ML20044G858

Forwards Response to NRC 930128 Request for Addl Info on Simplified BWR Design

ML20044G858
Person / Time
Site:	05200004
Issue date:	05/28/1993
From:	Marriott P GENERAL ELECTRIC CO.
To:	Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
MFN-083-93, MFN-83-93, NUDOCS 9306040360
Download: ML20044G858 (25)

Text

s GE Nuclear Energy av., ow. o,,

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' May 28,1993

e m ca w s MFN No. 083-93 Docket STN 52-004 1

Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l

t Attention:

Richard Borchardt, Acting Director Standardization Project DireJorate

Subject:

NRC Requests for Additional Information (RAls) on the Simplified Boiling l

Water Reactor (SBWR) Design i

Reference:

Transmittal of Requests for Additional Information (RAls) for the SBWR Design, i

Letter from J. W. Thompson to P. W. Marriott January 28,1793 The reference requested additional information on the SBWR Des (n. As part of the response to this request, GE is submitting responses to the following RAls:

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1) RAls No. ECGB.14,.16, and.21,

2) RAI No. EELB.1,

3) RAIs No. EMCB.4,.7, and.10, u

4) RAls No. HHFB.0,.1,.4.1,.4.3, and.4.4,

5) RAI No. lilCB.8,

6) RAls No. RPEB.6, and.13,

7) RAls No. SPLB.ll, and.26, and

8) RAls No. SRXB.3,.5,.24,.28, and.45.

i Sincerely, s

P. W.

farriott, Manager Safety & Licensing M/C 444, (408) 925-6948 Enclosure.:

RAI Responses c,aoeam 9306040360'930528 g-PDR ADOCK 05200004 PDR Q

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T RAI Number: ECGil.14 Question:

During a, July 17,1991, meeting with GE, the stafT was informed that the SIlWR is mainly based on state-of-the-art light water reactor (LWR) technology of operating boiling water reactors (IlWRs) and AllWRs. Thus, only a limited amount of Jevelopment testing for certification of the GDCS, PCCS and depressurization valve (DPV) is needed.

Among those planned tests, only laboratory tests of compenent or scaled model were performed. Some planned full-scale tests for decay heat removal, mechanical performance of heat exchanger, isolation condenser, and their performance under integral system conditions will not be conducted until 1995. For those tests already performed, large uncertainties in heat transfer were reported. It appears that all thermal loads needed for design of mechanical components ard piping systems under passive operating and accident conditions ar e not yet been accurately defined. At this stage of the SIlWR design, how does GE ensure that the design loadings for mechanical componenis are adequate?

GE Response:

l Large uncertainties in heat transfer do not occur. Extensive beat transfer tests at UC llerkeley and MIT were conducted to develop a heat transfer correlation for use in the TRACG computer code. This model was then qualified against integrated system test data from the GIRAFFE test facility. See the response to SRXil.38,39,40 and SRXil.42 for additional information on this subject.

Design loading conditions and structural adequacy for new components such as the Passive Containment Cooler (PCC) and Isolation Condenser (IC) are established by a combination of the following:

1) Ily first defining the thermal, prc..u.c. and mechanical loading conditions which are predicted to occur during plant service. Transient and steady state performance analyses and seismic response spectra are used to define these conditions. Numbers of loading cycles are predicted or based on data from operating plants.

2) Designing the equipment for thermal, mechanical and pressure loads that bound, on the conservative side, the above loading conditions.

3) Confirming design adequacy by detailed and complete stress and fatigue evaluations.

4) In qualifying new equipment designs, tests may hc 4 me to confirm icading conditions or to show structural adequacy. For the PCC, full scale testing will be

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t conducted under conditions and operating times that greatly exceed those which are.

. predicted to occur in senice;(except for seismic loading effects which are included in j

the stress and fatigue evaluations above). For the IC, full scale cycle testing will be =

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followed by NDE (Non Destructive Examination) using ISI methods after approximately l

one third of the required IC operational cycles.

