ML20072E888

Forwards Response to NRC 940408 RAIs Re Simplified BWRs Design.Rev 1 to 25A5013, Isolation Condenser Sys

ML20072E888
Person / Time
Site:	05200004
Issue date:	08/02/1994
From:	Marriott P GENERAL ELECTRIC CO.
To:	Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20072E891	List: ML20072E888 ML20072E906
References
MFN-090-94, MFN-90-94, NUDOCS 9408230134
Download: ML20072E888 (20)

Text

,

GE Nuclear Energy amenecme comury I15 Cartou Avenue. Shr Jaw CA 95125 August 2,1994 MFN No. 090-94 Docket No STN 52-004 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Richard W. Borchardt, Director Standardization Project Directorate

Subject:

NRC Requests for Additional Information (RAIs) on the Simplified Boiling Water Reactor (SBWR) Design

References:

1. Transmittal of Requests for Additional Information (RAls)

Regarding the SBWR Design, Letter from M. Malloy to P. W. Marriott dated April 8,1994.

2. MFN No. 084-94, RAls In Process at GE, Letter from P. W. Marriott to R. W. Borchardt, June 15,1994.

The Reference 1 letter requested additional information regarding the SBWR Design. In partial fulfillment of this request, GE is submitting Attachment I to this letter which transmits the responses to the following RAls in the 440.7 -

440.58 series which focus on thermal-hydraulic testing and analysis activities:

440.13 440.15- 440.17 440.32- 440.33 440.36- 440.58 The GE response to the remaining RAls in this series will be provided following resumption of NRC's review of the SSAR in design-related areas.

Sincerely,

) L W . jfnff *

) P. W. Marriott, Manager O6 Advanced Plant Technologies M/C 781, (408) 9254i948 Attachment 1, " Responses to NRC RAls" cc: M. Malloy, Project Manager (w/2 copies of Attachment 1)

F. W. Hasselberg, Project Manager (w/l copy of Attachment 1) 2ggg M

! RAI Number: 440.13 Question:

In SSAR Section 1.2.2.8.1 (Nuclear Fuel), delete the following statement which is not correct: " Fuel design for the SilWR Standard Plant is not within the scope of the certified design."

In SSAR Section 1.2.2.8.2 (Fuel Channel), delete the following statement which '

is not correct: " Fuel channel design for the SilWR Standard Plant is not within l the scope of the certified design."  !

In SSAR Section 1.2.2.8.3 (Control Rod), delete the following statement which is not correct: " Control rod design for the SilWR Standard Plant is not within the l scope of the certified design." l GE Response: 1 These changes are incorporated in the attached revised SilWR SSAR pages.

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25A5113 R:v. A

'S~WR standan! Safety Analysis Report l

1 The NEMS has no access to the safety-related data base; however safety-related data can I

be read by the NEMS on the optically-isolated memor7 portions of the Essential Multiplexing System (EMS) Local Multiplexing Units (LMU). This data can be read by any NEMS multiplexing units that is configured to do so. The NEMS cannot write data to any portion of the EMS.

The NEMS consists of two types of multiplexing units: Local Multiplexing Units (LMU),

and Control Room Multiplexing Units (CMU) connected via fiber optic cables. The NEMS also includes network gateways which allow transfer of data between data highway systems.  ;

Throughout the plant, LMUs are located in local plant areas to acquire sensor data and transmit this data to the any equipment that requires it. The LMUs also receive processed signals from the control room for command of control system actuators.

CMUs are located in the control room to transmit and receive data for the logic I processing units of the plant control systems.

Allinterconnections are fiber optic data links. Within each NEMS highway, the system uses redundant links for greater reliability.

There are a number of NEMS highway systems that are routed throughout the plant.

These systems all have CMUs located in the main control room. Gateways connect the multiple NEMS highway systems to allow for transfer of data between NEMS highway systems.

1.2.2.8 Nuclear Fuel 1.2.2.8.1 Nuclear Fuel FucklengrJor-the41RVFrStandard44 ant 4s not-wnhin-th@edfieddesgnr It is intended that the specific fuel to be used in any facility which has adopted the certified design be in compliance with U.S. NRC approved fuel design criteria. This strategy is intended to permit future use of enhanced / improved fuel designs as they become available. However, this approach is predicated on the assumption that future fuel designs will be extensions of the basic fuel technology that has been developed for boiling water reactors. Key characteristics of this established InVR fuel technology are:

a Uranium oxide based fuel pellets; a Zirconium-based (or equivalent) fuel cladding; a All material selected on the basis of BWR operating conditions; a Multi-rod fuel bundles in an N lattice; and 1.2-62 General Plant Description - Amendment 1

25AS113 Rsv. A

'S"WR Srndard Safety Analysis Report 1.2.2.8.2 Fuel Channel Fuel-ehanne! design for4he41RVR-imoH4dnin-the scope of We cerdfied41esign,It is intended that the specific fuel channel to be used in any facility which has adopted the certified design be in compliance with U.S. NRC approved fuel channel design criteria.

