptions and drawings in the SAR and determines that provisions are incorporated in the design to prevent unacceptable flooding of areas containing safety-related equipment or to mitigate the consequences of flooding.
NuScale Nonproprietary In the review of FSAR Tier 2, Section 10.4.5, the staff could not find any provision to meet the GDC 4 criteria, as it relates to dynamic effects such as hydraulic transients (e.g., water hammer), during plant startup and shutdown, normal operation, and accident conditions. Also lacking from the FSAR was information related to the failure of CWS expansion joints and resulting floods.
The applicant is requested to provide additional CWS expansion joint failure and hydraulic transient (e.g., water hammer) information including any supporting figures and drawings. The FSAR is to be modified accordingly.
NuScale Response:
SRP 10.4.5,Section II.1 indicates that the requirements of GDC 4 are met when flooding of safety-related areas that result from a malfunction or a failure of a component or piping of the CWS does not have unacceptable adverse effects on the functional performance capabilities of safety-related systems or components.
FSAR Section 10.4.5.3 indicates that the turbine building (TB) does not contain safety-related systems or components. Even if a CWS expansion joint were to fail, either from within or outside the TB, since the TB contains no safety-related systems or components there is not a possibility of damage to safety-related SSCs in the TB resulting from a CWS expansion joint failure.
FSAR Section 10.4.5.3 notes that water from a CWS expansion joint leak would drain through the TB doors and vent openings, and then away from other site buildings as controlled by the site grading. Additionally, the reactor building (RXB) and control building (CRB) are the only safety-related buildings on a NuScale site. The RXB contains SSCs important to safety and the CRB supports the module protection system by housing and providing structural support. FSAR Section 3.4.1.4 notes that flooding of the RXB or CRB caused by external sources does not occur and that failure of equipment outside the CRB and RXB cannot cause internal flooding.
Flooding exterior to the RXB and CRB would not impact the ability of the plant to safely shutdown as the structures are sealed, penetrations below grade are minimized as described in FSAR Section 3.4.2.1, and the passive safety system design of the reactor module does not require any water systems outside of the ultimate heat sink pool to provide shutdown cooling.
Therefore, the FSAR content described above demonstrates that flooding from a CWS expansion joint failure will not impact a safety-related area. Since a safety-related area will not be affected, the functional performance capabilities of safety-related systems or components in that area cannot be affected, which satisfies SRP 10.4.5,Section II.1.
Impact on DCA:
There are no impacts to the DCA as a result of this response.
NuScale Nonproprietary Response to Request for Additional Information Docket No.52-048 eRAI No.: 8878 Date of RAI Issue: 07/21/2017 NRC Question No.: 10.04.05-3 Leakage 10 CFR 52.47(c)(2) requires that a standard design certification of a nuclear power reactor design that uses simplified, inherent, passive, or other innovative means to accomplish its safety functions must provide an essentially complete nuclear power reactor design except for site-specific elements such as the service water intake structure and the ultimate heat sink, and must meet the requirements of 10 CFR 50.43(e).
SRP 10.4.5,Section III.2 states that the circulating water system (CWS) design should have the capability to detect leaks and to secure the system quickly and effectively.
In the review of FSAR Tier 2, Section 10.4.5, the staff was unable to find information addressing the CWS capability of detecting and controlling leaks.
The applicant is requested to provide additional CWS leak detection information including any supporting figures and drawings. The FSAR is to be modified accordingly.
NuScale Response:
SRP 10.4.5,Section II.1 indicates that the requirements of GDC 4 are met when flooding of safety-related areas that result from a malfunction or a failure of a component or piping of the CWS does not have unacceptable adverse effects on the functional performance capabilities of safety-related systems or components.
The NuScale design does not provide for direct CWS leak detection. Although not credited, large leaks would be detected by a loss of main condenser vacuum which is monitored during normal operation. FSAR Section 10.4.5.3 has been revised to include this additional detail.