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5 RAI Number: ECGB.16 Question:

l in SSAR Subsection 3.9.3.4 and Table 3.2-1, portions of the isolation condenser system (ICS) outside containment are identified as being classified as ASME Class 2 (QG I1).

j The staff agrees that this classification is consistent with the exceptions allowed in 10 ~

CFR 50.55a(c)(2). However, because portions of the ICS are subjected to reactor operating temperature and pressure continuously during operation, and because this system is used to transfer decay and residual heat from the reactor after it has been l

shutdown and isolated, it appears that some of the Class 2 steam supply and condensate piping and the condensers could be subjected to a significant number of thermal transients. Provide a description of the methodology to be used to account for the cffects of fatigue in this Class 2 portion of the ICS. In addition, in Figure 21.5.4-1, i

identify the exact location of the safety classification change from Class 1 to Class 2 in

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both the steam supply and condensate return piping.

GE Response:

Tier i Design Certification Document Figure 2.4.1-1 more clearly shows the exact transition between Class 1 and 2 as follows: The steam piping that penetrates the containment is Class I up to and including the flow restrictors shown on Figure 2.4.1-1, and the condensate drain line that penetrates the containment is Class I up to the two smaller branch line connections. Fatigue evaluation showing fatigue usage is withis acceptance limits will be done for the whole IC unit,i.e., both the Class 1 and Class 2 i

portions, according to the requirements of Class I components, ASME III, NB-3222.4 and NB-3222.5.

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RAI Number: ECGIl.21 Question:

In the SSAR, GE has committed that the MOV equipment specifications will require the incorporation of the results of either ir. situ or prototype testing with full flow and-differential pressure to verify the proper sizing and switch settings of the valves. GE also committed that all SilWR safety-related piping systems will incorporate provisions-for testing to demonstrate the operability of check valves under design basis j

conditions.

Ilased on operating experience, the staff has determined that a similar commitment is s

needed for the specifications for other povu operated valves to incorporate the results of either in plant or prototype testing to verify design basis capability.11ased on past experience with estimating thrust and torque requirements and other parameters for valve operation, the staff helieves that this assurance cannot be provided by analytical' approaches alone and will require that proper sizing and adjustment of other power-operated valves be verified by a generic ITAAC. GE shoald develop an acceptable i

generic ITAAC ihr demonstrating the capability of other power-operated valves.

GE Response:

Development of an acceptable generic ITAAC for demonstrating the capability of appropriate safety-related check valves and other safety-related power operated valves is underway at this time and will be added to the SilWR generic ITAACs as soon as the efrort is completed.

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RAI Number: EELB.1 5

'i Question:

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If not addressed in the forthcoming SSAR Section 1.8, " Interfaces for Standard Design,"

and Section 1.9, "Conformance with Standard Review Plan," then please provide an explanation and how the SBWR incorporates into the design the reliability assurance.

j program (RAP) that addresses the TS, insenice inspection / inservice testing programs, the maintenance programs, plant procedures, and the security program (see.

Commission paper SECY-894)13, " Design Requirements Related to the Evoluationary 1

Adv;mced Light Water Reactors (ALWRs)," for further details).

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i The objective of the D-RAP is as stated in section 17.3.

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e RAI Number: EMCB.4 i

l Question:

SSAR Section 5.3 Reactor Vessel SSAR Paragraph 5.3.1.2, "Special Procedures Used for Manufacturing and Fabrication,"

specifies maximum limits on copper, p,osphorous and sulfur for base and weld materials in the beltline region. The applicant must also include a maximum limit of 0.05 vanadium for weld materials in the beltline region.

t For staff position regarding compliance with the recommendations of RG 1.50, see Section 5.2.3, Question 5.

