Text

Forwards Responses to RAIs Transmitted by NRC Re Sbwr Design.Draft Rept Entitled, Sbwr Containment Natural Aerosol Removal Also Encl
	ML20064F625
Person / Time
Site:	05200004
Issue date:	02/24/1994
From:	Leatherman J; GENERAL ELECTRIC CO.
To:	Borchardt R; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20064F626	List: ML20064F625; ML20064F644;
References
	MFN-021-94, MFN-21-94, NUDOCS 9403150373
	Download: ML20064F625 (91)
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'- 1 GE Nuclear Energy

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February 24,1994 MFN No. 02194 Docket STN 52-004 t

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l Document Control Desk U.S. Nuclear Regulatory Commission Washington DC 20555 Attention: Richard W. Borchardt, Director I Standardization Project Directorate .j

Subject:

NRC Requests for AdditionalInformation (RAls) on the Simplified Boiling Water Reactor (SBWR) Design

References:

Transmittal of Requests for AdditionalInformation (RAls) Regarding the SBWR ;

Design, Letter from M. Malloy to P. W. Marriott dated November 29,1994 1 The Reference letter requested additional information regarding the radiological impacts contained in Chapters 2,6 and 15 of the SBWR Standard Safety Analysis Report (SSAR). In .

fulfillment of this request, GE is submitting Attachment 1 to this letter which contains responses to -

l these RAls.

The responses to RAls 470.4,.8,.11,.26,.33 and .34 indicate that analyses will be performed in ;

support of these RAls. Based on discussions with the NRC reviewers on their need dates for ;

receipt of this information, GE proposes to transmit these analyses by that expressed need date of May 1,1994.

Sincerely, -

]. - .

,se: g&imo }

J. E. Leath'erman SBWR Certification Manager MC-781, (408)925-2023 00180 Attachment 1, " Responses to NRC RAls" cc: M. Malloy, Project Manager (NRC) (w/2 copies of Attachment 1)

F. W. Hasselberg, Project Manager (NRC) (w/1 copy of Attachment 1)

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Meteorology (Chapter 2)

RAI Number: -470.1 Questiori: ,

Section 2.0, " Site Characteristics," of the SilWR standard safety analysis report (SSAR) defines the envelope of site-related parameters which the S11WR standard plant is designed to acconunodate. Provide a table showing the envelope of SilWR standard plant site design parameters, including but not limited to the (1) tornado design basis and (2) bounding atmospheric relative !

concentrations (C/Q) for the exclusion area boundary and for the low population zone. The bounding C/Qvalues should provide assurance that (1) the radiological ef1luent release limits associated with normal reactor operation (specified in 10 CFR Part 50, Appendix 1) will be met, and (2) the radiological consequence, of a range of postulated accidents, up to and including the limiting design-basis accident (DilA) considered, will be acceptable for an individual located at the nearest boumla,y of the exclusion area for a specified time, GE Response:

SilWR Standard Plant Site Dmien Parameters .,

Parameter SSAR Reference 'l

Regional Climatology Section 2.3.1 ;

(Including Tornado Design Ilasis) i

Normal X/Q Section 12.2

DllA x/Q Chapter 15 -

- Exclusion Area lloundary Table 15.414 .,

- Low Population Zone Table 15.415 i

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RAI No nber 470.2 l Question:

Local meteorology in SSAR Section 2.3.2 and the.on-site meteorological measurements program in Section 2.3.3 should be designated as combined operating license (COL) applicant action items.

GE Response:

1 Please see attached revised SSAR sections 2.3.2 and 2.3.3.

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25A5113 Rev. A SBWR Standant SafetyAnalysis Report i

1 The manmum design ambient tempemture corresponding to a one percent

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exceedance value is 37.8 C (100 F) dry bulb with a coincident wet bulb tempenture of 25 C (77 VJ and 26.7 C (80 F) for non<oincident wet bulb.The minimum design temperature corresponding to a one percent exceedance value is-23.3 C (-10 F).

The zero percent exceedance dry bulb temperature is 46.1 C (115 F) with a ceincident wet bulb temperature of 26.7 C (80 F) and 27.2 C (81*F) for non-coincident wet bulb.

The minimum temperature for this exceedance value is -40 C (-40 F). ;

The ruaximum rainfall rate for roof design is 0.49m/h (19.4 in./h), which is based on l the probable maximum precipitation (PMP) for one hour over one square mile with ratio of 5 minutes to one hour PMP of 0.32, as found in National Weather Source Publication HMR No. 52. The maximum snow load for roof design is 2394 Pa (50 lb/

sq ft).

The 10,000,000-year tornado has a maximum wind speed of 134m/s (300 mi/h), a translational velocity of 26.8m/s (60 mph), and a radius of 45.7m (150 ft).The maximum atmospheric pressure differential is 13.8 kPa (2 psi) and the rate of pressure thange is 4.85 kPa/s (1.2 psi /s). The missile specta is per Spectra 1 ofStandard Review Plan 3.5.1.4. Missile velocity is 35% of the maximum horizontal wind speed with an altitude of 9.1m (30 ft) above grade for large soft and rigid missiles. Small rigid missiles are postulated at all elevations.

2.3.2 Local Meteorology COL aoplicants will provide local me teoroloev for NRC review.

2.3.3 On-site Meteorological Measurements Program COL at'plicants will pigvide the on-site meteorolocical measurements procram.

2.3.4 Short-Term Diffusion Estimates Short term diffusion estimates are given in Chapter 15. The COL applicant will proside short term diffusion estimates in accordance with Regulatory Guide 1.145 for comparison to dose values given in Chapter 15. They must be shown to result in doses less than stipulated in 10CFR100 and in the applicable portions of SRP Sections 11 and

15. (See Subsection 2.7.2 for COL license information requirements.)

2.3.5 Long-Term Diffusion Estimates Long term diffusion estimates are given in Chapter 12. The COL applicant will provide long term diffusion estimates in accordance with Regulatory Guide 1.112 for comparison to Chapter 12 values. (See Subsection Subsection 2.7.2 for COL license information requiremen:s.)