This strategy is intended to pennit future use of enhanced / improved fuel channel designs as they become available. However, this approach is predicated on the assumption that future fuel channel designs will be extensions of the basic technology that has been developed for boiling water reactors. The key charactedsdc of this established BWR fuel channel technology is the use of zirconium-based (or equivalent) fuel channels which preclude cross-flow in the core region.

The following is a summary of the principal requirements which must be met by the fuel channel supplied to any facility using the certified design:

)

m The material of the fuel channel shall be shown to be compatible with the reactor l environment a The channel will be evaluated to ensure that channel deflection does not preclude control rod drive operation.

m The effects of channel bow will be included in the fuel rod critical power evaluations.

1.2.2.8.3 Control Rod Quatrol-red <lesign-for-die-SBWhiet-widdn-die-scope of 6e certified design. It is intended that the specific control rod to be used in any facility which has adopted the certified design be in compliance with U.S. NRC approved control rod design criteria.

This strategy is intended to permit future use of enhanced / improved control rod designs as they become available. However, this approach is predicated on the assumption that future control rod designs will be extensions of the basic technology that has been developed for boiling water reactors. Key characteristics of this established BWR control rod technology are:

a Control rods perfonn dual functions of power distribution shaping and reactivity control, a The control rod has a crucifonn cross-sectional envelope shape.

m The control rod has a coupling at the bottom for attachment to the CRD.

m The control rod has an upper bail handle for transporting.

k 1.2-66 General Plant Description - Amendment 1

l l

l RAI Number: 440.15 Question:

In SSAR Section 3.1.3.7 (page 3.1-30), Criteria 26, Reactivity Control System Redundancy and Capability, and SSAR Section 3.1.3.8 (page 3.1-31), Criteria 27, Lombined Reactivity Control Systems Capability, add Reference 9.3.5, Standby I.,gm,d i Control System, to the list of SSAR chapters / sections.

GE Response:

GE will make tilis change to tiie SBWR SSAR.

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l RAI Number: 440.16 l l

Question:

GE's response to RAI SRXB 7 was not satisfactory. The staff does not agree with the GE statement that "There are no requirements for a safety-related high pressure injection system.. " General Design Criteria (GDC) 33 of 10 CFR 50 Appendix A requires a safety-related system for protection against small breaks in the reactor coolant pressure boundary and the SBWR design does not include a safety-related injection system to satisfy the GDC 33. (Reference SSAR Section 3.1.4.4, Reactor Coolant Make Up.)

GE Response:

The SilWR does not have a safety-related high pressure makeup system. The SBWR, as a passive plant, takes exception to GDC 33. The SBWR relies on the ADS and GDCS system for core flooding if non-safety related makeup from the CRD system is unavailable. The safety advantages of passive technology are

• cxpected to provide a lower probability of core damage than non-passive technology in the case of a small break accident.

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RAI Number: 440.17 Question:

In SSAR Table 3.2-1 (page 3.2-16),J11, add control rods to the list.

GE Response:

GE will make this change to the SilWR SSAR.

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. i RAI Number: 440.32 Question:

What is meant by " overshoot the relief valve setpoint" in SSAR Section 5.2.2.3.3, Safety / Relief Valve Capacity? Figure 5.2-6 shows that peak vessel pressure is independent of valve capacity.110w can the peak pressure be completely independent of the valve capacity? liow many (minimum number of) safety-relief valves are required to meet the ASME Iloiler and Pressure Vessel Code limit of 1375 psig? Explain in detail the significance of the operation oflow and high set point valves shown in the figure.

GE Response:

Please see attached revised SSAR Section 5.2.2.3.3.

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25A5113 Rsv. A Standant Safety Analysis Report SBWR a Reclosure pressure setpoint (% of opening setpoint) bodi modes:

- maximum safety limit (used in analysis): 98

- minimum operational limit: 88 The opening and reclosure setpoints are assumed at a conservatively high level above the nominal setpoints. This is to account for initial setpoint errors and any instrument setpoint drift that might occur during operation. Typically, die setpoints in the analysis (on average) are assumed to be at least 1% above the actual nominal setpoints.

Conservative SRV response characteristics are also assumed; therefore, the analysis conservatively bounds all SRV operating conditions.

Safety / Relief Valve Capacity Sizing of the SRV capacity is based on establishing an adequate margin from the peak vessel bottom pressure to the vessel code limit (9.481 MPa gauge) in response to the reference transients.

The method used to determine total valve capacity is as follows.

Whenever the system pressure increases to the valve spring set pressure of a group of valves, these valves are assumed to begin opening and to reach full open at 103% of the valve spring set pressure. The lift chamcteristics assumed are shown in Figure 5.2-3.