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SSAR Paragraph 5.3.1.6.1, " Compliance with Reactor Vessel Materials Surveillance Program Requirements," states that three capsules are prosided to meet the 10 CFR Part 50 Appendix H requirements. The staff finds this commitment not acceptable since the SBWR is designed for a 60-year life. The applicant must commit to proside at least

' four capsules and require a minimum capsule lead factor of 1.

SSAR Paragraph 5.3.1.8, " Regulatory Guide 1.65," states that the RPV studs, nut, and washer materials will be ultrasonically examined after final heat ta ratment and prior to-treading. The applicant must also commit to surface examine those items using magnetic particle or liquid penetrant examination after final heat treatment and prior to treading.

SSAR Paragraph 5.3.3.2.1, " Summary Description," sta'e, that the interior of the RPV is clad with stainless steel weld overlay and the bottom icad is clad with Ni-Cr-Fe alloy.

l The applicant must specify the cladding process esec. and identify the weld materials by specification and type.

1 SSAR Paragraph 5.3.3.2.2, " Reactor Vessel Design Data," states that CRD forged stub tubes fbr the CRD housing are made of ASME Sib 564 materials. The applicant must specify which grade of materials will be used. The applicant should also include the material specifications for the RPV drain nozzles and partial penetration f

instrumentation water level nozzles.

SSAR Paragraph 5 3.4, " COL License Information," should be revised to reflect that the COI applicant is to proside to the NRC staff for review actual PT limits cunes for the-specific RPV.

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- GE Response:

a)' The' SSAR will be changed to specify a maximum limit of 0.05% Vanadium for we't materials in the beltline region.

For response to the staff position concerning compliance with RG 1.50, see RAI #

EMCll.2, Question 5. SSAR Para. 5.3.1 A " Regulatory Guide 1.50" will be revised to.

incorporate the NRC comment.

b) The SSAR Para 5.3.1.6.1 will be changed to provide four capsules located in the f

beltline region, with a minimum lead factor of 1.

j c) The SSAR Para 5.3.1.8 will be changed to require surface examination of studs, nuts f

and washer materials using magnetic particle or liquid penetrant examination after

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final heat treatment and prior to threading.

i d) SSAR Para 5.3.3.2.1 will he revised to include the following: A variety of welding j

processes, such as electroslag, submerged arc, manual welding etc. are used for i

cladding depending upon the location and configuration of the item in the vessel..

Cladding in the "as-clad" condition may be acceptable for some deposits made with-automatic processes such as submerged arc welding, gas metal arc welding, and i

electroslag welding. For other processes, particularly where manual welding is cmployed, some grinding or machining is required. Workmanship samples are prepared for each welding process in the "as-clad" condition and for typically ground 1

surfaces.

The welding material used for cladding in the shell area is ASME SFA 5.9 or SFA 5.4, type 309L for the first layer and type 308L or 316L for subsequent layers. For the bottom f

head cladding, the welding material is ASME SFA 5.14, type ERNiCr3.

c) SSAR Para 5.3.3.2.2 will be revised to specify that the stub tube material is ASME Sib 564, grade N06600. The material specifications for the drain nonles and water level instrumentation nonles are specified in table 5.2-4.

O The SSAR Para. 5.3.4 will be changed to state the COL applicant will prmide actual.

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P/T curves for the RPV.

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RAI Number: EMCII.7 '

Question:

. SSAR xection 6.1.2 Organic Materials The SSAR imposes radiation exposure limit fbr organic materials of lE10 rads. GE should state that this limit applies for the whole life of the plant.

COL License Information should require that protective coatings in post-accident environments consider the generation of hydrogen from Zn containing primers and I'

topcoats.

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t The protective coatings should meet the requirements of ANSI 101.2, " Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities."

GE Response:

i The comment regarding the radiation exposure limit for organic materials is acceptable; the statement should apply for the entire life of the plant.