232 Meteorology- Amendment 1 DRAFT 3294

Control Room Habitability (Chapter 6) '

RAI Number 470.3 Question:

The SI1WR main control room is located on the ground level and within the reactor building, adjacent to the senice and turbine buildings. During and following a loss-of coolant accident (l.OCA), which is the controlling DIlA for -

the radiological consequence to the control room operators, the radiation 4 exposures to the operators will consist of contributions from airborne fission- -

products entered into the control room and direct gamma radiation from the surrounding buildings and process equipment. For determination of the gamma radiation dose to the control room operators, state the major gamma radiation sources, including the main steam lines, and the shielding provided (Doors and control room wall thicknesses).

t GE Response:

The gamma dose contributions to operators in the main control room are given in subsection 15.6.5 and consists of the following contributions: (1) airborne noble gases and iodines - 0.13cSv (= rem), and (2) shine from the corridors, safety envelope, and containment - I cSv(= rem). See attached revised Figure 15.44b which corrects a labeling error to read cSV. The main steam lines contribute negligible doses. The shielding provided around the main control room is shown on figure 21.12.3-1. Sheet 8.

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15.6 60 Decrease in Reactor Coolant Inventory- Amendment 1 DRAFT l 2/2f@4

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- RAI Number 470.4 Question:

6SAR Section 6.4.2 states that the emergency breathing air system (EllAS) is--

automatically initiated upon automatic isolation of the sealed emergency ,

operation area (SEOA) by a high radiation signal from the normal control room '

ventilation system. Show this feature in SSAR Figures 21.6.4-1 and 21.9.4-1.

GE Response:

1 See revised SSAR sectin 6.4.2, attached, which states that the embrgency i breathing air system (EllAS) is manually initiated.

The LOCA DllA analysis in section 15.6.5 which assumes automatic initiation of the EllAS, will be reviewed and revised accordingly (see question 470.19).

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25A5113 Rev. A SBWR standard Safety Analysis Report 6.4.2 System Design Figure 21.6.4-1 provides the EBAS P&ID. The CREIIVAC P&lr' aa.1 system description are provided in Subsection 9.4.1.

fhe EBAS is a redundant safety-related system which supplies stored, compressed air to the SEOA for breathing and for pressurization to minimize inleakage. The EBAS is -

manually initiated by control room nersonnel. automatically 4aitiated-upon-automatic-isolati<nvohhe-SEGA4ause4hu-high-radiationet-the-normal-controkoonvventilatiom i syst em-(CRE11VAC)-air-intakerThe-GRE L IVAG-system 41uetsweels<smt oma tically-isolated 4>y4hinigna!. h EllAS-alsmmdem4>e-manually-initiated 4>y-entrohoom-personnel.The system consists of two redundant, completely independent trains of bottled air. Each redundant train of air bottles has been sized to provide sufficient breathing quality bottled air to maintain a positive pressure in the SEOA for a minimum of72his.

6.4.3 Sealed Emergency Operating Area (SEOA)

The SEOA includes the following areas:

a Main control room; a Shift supervisor room; a Shift supervisor clerVs room; a Shift supervisor conference room; a Operator's area; a Shift technical advisor's room; a Switch and tag room.

This space constitutes the operation control area which can be isolated for an extended period if such is required by the existence of a LOCA or high radiation condition. 1 Radiation Protection i Description of control room instrumen tation for monitoring of radioactivity is given in Sections 11.5 and 12.3. i Shielding Design The control room shielding design is based upon protecting personnel from radiation resulting from a design basis LOCA. The radioactive sources existing under normal operating conditions are not determining factors for the shielding design.

6.44 ContmlRoom Habitability Systems - Amendment 1 DRAFT

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. . - , + . . . _ + __u _.. 4 - 4. .,a- - 2 -- # . _A RAI Number: 470.5 Question:

In the analysis of the control room operator radiation doses following a LOCA, full credit is taken for fission-product removal by the control room envelope heating, ventilation, and air conditioning (CREHVAC) system after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months into a LOCA. This is unacceptable since the CREHVAC system is neither chtssified nor qualified as an engineered safety feature (ESF) system and needs '

to be addressed. (Th( 'R CREHVAC system is a single train non-safety-grade system.) ;

GE Response:

r The analysis assumes the use of the CREHVAC system as the more conservative dose consequence model. The reasoning behind this analytical decision is that the EllAS system which consists of a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months bottled air system can be easily replenished by trucking or flying in replenished bottles so that an uncontaminated source of air can he supplied to the EllAS indefinitely. Under.

these circumstances, iallow to the control room would consist only of that unfiltered inflow created by personnel entering and exiting the control room. ;

By using the CREHVAC in the analysis, an additional source of contamination is introduced into the control room through the fiher system since the filters are assumed to operate at 95E This then provides the most bounding analysis in the operation of the control room systems.

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i RAI Number: 170.6 1 Question 1 SSAR Table 15fr13 shows the control room atmospheric relative concentrations j (C/Q) used and the resulting control room operator dcscs. I.ist the _ major parameters, assumptions, and methodologies used in determining the C/Q j values and control room operator doses. l GE Respone l The control room X/Q is calculated using the parameters found in Table 15fr9 111 along with the process that is explained in section 15.6.5 Control Room'(page !

15.6-16). -)

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RAI Number: 470.7 Question:

The staff's evaluation of Electric Power Research Institute's (EPRI's) Light-Water Reactor Utility Requirements Document for passive plant designs (EPRI's Passive Requircraents Document) accepts the use of a passive, safety-grade control room pressurization system that would use bottled air to keep the operators' doses within the limit of General Design Criterion (GDC) 19 and Standard Review Plan (SRP) Section 6.4 (Revision 2) for the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the DIIA. The evaluation addresses use of safety-grade connections for the pressurization system to allow the use of ojrsite portable air supplies after 72

. hours to minimize the operators' doses for the entire duration of a DIIA (30 days). In either case (use of bottled air for the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months only or the use of offsite portable air after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the entire duration of a DilA), GE should specify in the Si1WR Tier 1 Design Certification Document; i.e.,'in the inspections, tests, analyses, and acceptance criteria (ITAAC); that the feasibility and capability of the safety-grade bottled air supply system to maintain positive pressure in the control room envelope should be demonstrated.

GE Response ,

When the Tier I DCD is amended it will incorporate a test to verify that the EllAS can maintain a positive pressure in the control room and an inspection of the as-built configuration to verify that a temporary air supply can be established. A draft copy of this material is attached. The requirement to ,

insure an adequate supply of bottles is not a design commitment but an operational licensee commitment and should be addressed by the operational licensee.