FSAR Section 10.4.5.3 indicates that the turbine building (TB) does not contain safety-related systems or components. Even if CWS leakage were to occur and progress undetected to the level of flooding, since the TB contains no safety-related systems or components there is not a
NuScale Nonproprietary possibility that a safety-related structure, system, or component (SSC) will be affected.
FSAR Section 10.4.5.3 notes that water from a CWS expansion joint leak would drain through the TB doors and vent openings, and then away from other site buildings as controlled by the site grading. Additionally, the reactor building (RXB) and control building (CRB) are the only safety-related buildings on a NuScale site. The RXB contains SSCs important to safety and the CRB supports the module protection system by housing and providing structural support. FSAR Section 3.4.1.4 notes that flooding of the RXB or CRB caused by external sources does not occur and that failure of equipment outside the CRB and RXB cannot cause internal flooding.
Flooding exterior to the RXB and CRB would not impact the ability of the plant to safely shutdown as the structures are sealed, penetrations below grade are minimized as described in FSAR Section 3.4.2.1, and the passive safety system design of the reactor module does not require any water systems outside of the ultimate heat sink pool to provide shutdown cooling.
Therefore, the FSAR content described above demonstrates that flooding resulting from a large CWS leak will not impact a safety-related area. Since a safety-related area will not be affected, the functional performance capabilities of safety-related systems or components in that area cannot be affected, which satisfies the acceptance criteria of SRP 10.4.5,Section II.1.
Impact on DCA:
FSAR Section 10.4.5.3 has been revised as described in the response above and as shown in the markup provided in this response.
NuScale Final Safety Analysis Report Other Features of Steam and Power Conversion System Tier 2 10.4-22 Draft Revision 1 A trip of one of the six NPMs served by the circulating water subsystem does not affect the other five. The design requires one CWS pump for every two main condensers that are online to provide full flow.
To accommodate flooding due to rain or makeup valve malfunction, basin overflow is piped to the single point discharge to the environment.
The circulating water pumps are not required during a DBA.
10.4.5.3 Safety Evaluation The CWS serves no safety function, is not credited for mitigation of a DBA, and has no safe shutdown functions.
General Design Criterion 2 was considered in the design of the circulating water system. No safety-related structures, systems, or components are affected by this system from the effects of natural phenomena such as earthquakes. The design and layout of the CWS include provisions that ensure that a failure of the system will not adversely affect the functional performance of safety-related systems or components.
The CWS system meets RG 1.29 in that the CWS is not located in areas that contain safety-related components and is not required to operate during or after an accident.
General Design Criterion 4 was considered in the design of the circulating water system. The design of the CWS provides protection of safety-related SSC from the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents.
RAI 10.04.05-3 The TGB does not contain safety-related equipment; thereby eliminating the possibility of damage as a result of CWS line break. Large circulating water system (CWS) leaks due to pipe failures will be indicated in the control room by a loss of main condenser (MC) vacuum. MC vacuum is a parameter that is monitored during normal operation (Section 10.4.2.2.3). Water from a circulating water pipe or expansion joint leak would drain through the building doors and vent openings, and then away from other buildings as controlled by the site grading. The soil around the TGB and the cooling towers is sloped away from structures, thus a failure of the water basin of the cooling towers has no effect on other structures. The flooding evaluation is addressed in Section 3.4.
General Design Criterion 5 was considered in the design of the circulating water system. The sharing of the CWS across six NPMs does not impair the ability of the other NPMs to perform their safety functions. The use of common CWS equipment to accomplish the cooling of six or less condensers does not have a significant effect on system availability and operability as described in Section 10.4.5.2.3.
General Design Criterion 60 was considered in the design of the circulating water system. Consistent with GDC 60, the CWS design controls radioactive material releases to the environment. The CWS is anticipated to contain negligible quantities of radioactive contaminants during power operation and during shutdown. Blowdown