Additional information regarding Zn-containing primers and topcoats is not-considered necessary. GE does not pian to use any Zn coatings, as none have been qualified for nuclear service to the relevant ANSI standards. Current Zn-rich coatings -

will not meet EPA regulations as well.

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All epoxy coadngs will meet the requirements of ANSI N101.2, N101.4, and N5.12, as l

well as Regulatory Guide 1.54.

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.. j RAI Number: EMCil.10

- Question:

s SSAR Section 9.3.9 Hydrogen Water Chemistry i

- The design of the hydrogen water chemistry system in SBWR should meet.the -

requirements of EPRI Report NP-4500-SR-LD, " Guidelines for Permanent BWR Hydrogen Water Chemistry Installations."

t GE Response:

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. SSAR Section 9.3.9.1 indicates that EPR] report NP-4947-SR "BWR Hydrogen Water Chemistry Guidelines" (Reference 9.3-1) is utilized.

RAI EMCII.10 suggests utilization of EPRI report NP-4500-SR-LD " Guidelines for

. Permanent ilWR Hydrogen Water Chemistry Installations". This was a limited l

' distribution document which was ultimately issued as NP-5283-SR-A (same titic). We will revise Section 9.3.9.1 to reference both.NP-4947-SR and NP-5283-SR-A for use as~

appropriate.

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RAI Number: H i._. >

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i Several sections in the SSAR do not contain enough information to perform a complete human factors review. These SSAR sections are 1.8,13.2.1-2,13.5.2 and SSAR Chapter 18,:

These deficiencies are discussed in RAI numbers HHFB.1-4.16.

7 GE Response:

The February 28,1993, submittal includes Appendix 18F which contains the mimmum inventory of controls, displays and alarms required by the operators to implement the j

SilWR Emergency Procedure Guidelines and important operator actions as defined in the probabilistic risk assessment. With the material in Appendix 18F, Chapter 18 is

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complete.

Section 1.8 is a summary description of material that is contained in the rest of the l

SSAR and is a part of the February 28,1993 SSAR submittal.

l Subsectionsl3.2.1-2 and 13.5.2 contain no discussion because they each address topics -

which are outside the scope of the SilWR Standard Plant Design.

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RAI Number: HHFB.1

-i ua Question:

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.SSAR Section 1.8, " Interfaces for Standard Design"

- l l-This section was not included in the application. The applicant states that this'section l

i will be submitted on February 28,1992.

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I GE Response:

l SSAR Section 1.8 for SBWR is titled " Summary of COL License Information," conforms with Standard Review Plan (SRP) Section 1.8, and was submitted on February 28,1993.

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-i RAI Number: HHFil.4.1 -

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Question:

ITAAC were not included in the application. The applicant states that SSAR Chapter 18 q

will be sulunitted on February 28,1992.

.. i GE Response:

The SilWR ITAAC were submitted to 8.he.NRC staff for review by February 28,1993. The-ITAAC fbr Chapter 18 were included in that submittal.

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P RAI Ntunber: HH FIl.4.3 I

Question:

1 The applicant has not described how the GSIs and USIs invohing human-system concerns are resolved in the SilWR design.

GE Response:

.The resolutions of USIs and GSIs invohing human-system concerns (i.e., NUREG-0933 i

items HFl.1, HF4.4, HF5.1, and HF5.2) are described in Appendix 19H, ". Resolution of USIs/GSIs," Subsections 19H.2.57 and 19H.2.58. Appendix 19H was submitted on-February 28,1993.

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. RAI Numbcr: ' H H FB.4.4 i

Question:

.i The applicant has not described MMIS design goals in " operator centered" terms.

l GE Response:

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The primary man-machine interface design goal is to " facilitate safe, efficient and i

reliable operator performance....". This is considered by General Electric to be

~j operator centered terminology because the operator is the central figure upon which the design goal is focused.