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b A N /~1 h' y 25A5354 Rev. A ' ' 4V EE A SBWR cenised Design uaterial 2.15.6 Reactor Building HVAC Design Description ControlRoom Area Ventilation System The Control Room Envelope lleating, Ventilating, and Air Conditioning System (CREIIVAC) provides a controlled emironment for personnel comfort and safety and for the operation of equipment in the main control room envelope which includes the Niain Control Room (NICR), the Technical Support Center (TSC), the computer room, and adjacent rooms supporting the control room. Figure 2.15.ti1 shows the basic configuration and scope. Except for the CREIIVAC isolation dampers that penetrate the Scaled Emergency Operating Area (SEOA) boundary, and the associated air monitors and controllen that provide the isolation signals, the CRElIVAC System is classified as non-safetv-related/The Emergency Breathing Air System (EBAS) is a /C ,

D redundant safety-related system which supplies stored, compressed air to the SEOA for 3

breathing and pressurizing the SEOA to 34.5 Pa for leakage rates of up to 17 m /h to 'h minimize in leakage. The EBAS is manually initiated upon isolation of the SEOA. Each , , ,

redundant train of bottled air can prmide sufficient breathing quality air to maintain a ',s

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positive pressure in the SEOA for a minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and includes provisions for

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connections of temporary air supplies at locations which are accessible under DBA conditions.

Normal Operating Mode in the normal mode, one supply fan and one exhaust fan operates. The system operates by recirculating a portion of the exhaust air (return air) and mixes it with outside air.

The CREllVAC System maintains a slight positive pressure in the Control Room Envelope during nonnal operation The SEOA isolation dampers are closed upon receipt of an isolation signal from the Irak Detection & Isolation System (LD&lS).

High Radiation Mode On receipt of a Process Radiation Monitoring System (PRNfS) signal for high radiation in the outside airintake plenum, automatic controls shut the normal outside air damper upstream of the air conditioning unit, opens the outside air and return air dampers upstream of the supplementary (High Efficiency) filtration unit, and start an associated recirculation fan.

Smoke RemovalMode The smoke removal mode may be manually initiated so that outside air is supplied through the air conditioning units and exhausted to the atmosphere without recirculation. If high airborne concentrations of chlorine, smoke, high radiation or

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1 ot her site specific toxic gases are detected in the intake plenum, automatic controls simt down the utility fans, shut the CRE isolation dampers, and fully open the return air ;

damper (the CRElIVAC continues to run in the 100% recirculation mode). If high concentrations of chlorine or other toxic gases are detected, the controls initiate the same automatic operation and also start the supplementary filtration system.

The CRElIVAC System penetrations of the SEGA and the isolation dampers are classified as Seismic Category I. The Control Room Envelope is located in the Reactor Building.

Each SEOA isolation damper requiring electrical power is powered from the Class 1E dhision as shown on Figure 2.15.6-1. In the Cl?EllVAC, independence is prcaided between Class IE divisions, and also herween Class lE divisions and non-Class 1E equipment.

Fire dampers with fusible links in IIVAC duct work close under air flow conditions.

The CREIIVAC System has the following displays and controls in the MCR:

a Control and status indication for the active components shown on Figure 2.15.6-1.

m Parameter displays for the instruments shown on Figure 2.15.6-1 except for the smoke detectors.

The pneumatically-operated isolation dampers, shown on Figure 2.15.6-1, fail to the closed posidon in the event ofloss of pneumatic pressure or loss of electrical power to the valve actuating solenoids.

Interface Requirements Toxic gas monitors will be located in the outside air intakes of the CREllVAC System, if the site is adjacent to toxic gas sources with the potential for releases of significance to plant operating personnelin the SEGA. These monitors should have the following requirements:

a Be located in the outside air intakes of the Control Room Envelope a Be capable of detecting toxic gas concentrations at which personnel protectii actions must be initiated. l I

Spent Fuel Pool Area Ventilation Systern The Reactor Building Ventilation System for the spent fuel pool area is called the Refueling and Pool Area Ventilation System (REPAVS).

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25AS354 Rev A j SBWR certised Design Material r ., q , r gqpp o e 1" % y The REPAVS prosides a controlled emironment for personnel comfort and safety, and for proper operation and integrity of equipment in the refueling area. Figure 2.15.42 shows the basic system configuration and scope.

Except for the portion of the system that isolates the refueling area boun6ay, the REPAVS is classified as non-safety-related.

Normal Operating Mode in the normal operating mode, one air conditioning unit operates with r. single supply fan and exhaust fan.

The REPAVS maintains a negative pressure in the refueling area relative to the outside atmosphere.

The REPAVS isolation dampers are dosed and the ventilation supply and exhaust fans are shut down in the event of high radioactivity detected by the PRMS in the exhaust air from the areas served by REPAVS.

Instrumentation and controls for the REPAVS are located at the local panels. General trouble alarms for each local panel and hand switches with position indicators for the isolation dampers are located in the main control room.

The REPAVS penetrations of the refueling area bounday and isolation dampers are classified as Seismic Categon I. The REPAVS is located in the Reactor Iluihling.

Each REPAVS isolation damper requiring electrical power is powered from the Chiss 1E division as shown on Figure 2.15.42. In the REPAVS, independence is provided between Class IE divisions, and also between Class 1E divisions and non-Class 1E equipment.

The pneumatically-operated isolation dampers, shown on Figure 2.15.42, fail to the closed position in the event of109 of pneumatic pressure or loss of electrical power to the valve solenoids.

Reactor Building HVAC System The Reactor Iluihiing HVAC System consists of the following stiosystems:

a Clean Area Ventilation System (CIAVS); and-a Controlled Area Ventilation System (CONAVS).

The reactor building ventilation systems provide a controlled environment for personnel comfort and safety, and for proper operation and integrity of equipment.

Except for the CONAVS safety envelope isolation dampers, the Reactor lluilding iIVAC Systems are classified as non-safety-related. Figure 2.15.43 shows the CONAVS basic configuration.

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25A5354 Rev. A SBWR certmedcesion Materiai B h.

NormalOperating Mode The CIAVS is a recirculating ventilation system that prmides filtered, heated or cooled, and humidified air to the clean areas of the reactor building.

The CONAVS is a once-through ventilation system that provides filtered and heated or ,

cooled air to the potentially contaminated areas of the reactor building including the l

safety envelope.

For the CONAVS, interlocks allow the supply fans to run only when negative pressure is established in the associated spaces.

High Radiation Mode The CONAVS supply and exhaust for areas within the safety envelope are automatically shutdown and the containment isolation dampers closed if a contairunent isolation signal is received or high airborne radioactivity is detected by the PRMS in the exhaust duct. If high airborne radioactivity is detected in any areas scived by the CONAVS that are outside the safety envelope, the ventilation supply and exhaust is shut down for that area and the corresponding isolation dampers closed. Safety envelope isolation dampers automatically shut and associated fans stop if off-site power is lost. ;

Smoke RemovalMode i The purge exhaust fims are also used for smoke removal from areas in the reactor building and the refueling floor that are affected by a fire.