5.2.2.3.3 Evaluation of Results Safety / Relief Valve Capacity The required SRV capacity is determined by analyzing the pressure rise from a turbine / generator trip with bypass failure transient with pressure scram. Results of this analysis are given in Figure 5.2-5. The peak vessel bottom pressure calculated is 8.73 MPa absolute (1266 psia), which is well below the acceptance limit of 9.481 MPa gauge (1375 psig). Figure 5.2-5 shows the MSIV isoladon transient with flux scram, which is slightly milder than the turbine trip. The pressurization is not dynamic and donnot+igniGeamiv+versiemhe+elief dve utpoint ceases to increase once a sincle relief valve onens. Figure 5.2-6 shows that peak vessel pressure is only a fimction of the ,

valve serpoint.

This is because the hicher steam volume-to-oower ratio of the SBWR causes the cressure ste prior to scram to be much lower than onemting BWRs f0.34 MPa/sec (38 nsi/sec) versus 0.57 MPa/sec (82 osi/sec) 1. After a scram. the nressure rates due to stored encrev release are corresnondincly lower. The neak nressure in these events is the relief valve setnoint because. at these low nressurization rates. with larce marcin to the SRV setnoint. the cressure increase is efTectively terminated by any sincie relief valve opniinc. As shown on Ficure 5.2 6. the neak nressure is the same when 1. 2. 3. or 4 low g.1 point valves are available. If all low setnoint valves are assumed to fail. the neak 5.2-8 Integrity of Reactor Coolant Pressure Boundary- Amendment 1

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2SAS113 Rsv. A standant sanrty Analysis Report SBWR oressure steos un to a higher value.which is essentially the hich oressure setpoint.

nrovided any hich setnoint valve is available.

Pressure Drop in inlet and Discharge Pressure drop in the piping from the reactorvessel to the valves is taken into account in calculating the maximum vessel pressures. Pressure drop in the discharge piping to the suppression poolis limited by proper discharge line sizing to prevent backpressure on each SRV from exceeding 40% of the valve inlet pressure, thus assuring choked flow in the valve orifice and no reduction of valve capacity due to the discharge piping. Each SRV has its own separate discharge line.

5.2.2.3.4 System Reliability The system is designed to satisfy the requirements ofSecdon III of the ASME Code. The consequences of failure are discussed in Subsection 15.1.4 of this report.

5.2.2.4 Testing and Inspection Requirements The inspection and testing of applicable SRVs utilizes a quality assurance program which complies with Appendix B of 10CFR50.

The SRVs are tested at a suitable test facility in accordance with quality control procedures to detect defects and to prove operability prior to installation.The following tests are conducted:

(1) hydrostatic test at specified test conditions (ASME Code requirement based on design pressure and temperature);

(2) thermally stabilize the SRV to perform quantitative steam leakage testing at 1.03 MPa (150 psi) below the SRV nameplate value with an acceptance criterion not to exceed 0.45 kg/hr (1 lb/hr) leakage; (3) full flow SRV test for set pressures and blowdown where the valve is pressurized with saturated steam, with the pressure rising to the valve set pressure (during production testing the SRV is adjusted to open at the nameplate setpressure i1%); and l

(4) response time test where each SRV is tested to dem.onstrate acceptable response time based on system requirements.

The valves are installed as received from the factory.The valve manufacturer certifies that design and performance requirements have been met.This includes capacity and blowdown requirements. The serpoints are adjusted, verified, and indicated on the valves by the vendor. Specified manual and automatic initiated signal for power actuation (relief mode) of each SRV is verified during the preoperational test program.

S.2-9 Integrity of Reactor Coolant Pressure Boundary - Amendment 1

RAI Number: 440.33 Question:

In SSAR Figure 5.2-1, add units to all the numerical values shown in the figure.

GE Response:

All units of numeric values in Figure 5.2-1 are nun and there is a note to that effect on the attached revised figure.

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25A5113 Rsv. A S WR standardsa% Analysis Report inso

= 15750  :

CONTAINENT BOUNDARY (THE DARK BLACK Lpi E) l N

i , 4 SRVs PER STEAMLINE i

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I N ST ~ 3154-+  :

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f SUPPRESSION POOL

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Note: All units are measured in mm Figure 5.2-1 Safety / Relief Valve Schematic Elevation Integrity of Reactor Coolant Pressure Boundary- Amendment 1 5.2-61

RAI Number: 440.36 Question:

Provide a diagram showing the isolation condenser system (ICS) design parameters: Pressures, temperature, and flow rates. Submit the process flow diagram for the ICS.

GE Response:

The process flow diagram 107E6073 Rev.1 is attached.

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RAI Number: 440.37 Question:

The isolation condenser for the SilWR is a vertical heat exchanger which is significantly different from the IC in operating plants. Provide a detailed description and drawing of the SilWR IC.

GE Response: ,

The Isolation Condenser System (ICS) specification 25A5013 Rev.1 (attached) contains a detailed description of the isolation condenser system. A copy of the Isolation Condenser (IC) drawing is attached.

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