Also,it is specified that, to achieve this goal, the devices which make up the MMIS "shall he implemented in a manner consistent with good human factors engineering practices". This, in effect, makes the entire section operator centered because good human factors engineering practices take into consideration the operator's strengths, weaknesses, skills, size and other human characteristics and assure that these characteristics will he factored into the MMIS design.

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RAI Number: HICIL8

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The unresolved safety issues / generic safety issues l(USI/GSI) applicability has not been ij addressed.- The SSAR states that this will be provided in the February 28,1993,-

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submittal.

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3 GE Response:

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i Applicability of USIs/GSIs is addressed and documented in Appendix 19H, "Resolutwn -

of USIs/GSis," and was submitted on February 28,1993.

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RAI Number: RPEh.6 Question:

SSAR Section 14.2.8 Individual Test Description 1

It is the staff's position that test abstracts in this section of the SSAR should address the following:

The level of detail in the test abstracts is insufficient to determine conformance with RG 1.68, Position C.')

Several test abstracts include imprecise acceptance criteria (i.e., acceptable, allowable, design, estimated, expected, selected, specified, appropriate, and within limits). Individual test abstracts should clearly specify the bases for determining acceptable system and component performance. Acceptable criteria include specific references to regulatory guides, technical specifications, assumptions used in the safety analysis, other GESBWR-DC sections, and applicable codes and standards.

GE Response:

The AllWR SSAR equivalent section is currently being modified as described in this i

RAl. Thus, the SBWR SSAR Section 14.2 will be modified to be more descriptive in referencing the appropriate standards and design documents. Section 14.2 will be modified to conform to these standards in future SBWR SSAR revision.

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4 RAI Number: RPEll.13 Question:

SSAR Section 17.3.9 describes O-RAP acthities and the process used to identify which actisities are needed for a particular SSC. Periodic testing is one of the 0-RAP actisities -

i described in this section. However, the description of periodic testing is not clear as it relates to inspection of SSCs such as tanks and pipes. (For example, is inspection the only periodic testing that will be perfbrmed on SSCs such as tanks and pipes, or will inspections be limited to SSCs such as tanks and pipes?) GE should clarify what periodic testing will be applied to SSCs such as tanks and pipes in SSAR Section 17.3.9.

GE Response:

The SSCs mentioned in Section 17.3.9 are examples, not a total maintenance program.

defined for any SSC, whether pumps, instrument circuits, tanks, pipes, or heat exchangers. Section 17.3.9 says. "some SSCs may require a combination of activities to assure that their perfbrmance is consistence with that assumed in the PRA". This is not the section in which to specify total testing or maintenance actisities related to tanks and pipes.

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-i RAI Number: SPLB.11.

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-f Identify all watertight doors' and hatches on the general arrangement drawings.

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All the watertight doors and hatches are currently identified on the general arrangement drawings. The general arrangement drawings show the water tight doors -

with the symbol as shown in the symbol legend. There are no watertight hatches.

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RAI Number: SPLIL26' j

s Question:

o SSAR Section 9.2.1.2 states that the plant service water system (PSWS) rejects heat from non-safety-related components in the acactor and turbine buildings to the

-l environment. Heat is rejected to the ensironment by mechanical-draft cooling towers (site specific). However, the water source for the PSWS was not addressed. Section 9.2.9 states that the COL applicant shall provide the design of the senice water basin 'or -

j other site-specific water supply. Provide provisions or reference criteria for the design j

of PSWS water intake devices.

P GE Response:

i Please note that the PSWS is a closed loop system with heat rejection by the mechanical draft cooling towers. There are no significant water requirements. - Initial system fill, j

evaporative and windage losses as well as blowdown will be supplied, per Section'9.2.1.2, l

by the Station Water System which is site specific and will be COL applicant furnished '

l per Section 9.2.9.