Insinunentation and controls for the CONAVS are located at the local panels. General trouble alarms for each local panel and hand switches with position indicators for the isolation dampers are located in the main control room.

The CONAVS penetrations of the safety envelope and the isoladon dampers are classified as Seismic Category 1. The CIAVS and CONAVS are located in the Reactor l Iluilding.

Each CONAVS isoladon damper requiring electrical power is powered from the Class lE division as shown on Figure 2.15.43. In the CONAVS, independence is provided between Class lE divisions, and also between Class IE disisions and non-Chiss IE equipment.

1 The pneumaticdly-operated isolation dampers, shown on Figure 2.15.G3, fail to the j closed position in the event ofloss of pneumatic pressure or loss of electrical power to I the valve solenoids. .

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! Table 2.15.6-1 Control Room Envelope HVAC (CREHAVC) System M a CD

{ inspections, Tests, Analyses and Acceptance Criteria k Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. The basic configuration of the CREHVAC 1. Inspections of the as-built system will be 1. The as-built CREHVAC System conforms s System is as shown on Figure 2.15.6-1. conducted. with the basic configuration shown on ,,y y Figure 2.15.6-1. J

$ 2. The CREHVAC areas are maintained at a 2. Tests will be conducted on the as-built 2. The CREHVAC areas are maintained at a .j 8 minimum pressure of 3.2 mm water CREHVAC System in the normal mode of minimum pressure of 3.2 mm water f gauge above the outside atmosphere. operation. gauge above outside atmosphere. ,@

a 2 3. The CREHVAC SEOA isolation dampers 3. Tests will be conducted on the CREHVAC 3. Upon receipt of a simulated signal, SEOA -qq are closed upon receipt of an isolation System using simulated LD&ls isolation isolation dampers are automatically "

signal from the LC&lS. signals. closed. d

4. Each redundant train of EBAS provides a 4 Tests and analysis will be performed on 4. EBAS air storage capacity in each train is I

stored air supply capacity sufficient to the EBAS. sufficient to maintain a positive y maintain the SEOA at a positive differential pressure of greater than or s differential pressure of 34.5 Pa for 72 equal to 34.5 Pa for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . $

hou rs. $

5. The smoke removal mode is automatically initiated upon receipt of a

5. Tests wi!! be conducted of the smoke removal mode.

5. Upon receipt of a simulated signal, the following automatic actions occur:

[

smoke detector signal. a. The utility fans shut down.

b. CRE isolation dampers close.

c. Return air dampers open (the CREHVAC System operates in a 100%

recirculation mode).

w

, b.

g.

3 E:L 9 $

t a a

.n 1

l i

t Table 2.15.6-1 Control Room Envelope HVAC (CREHAVC) System (Continued) m m -

$ inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria

6. Each CREHVAC SEGA System isolation 6. 6.

damper requiring electrical power is a. Tests will be performed on the a. The test signal exists only in the Class powered from the Class 1E division, as CREHVAC System by providing a test 1E division under test in the CREHVAC shown on Figure 2.15.6-1. In the signal in only one Class 1E division at System.

CREHVAC System, independence is a time. b. In the CREHVAC System, physical provided between Class 1E divisions, and separation or electricalisolation exists between Class 1E divisions and non-Class b. Inspection of the as-butt Class 1E between Class 1E divisions. Physical 1E equipment- divisions in the CREHVAC System will . .

separation or electrical:. solation exists be performed. ,

between these Class 1E divisions and non-Class 1E equipment.

7. Fire dampers with fusible links in HVAC 7. Type tests of fire dampers in a test facility 7. Fire dampers close under system air flow duct work close under air flow conditions. will be performed for closure under conditions. y system air flow conditions. E C

8. Main control room displays and controls 8. Inspections will ba performed on the 8. Displays and controls exist or can be T provided for the CREHVAC System are as main control room displays and controls retrieved in the main control room as l' defined in Section 2.15.6. for the CREHVAC System. defined in Section 2.15.6. [

9. The pneumatically-operated CREHVAC 9. Tests will be conducted on the as-built 9. The CREHVAC SEOA isolation dampers isolation dampers, shown on Figure CREHVAC System pneumatic isolation shown on Figure 2.15.6-1 fail to the closed 2.15.6-1, fail to the closed position in the dampers. position on loss of pneumatic pressure or event of loss of pneumatic pressure or loss of electrical power to the valve %

2 actuating solenoids. '

[ loss of electrical power to the valve actuating solenoids.

1,=

7 .-

q

~ 2;

's j

x

[3 2-5.? $

G n c ; g ;

l _, 4

% v:r :r:-

c3 ,

i %s tr a

as w .,1 i

E Table 2.15.6-2 Refueling and Fuel Pool Area Ventilation System (REPAVS) M n II3

[ Inspections, Tests, Analyses and Acceptance Criteria ks- Design Commitment inspections, Tests, Analyses Acceptance Criteria

$ 1. The basic configuration of the REPAVS is 1. Inspections of the as-built system will be 1. The as-built REPAVS conforms with the g as shown on Figure 2.15.6-2. conducted. basic configuration shown on y Figure 2.15.6-2.

$ 2. The REPAVS maintains a negative 2. Tests will be conducted on the as-built 2. The REPAVS maintains a negative 3 pressure in the refueling area relative to REPAVS in the normal mode of operation. pressure in the refueling area relative to I the outside atmosphere. the outside atmosphere.

s Upon receipt of a simulated signal,

3. The REPAVS isolation dampers are closed 3. Tests will be conducted on the REPAVS 3.

upon receipt of an isolation signal from using simulated LD&lS isolation signals. REPAVS isolation dampers are the LD&lS. automatically closed.

4. Each REPAVS isolation damper requirLig 4. 4.

electrical power is powered from the a. Tests will be performed on the a. The test signal exists only in the Class %

Class 1E division, as shown on Figure REPAVS by providing a test signal in 1E division under test in the REPAVS. k 2.15.6-2. In the REPAVS, , independence .is

, only one Class 1E division at a time. b. In the REPAVS, physical separation or E, provided between Class 1E divisions, and between Class 1E divisions and non-Class b. Inspection of the as-built Class 1E electricalisolation exists between g 1E equipment. divisions in the REPAVS will be Class 1E divisions. Physical b performed. separation or electricalisolation exists between these Class 1E divisions and non-Class 1E equipment.