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Section 9.2.9, Plant Service Water Basin, states, in part, that the COL applicant shall proside the design of the senice water basin or other site specific water supply. If,on a-

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site specific basis, the Plant Senice Water Basin was not a closed loop per the reference plant, then the COL applicant would provide the design of that site specific water supply intake. It is expected that any intake structure design would satisfy appropriate j

Civil Structural Society and Hydraulics Institute Standards.

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RAI Number: SRXil.3 j

Question:

SSAR Section 1.2.2.7, Reactor Protection System. Unlike in current boiling water reactors (BWRs), the reactor is also scrammed on reactor pressure vessel (RPV) level 8.

Is there any safety significance for this trip or is this only for the main turbine-protection?

GE Response:

There is no safety significance for the reactor scram function on RPV Level 8 and this reactor trip function is not provided for the main turbine protection.

j Main turbine protection upon RPV high water level for the SBWR is an assigned R

function of the Feedwater Control System (FWCS). The FWCS determines whether the Level 8 setpoint has been exceeded and then sends signals to the turbine control.

system which utilizes these signals to initiate a turbine trip upon the indication of high water level in the RPV. The Feedwater Control System itselfis provided with narrow range level signals from Nuclear Boiler System (NBS) sensors associated primarily with l

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functions of the Feedwater Control System. These non-safety-related sensors are separate from the safety-related NBS narrow range and wide range level sensors associated with the reactor safety and protection systems' functions. The setpoints for turbine protection functions may or may not correspond _to the same actual water level trip setpoints used by the RPS. Other trip setpoints of the Feedwater Control System i

(e.g., reactor pressure vessel high water level feedwater runback and feedwater pump

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trip setpoints) also may or may not exactly correlate with corresponding trip setpoints used by safety systems.

It should also be pointed out that the previous generat%n of GE BWRs,i.e., the GE BWR/6 product line projects such as Grand Gulf, Riveraend, Clinton, etc., did have the RPV I.crel 8, as well as Level 3, reactor scram setpoints.

For the SBWR project, the RPS RPV water level 8 trip, resulting in reactor scram with subsequent substantial water level decrease, mitigates any reactor pressure vessel water inventory increase event and provides additional operational time to delay and potentially prevent (through appropriate Operator actions) the feedwater pump trip l

which should occur at Level 9 (which is higher than Level 8). This Level 8 scram function also, by anticipating the non-safety-related turbine trip which also occurs at 1

Level 8, provides additional mitigation of the increasing reactor pressure and neutron flux transients which would result as a consequence of such turbine trips at high reactor-power levels.

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1 RAI Number: SRXB.f>

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SSAR Section 1.2.2.4.3. Only RPV level 1 and not dqwell high pressure is given as the initiation signal for the GDCS. In current IlWRs, RPV low level and drywell high pressure signals are used, providing initiation diversity. Why is only the level signal j

used for starting the GDCS? Explain in detail why diversity is not necessag for the start-up of emergency core cooling systems (ECCS) in the SBWR.

i GE Response:

The initiation of GDCS on RPV level 1 is part of SBWR's emergency core cooling system response and is initiated via ADS logic. High drywell pressure is not used as a diverse initiation signal because the consequences ofinadvertent ECCS initiation in SilWR are significant due to the design of the DPVs. D well pressure transients are probable 9

during the 60 year operational life of the plant as a result of drywell cooling unit failures. If dqwell pressure was used as a diverse initiation signal for ECCS,' and filure of drywell coolers occurs and drywell pressure exceeds the initiation setpoint a relatively minor transient now results in unnecessary depressurization as ADS initiates to depressurize the RPV using SRVs and permanently open DPVs.

ECCS initiation signal diversity is assured in the ADS design by using redundant microprocessor based and harchvired logic channels and diverse level transmitters.

1 Refer to the Nuclear Boiler System (NBS) logic diagram, Figure 21.7.3-1, for tpecific design details. NILS logic is discussed in Subsection 7.7.1.

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1 RAI Number: SRXB.24 Question:

l We understand that the control rod has no velocity limiter. Discuss in detail the reason -

for ulocity limiter climination.