5. Main control room displays and controls 5. Inspections will be per.ormed on the 5. Displays and controls exist or can be .,43 .

provided for the REPAVS are as defined in main control room displays and controls retrieved in the main control room as ,

Section 2.16.6. for the REPAVS. defined in Section 2.15.6. - / -;

6. The pneumatically-operated REPAVS 6. Tests will be conducted on the as-built 6. The REPAVS isolation dampers shown on ,

isolation dampers, shown on Figure REPAVS pneumatic isolation dampers.

~

Figure 2.15.6-2 fail to the closed position pjd 2.15.6-2, fail to the closed position in the on loss of pneumatic pressure or loss of B,' y. ,s event of loss of pneumatic pressure or loss of electrical power to the valve electrical power to the valve actuating solenoids.

('

o y. q

(

actuating solenoids. E.

S 3 - p m

a e 9 : .

A p:4 y

r 7 n.

O Table 2.15.6-3 Controlled Area Ventilation System (CONAVS) M E W 6 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment . Inspections, Tests, Analyses Acceptance Criteria

1. The basic configuration of the CONAVS is 1. Inspections of the as-built system will be 1. The as-built CONAVS conforms with the as shown on Figure 2.15.6-3. conducted. basic configuration shown on p Figure 2.15.6-3. '=

2. The CONAVS maintains a negative 2. Tests will be conducted on the as-built 2. The CONAVS maintains a negative , U.))

pressure in the refueling area relative to CONAVS in the normal mode of pressure in the refueling area relative to %

P=q the outside atmosphere. operation. the outside atmosphere. JM

[.~d- 3. The CONAVS isolation dampers are 3. Tests will be conducted on the CON AVS 3. Upon receipt of a simulated signal, "I~i ~

r *J closed upon receipt of an isolation signal using simulated LD&lS isolation signals. CONAVS isolation dampers are ;

from the LD&lS. automatically closed. 'M y} k

' 4 Each CONAVS isolation damper requiring 4. 4.

'Y electrical power is powered from the c. Tests will be performed on the a. The test signal exists only in the Class y Class 1E dms_on,t as shown on CONAVS by providing a test signalin 1E division under test in the CONAVS. d;

,j Figure 2.15.6-3. In the CONAVS, only one Class 1E division at a time.

independence is provided between Class b. In the CONAVS, physical separation or E, 1E divisions, and between Class 1E d. Inspection of the as-built Class 1E electrical isolation ex.sts between ;

divisions and non-Class 1E equipment. divisions in the CONAVS will be Class 1E divisions. Physical b performed. separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.

p 5. Main control room displays and controls 5. Inspections will be performed on the 5. Displays and controls exist or can be g provided for the CON AVS are as defined main control room displays and controls retrieved in the main control room as 5! in Section 2.15.6. for the CONAVS. defined in Section 2.15.6.

6. The pneumatically-operated CONAVS 6. Tests will be conducted on the as-built 6. The CONAVS isolation dampers shown g isolation dampers, shown on CONAVS pneumatic isolation dampers, on Figure 2.15.6-3 fail to the closed p

= Figure 2.15.6-3, fail to the closed position position on loss of pneumatic pressure or B.

k

}

in the event of loss of pneumatic pressure or loss of electrical power to the valve loss of electrical pcwer to the valve actuating solenoids.

{

ta b actuating solenoids. E I ! S.

i:

B a

n.

" k l

MISSILE - HWS E SHIELD SMOKE & TOXIC GAS MONITORS

= - , .

m ,

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= - - - c---------- ---------- .

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! - " ',I' " -+- --i l ;

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~ " ~

3lNNS NNS!3 NT MISSILE ) l} ( COMPUTER ROOM l [

h

l SHIELD -- --

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~-

RECIRC FANS

,- ,~ ;

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RETURN / EXHAUST

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? SHIELD

-w ,

NOTE 1:

UTILITY EXHAUST FANS gg a f@

THE OUTBOARD ISOLATION DAMPER SOLENOID VALVES ARE POWERED BY CLASSlE DIVISION 1. THE INDOARD ISOLATION DAMPER SOLENOID VALVES ARE

( () hh 4 E

POWERED BY CLASS lE DIVISION 2. V V V SBWR-CDM-81 g AIR STORAGE BOTTLES g G; R g Figure 2.15.6-1 Control Room Envelope HVAC System g-

U %

tu AIR CONDITIONING UNIT r r TO REFUELING T 'A SUPPLY FANS

} J AREA [7l'f FROM HVAC yi}

,-: e, FRESH AIR ,_

.3 INTAKE PLENUM 7

'"04-

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,, %,.; UNIT 'qv

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NOTE 1 NOTE 1 ', NOTE 1 NOTE 1 r k P P P $

f a tre iI/ ( i e i ( TO PLANT JVi Vi J I I J STACK VEN T NNS 3 3 NNS NNS 3 3 NNS r

iP E EXHAUST g FANS E

B. REFUELING AREA S @

NOTE 1: THE OUTBOARD ISOLATION DAMPER E-O SOLENOtD VALVES ARE POWERED BY @4 CLASS 1E. DIVISION 1. THE INBOARD i ISOLATION DAMPER SOLENOID VALVES C3 b ARE POWERED BY CLASS 1E. OlVISION 2. 3 s c*

I g SBWR-CDM42 Rl n ~

e g 2 Figure 2.15.6-2 Spent Fuel Pool Area Ventilation System

, E-

$ M 2 II3 AIR CONDITION -

p p

[ FROM HVAC UNIT TO FROM TO PLANT N g ( FRESH AIR CONAVS CONAVS (

1 CONVAS EXHAUST fVENT 2 # INTAKE SUPPLY FANS AREAS AREAS FANS STACK k PLENUM (" }

l AIR CONDITION v[,Z'}

)

UNIT @p ._,AJ

{

..d D k TO PLANT "ww 31 SAFETY ENVELOPE EXHAUST FANS fVENT STACK ec+ rf FROM SAFETY SAFETY ENVELOPE AREA FROM SAFETY ENVELOPE AREAS g y ENVELOPE E O b

FROM CONAVS NOTE 3

p NOTE p

3 hAREA NOTE 1

NOTE p

1 FROM E

{ CONTAINMENT PURGE TO PLANT

"'?] AIR h hhSAFETY / l-- h I

EXHAUST EXHA ST _bVENT SYSTEM FILTERS Q" ENVELOPE FANS FROM (

'"y'"i NNS 3 3 NNS NNS 3 3 NNS CONAVS>

AREAS (h

FROM m )m '

SAFETY ENVELOPE ISOLATION REACTOR (

(TYPICAL) p BUILDING #

NOTE 1: THE OUTBOARD ISOLATION DAMPER CONTROLLED g AREAS =

SOLENOID VALVES ARE POWERED BY CLASS 1E. DIVISION 1. THE INBOARD $

=

ISOLATION DAMPER SOLENOID VALVES ARE POWERED BY CLASS 1E, DIVISION 2. ,t, SBWR-CDM-83 =

9 t i y h '

{ Figure 2.15.6-3 Controlled Area Ventilation System (

R RAI Number: 470.8 Question:

l SSAR Sections 6.2.3 and 6.5.3.3 describe the SilWR safety envelope design and SSAR Table 15.6.9 lists the safety envelope leakage rate and its air mixing efficiency as 25 percent per day and 50 percent, respectively. The leakage rate and air mixing efliciency are used in mitigation of offsite and control room operator radiological consequence assessments. Section 6.2.3 further states that the safety envelope is designed to he capable of periodic testing to assure that performance requirements are met.

Provide (1) the safety envelope free air volume, (2) detailed technical justifications for the 50-percent air mixing efficiency assumed and (3) the, leakage testing criteria, frequency, and procedures. The staff will require the SilWR Tier 1 Design Certification Document (ITAAC) to specify the safety envelope leakage rate and air mixing efficiency and the COL applicant technical specifications to specify the periodic integrated leakage rate testing.

GE Response:

(1) The safety envelope free air volume is approximately 7,900 m3 (2) An appendix will he added to chapter 15 detailing the technical justification of the 50 per cent mixing cHiciency in the safety envelope.

, (3) When the Tier I DCD is amended it will incorporate a test to verify the 25 4

per cent per day leakage tightness on the safety envelope. A draft copy of this material is attached The operational requirement is found in section >

16.3.6, SR 3.6.4.1.3 for safety envelope leakage testing.

Also please see Reply to RAI 470.30.

f I

25AS354 Rev. A SBWR certitied oesign Materiai g.g p gn 2.15.12 Reactor Building Structure 4F $

h.."s . .

Design Description The reactor building (RB) is a structure which houses and provides protection and support for the reactor systems, reactor support and safety systems, containment, refueling and spent fuel storage areas and equipment, main steam tunnel, MCR and other control areas, auxiliary area, health physics, laboratories, and much of the plant safety-related equipment. Figures 2.15.12-la through 2.15.12-lh show the basic configuration and scope of the RB*

The reactor building stmeture is integrated with that of a stepped cylindrical reinforced concrete containment vessel (RCCV); the RCCV is located on a common basemat and surrounded by three concentric boxes: the inner hox (safety envelope), the intermediate steel frame, and the outer box. The inner and outer boxes are made of )

reinforced concrete shear walls and the intermediate steel frame is made of structural steel framework with non-structuralwalls as required for radiation shielding, separation, etc. The building is partially embedded. The top of the RB basemat is located 16Am i 0.3m below the finished grade elevation.

All SBWR safety-related equipment is housed in the reactor building safety envelope, main steam tunnel, and pools located beneath the operating Door, with the non-safety-related systems and areas (including the MCR) surrounding this envelope. The safety envelope is leak tight for holdup and decay of fission products that may leak from the containment after an accident &he safety envelope holdup capability decreases releases W 7,e to the atmosphere by limiting leakage to 25% of the safety envelope free volume per day W P My at differential pressures of up to 6 mm of waterffhe building and systems are also

~~

arranged to separate clean and potentially contaminated areas, with separate stairway and elevator service for each area.

To protect against internal flood damage, the following design features are provided:

a Flood water in one division is prevented from propagating to other divisions by walls and floors as shown on Figures 2.15.12-la through 2.15.12-lh.

m Openings between divisions that are located below grade have watertight doors and sills. Other openings between divisions have sills. l m Disisional walls below grade are at least 0.8 meters thick. j

'i l

a Each divisional area has drains to channel 11ood water to its respective divisional below grade area.

The overall building dimensions prmided in Figures 2.15.12-la through 2.15.12-lh are provided for informadon only and are not intended to be part of the certified SBWR information.

Reactor Building Structure - Amendment 1 sQ g'r. n s r ., my 2.15.12 1 1

g e $$ h

,c.p3[N

o. Q ua7" t s

I'

25A5354 Rev. A SBWR certinadoesign uateriai 73 g ,- - g

{' N) h a Equipment necessary for safe shutdown is located above the maximum flood level or is qualified for flood conditions.

Watertight doors on RCCWS, FAPCS, and RWCU/SDCS rooms have open/close sensors with status indication and alarms in the main control room.

For fire protection, the RB prmides three hour fire barriers for separation of the four independent safe shutdmyn divisions.

The general building arrangement including watertight doors and sills for doonvays where needed for flood control is presented in Figures 2.15.12-la through 2.15.12-lh.

To protect against external flood damage, the following design features are provided:

a External wall thickness below flood level equal to or greater than 1.4 meters to prevent ground water seepage.

= No access openings in external walls below flood level.

m Penetrations in external walls below flood level provided with flood protection features.

s A tunnel connects the Radwaste Building, Turbine Building and Reactor Building for the liquid radwaste system piping. The penetrations from the tunnel to the Reactor Building will be watertight.

The reactor building is classified as Seismic Category 1. It is designed and constructed to accommodate the dynamic and static loading conditions associated with the various loads and load combinations which form the structural design basis. The loads are those associated with: ,

(1) Natural p'ienomena-wind, floods, tornados (including tornado missiles),

earthquakes, rain and snow.

(2) Internal events-floods, pipe breaks and missiles.

(3) Normal plant operation-live loads, dead loads, temperature effects and building vibration loads.

w) :. >

4., ..we s i' li b

!j $. gr..(f

% ? -J A"

( L{ke[" *I 2.15.12 2 Reactor Building Structure - Amendment 1

qwm 8 r]rZ'*q ncn2

, , .M

$' 'T~ Table 2.15.12-1 Reactor Building M a '

CO

[ r ~ 4 Design Commitment inspections. Tests, Analyses Acceptance Criteria -

h 1. The basic configuration of the RB is 1. Inspections of the as-built stmeture will 1. The as-built RB conforms with the basic d shown on Figures 2.15.12-1a through be conducted. configuration shown in Figures 2.15.12-1a y 2.15.12-1 h. through 2.15.12-1 h.

e ~

l 2.~The safety envelope pressure boundary has a leak rate equal to or less than 25%

2. An integrated leak rate test of the safety envelope will be conducted,

2. The safety envelope pressure boundary has a leak rate equal to or less than 25%

[ per day of the contained gas volume at a per day of the contained gas volume at a I 3:

! differential pressure equal to or less than maximum differential pressure of 6 mm

& 6 mm of water. of water.