GE Response:

The locking piston control rod drive mechanism used in current BWRs cannot detect separation of the control rod from the drive mechanism during normal rod movements. In order to prevent damage to the reactor coolant pressure boundary by the rapid reactivity increase which would result from a free fall of a control rod (rod 1

drop accident) from its fully inserted position to the position where the drive i

mechanism is withdrawn, a velocity limiter is provided on the control rod to. restrict the j

control rod free-fidl velocity to acceptable limits.

In contrast to the locking piston control rod drive, the fine motion control rod drive 9

(FMCRD) is designed to detect separation of the control rod from the drive.

mechanism. Two redundant and separate Class 1E switches are provided to detect the separation of either the control rod from the hollow piston or the hollow piston from the from the ball-nut (refer to Subsection 4.6.1.2.2 for a discussion'of the principle of; j

operation of the separation detection switches). Actuation of either of these switches' will cause an immediate rod block and imtiate an alarm in the control room, thereby-j preventing a rod drop accident from occurring.

As discussed in Subsection 15.4.9, for a rod drop accident to occur with the FMCRD

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i design, it is necessary for multiple, highly unlikely independent failures to occur simultaneously with the occurrence of a struck rod on the same FMCRD. This is displayed in the attached Figure SRX11.24-1, which shows the various failure paths leading to a control rod drop on the FMCRD. As a consenuence of the low probability of the simultaneous occurrence of these multiple independent events in each failure

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path, there is no basis to postulate a rod drop accident for the FMCRD design.

Ily designing the FMCRD to prevent a rod drop accident, the need for the velocity

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limiter is climinated. It is on this basis that the SBWR control rod design does not -

include a velocity limiter.

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RAI Number: SRX11.28 Question:

The isolation condenser system (ICS) is a safety-related system and the SBWR ICS design is similar to the ICS in operating plants like Dresden 2 and 3, Millstone 1, Nine Mile Point I and Oyster Creek. Ilut the operational experience with the ICS in those -

.i plants has been of concern to the staff. The stairs experience with operational events relating to the ICS has indicated numerous design deficiencies and several operational problems. Has GE perfbrmed a systematic study of the operational experience related to ICS plants? What design changes and improvements have been made to the SllWR ICS design to correct potential design deficiencies in operating ICS plants?

GE Response:

4 The fbilowing design improvements have been included in the SilWR ICS design:

1. Use of carbon steel supply piping, Inconel tubes with butt-welded end attachments, and low carbon or nuclear grade stainless steel condensate return piping which is resistant to IGSCC.

2. The condensate remrn lines are continuously sloped downward from the IC to an elevation below reactor water level to avoid the trapping and collapse of steam in the -

L drain piping.

3. The quality of the makeup to the IC pools is such that pool boiloff to atmosphere and the surroundings should not require cleanup.

4. Three IC loops are provided, either of which will allow reactor operation'at 80% of full power, and two or more IC loops will allow reactor operation at 100% or higher

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RAI Number: SRXB.45 l

Question:

For event 15.4.9, control rod drop accident, what would be the consequences if the

- separation-detection alarm failed for a stuck control rod, and it could drop to its j

maximum distance? Will the distance of the rod drop be limited so as to preclude unacceptable consequences?

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GE Response:

The system which detects control blade separation is safety grade with redundant:

components. Therefore it is incredible that the blade could separate and drop.

Independent from the control blade separation-detection alarm, the Rod Control and information system (RC&lS) has a rod worth minimizer (RWM) design that rest'ricts the -

maximum distance a rod could drop. The ganged withdrawal sequence of the RWM restricts this distance and thus minimizes the rod worth such that any unacceptable i

consequences are precluded. Best estimate calculations indicate that even if a rod drop-occurred, the 280 cal /gm fuel enthalpy limit would not be exceeded.

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