3 i 3. The top of the RB basemat is located 3. Inspections of the as-built structure will 3. The top of the RB basemat is located 16.4m 10.3m below the finished grade be conducted. 16.4m 0.3m below the finished grade elevation. elevation.

4. Inter-divisional walls, floors, doors and 4. Inspections of the as-installed inter- 4. The as-installed walls, floors, doors and penetrations in the RB have a three-hour divisional boundaries will be conducted. penetrations that form the inter- M fire rating. divisional boundaries have a three-hour $

fire rating. t

=

5. The RB has divisional areas with walls 5. Inspections of the as-built walls and 5. The as-built RB has walls and watertight 2 and watertight doors as shown on Figures watertight doors will be conducted. doors as shown on Figures 2.15.12-1a b 2.15.12-la through 2.15.12-th. through 2.15.12-1 h.

6. Main control room displays and alarms 6. Inspections will be performed on the MCR 6. Displays and alarms exist or can be provided for the RB are as defined in displays and alarms for the RB. retrieved in the MCR as defined in Section 2.15.12. Section 2.15.12.

7. A flooding event doe., not affect more 7. Inspections will be conducted of the 7. Equipment necessary for the safe than one division of safety-related location of equipment necessary for the shutdown of the plant is located above equipment. safe shutdown of the plant. the maximum flood height or is qualified for flood ccnditions. p B.

?

4 I%f-p-3&f ff t- ?:;-m R' gbj E 9

m%-

3:

!!t g dw~ Q G #p- k i

m

t Table 2.15.12-1 Reactor Building (Continued) m W

{ Design Commitment inspections, Tests, Analyses Acceptance Criteria g

8. The R/B is protected against external 8. Inspections of the as-built structure will 8. D floods by having: be conducted. a. External walls below flood level are

a. External walls below flood level that equal to or greater than 1.4m thick to are equal to or greater than 1.4m thick prevent ground water seepage. f"]r' wz to prevent ground water seepage.

b. Penetrations in the external walls -

b. Penetrations in the external wa:Is below flood level are provided with m -.

below flood level provided with flood protection features.

flood protection features. [M

c. Penetrations from the tunnel to the ""P I Watertight penetrations to the Reactor Reactor Building are watertight.

c.

Building from the tunnel that ;

connects the Radwaste Building.

Turbine Building and Reacto; Building for the liquid radwaste system piping. ,

w

9. The RB is able to withstand the structural 9. A structural analysis will be performed 9. A structural analysis report exists which P,;

design basis loads as defined in Section which reconciles the as-built data with concludes that the as-built RB is able to $

2.15.12. structural design basis as defined in withstand the structural design basis if Section 2.15.12. loads as defined in Section 2.15.12. (

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RAI Number: 470.9 Question:

SSAR Section 6.5.3.3 describes lission-product holdup, as well as the platcout mechanism, in the safety envelope. However, the staff noticed that holdup and .

raixing credits (fbr decay) are claimed, but that no credit is taken for fission.

product platcout. Clarify this apparent discrepe y.

GE Respone:

No credit is taken in the I.OCA DilA analysis for fission product plateout in the safety envelope or in the reactor building. This approach is conservative as it is cicar, based upon considering residence times and room conditions (temperature, humidity, and pressure), that fission product platcout will occur.

and will be significant in the safety envelope and reactor building.

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RAI Nmnber: 470.10 Quesdon:

SSAR Section 6.2.2 describes the passive containment cooling system (PCCS). '

The PCCS removes the core decay heat rejected to the containment after a .

1.OCA. Should the PCCS heat exchanger tubes fitil, the PCCS will provide a ;

potential bypass pathway for the SIMR containment, releasing radioactive '

fission-products from the containment atmosphere to the reactor building through the passive containment cooling pool water.. Provide the radiological i consequence assessments, complete with the major assumptions and ;

parameters used, for the PCCS heat exchanger tube fitilure. .;.

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GE Responre:

No failure or leakage of this system is anticipated based upon the conservatisms l in design, manuliscture, and use. The PCCS tubes are high quality tubing ,

designed to pressures twice the maximum operational requirement. In addition i these tubes are subject to regular test under AppendixJ of 10 CFR 50.

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Radiological Consequences of DBAs (Chapter 15)

RAI Number: 470.11 Question:

SSAR Section 15.6.2 describes failure of a small line carrying primary coolant. ,

outside containment, and SSAR Table 15.6-1 lists the assumptions and parameters used in the radiological consequence assessment for this failure.

Provide the technical bases, complete with applicable references, for the following assumptions:

(1) the period of 10 minutes for the operator to detect the event (2) the period of 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for the operator to scram the reactor to reduce l reactor pressure (3) 13,000 kg of reactor primary coolant released into the reactor building (4) 5,000 kg (out of 13,000 kg losO icactor primary coolant being flashed to stean!

(5) magnitudes of iodine spiking (6) iodine plate out fraction of 50 percent (7) reactor building leak rate of 200 percent per hour GE Response:

The basis for the instrument line break accident is found in GE Report NEDO-21143-1 (SSAR Reference 15.6-2) which assumes the following:

(1) A small line containing primary coolant ruptures in the reactor )

huilding. For purposes of evaluation the small line chosen was an instrument l line. All small lines without isolation valves are required to have a one-quaniu l inch flow restrictor installed in them to limit flow from the line. Following the I break, within one minute the releasing primary coolant will alarm the control room via sensors (temperature or samp flow) found in all rooms where primary containment lines are located. The ten minute period fc.r operator action is an assumption permitting the operator suflicient time to analyze the occurrence prior to action on the operators part (2) The time period of 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is the time necessary to bring a typical inVR from full power operations to cold shutdown.

(3) The 13,000 kg of fluid is the calculated mass release through a one-quarter inch orifice for a period of six hours to bring a InVR at initial full power

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conditions to cold shutdown (see graphs in NEDO-2114S1, SSAR Reference 15.62).

(4) The mass of Duid Hashed to steam is 2.270kg (5,000 pounds). See ;

attached revised SSAR Table 15.G1. It is based upon the pressure and temperature of the released Huid for a full shutdown event and is based upon the cmves presented in NEDO-2114SI (SSAR Reference 15.42). ;

(5) GE has established a spiking release value per bundle fbr IlWRs (APED-5756, SSAR Reference 15.48) which is assumed available for release in a f depressurization event. Liitially ten percent of this inventory is released to the primary vessel water at accident initiation. The remaining ninety percent is released proportional to the decrease in pressurization during the accident.

Details of the analytical model are presented in NEDO-2114Sl(SSAR Reference 15.42).

(6) The iodine plate out fraction is based upon the expected release pointin ,

that all such lines are located in the lower safety envelope or reactor building ,

and must traverse the HVAC or several Doors of reactor building prior to exiting .

to the environment. The condensation of the Dashed steam traversing this pathway will remove at least 50 percent without consideration of other natural mechanisms and is typical for events of this type based upon SRP 15.4.9,

Appendix A,111.11 for releases into large cool volumes.

(7) The reactor building turn over rate is one and one half air exchanges per hour. The release rate of two exchanges per hour is a conservative assumption t considering that a release of such magnitude will not overwhelm the HVAC' for any extended period of time. However more recent evaluations indicate that the air exchange rate of 1.5 exhanges pe- hour (actually 1.36 exchanges per hour) do not correctly reDect the air exchange rate in the safety envelope, which is 2.35 exchanges per hour. The analysis will therefbre be revised in a future amendment to an air exchange rate of three exchanges per hour. This should consititute only a fractional change in the resulting dose.

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25A5113 Rev. A SBWR Standard Safety Analysis Report -

Table 15.6-1 Instrument Small Line Break Accident Parameters l Data and assumptions used to estimate source terms A. Power level 2040 MWt B. Mass of fluid released 13,000 Kg (28,600 lbm)

C. Mass of Fluid flashed to steam 57 000 2270 Kg (4h000 5ppQ Ibm)

D. Mass of fluid in reactor 246,600 Kg (544,000 lbm)

E. Duration of accident 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months F. Number of bundlesin core 732 G. Number of rods per bundle 60 H. Ime from SCRAM to lodine Gpike 0.0 sec 11 Data and assumptions used to estimate activity released A. lodine water concentration Subsection 15.6.4.5, case 2 B. lodine Spiking 1-131 7.8E+4 MBq/bundie (2.1 Ci/ bundle) 1-132 1.2E+5 (3.2) 1-133 1.9E+5 (5.1) 1-134 2.0E+5 (5.4) 1-135' 1.8E+5 (4.8)

C. lodine plateout fraction 50%

D. Reactor Building Flow rate 200%/ hour E. SGTS Filter Efficiency none assumed F. Time from SCRAM to lodine Spike 0.0 sec til Dispersion and Dose Data A. Meteorology Table 15.6-3 B. Site Boundary distances Table 15.6-3 C. Method of Dose Calculation Reference 15.6-2 D. Dose conversion Assumptions RG 1.109,and Fed Guide Report 11 E. Activity Inventory / releases Table 15.6-2 F. Dose Evaluations Table 15.6-3 Deeresse in Reactor Coolant inwntory- Amendment 1 DRAFT 15.6 23 2.2584

2EAS113 Rev. A SBWR Standant Safety Anclysis Report Table 1.6-1 Referenced Reports SBWR SSAR Report No. Title Section No.

22A7007 GESSAR 11,238 Nuclear Island, BWR/6 Std Plant, General 3.7, 19.2, 19.3, Electric Company,3/82, & Amendments 1-21. 19D3,19D.7, 19E.2,19E.3 APED-5750 Design and Performance of GE-BWR Main Steam Line 5.4 Isolation Valves, General Electric Company, Atomic Power Dept., 3/69 APED-5756 N, R, Horton, W. A. Williams and K-4% J. W. Holtzclaw, 6 15m6,15.7 Analytical Methods for Evaluating the Radiological Aspects of General Electric BWRs, S/-76 3/$2 NEDO-10029 An Analytic Study on Brittle Fracture of GE-BWR Vessel 5.3 Subject to the Design Basis Accident NEDO-10871 J.M.Skarpelos and R. S. Gilbert, Technical Derivation of 12.2.

BWR 1971 Design Basis Radioactive Materials Source Terms,3/85 NEDE-10958-PA, GE BWR Thermal Analysis Basis (GETAB): Data 48.16 NEDO-10958-PA Correlation and Design Application,1/77 NEDO-11209-04A Nuclear Energy Business Operations Quality Assurance 17.1 Program Description, Rev. 8,3/89 NEDO-20159 BWR Radioactive Waste Treatment System,896 15.7 NEDO-20206 D. R. Rogers, BWR Turbine Equipment N-16 Radiation 12.2.

Shielding Studies,1203 NEDO-20533 W. J. Bilanin, The GE Mark lil Pressure Suppression 6.2.7 Containment Analytical Model,6/74 NEDO-20533-1 W. J. Bilanin, The GE Mark 111 Pressure Suppression 6.2.7 Containment Analytical Model, Supplement 1,995 NEDE-20566 Analytical Model of Loss of Coolant Accident in 3.9.7 Accordance with 10CFR50, Appendix K, Proprietary Document,11R5 NEDE-20566-P-A Analytical Model of Loss of Coolant Accident in 6.3.7 Accordance with 10CFR50, Appendix K,9/86 NEDM-20609-01 P. P. Standcavege and D. G. Abbott, Liquid Discharge 12.2.

Doses - LIDSR Code,8/76 NEDO-21052 F. J. Moody, Maximum Discharge Rate of Liquid-Vapor 6.2.7 Mixtures from Vessels, General Electric Company,995 NEDO-21143-1 H. Careway, V. Nguyen, and P. Stancavege, Radiological 15.7 Accident - The CONACO3 Code,12/81 NEDO-21159, Airborne Releases from BWRs for Environmental impact 12.2 3 NEDO-21159-2 Evaluations,12/84 NEDE-21175-P BWR/6 Fuel Assembly Evaluation of Combined SSE and 3.9.7 Loss of Coolant Accident Loadings,1196 NEDE-21354-P BWR Fuel Channel Design and Deflection,996 3.9.7 1.6-2 Materialincorporated by Reference - Amendment 1 DRAFT NTN.1

l-2SAS113 Rev. A 1

SBWR Standant Safety Analysis Report The totalintegrated mass of fluid released into the reactor building is 13,000 kg (28,660 lbs) with approximately 5,000 kg (11,000 lb) being flashed to steam.

15.6.2.5 Radiological Analysis General The radiological analysis is based upon conservative assumptions considered acceptable to the NRC. Though the SRP does not provide detailed guidance, the assumptions found in Table 15.6-1 assume that all of the iodine available in the flashed water is transported via the reactor building IIVAC system to the environment without prior treatment. Other isotopes in the water contribute only negligibly to the iodine dose.

Fission Product Release ne iodine activity in the coolant is assumed to be at the maximum equilibrium Technical Specification limit (see Design Basis Analysis in Subsection 15.6.4.5, case 2) for continuous operation. The iodine released to the reactor building atmosphere and to the environment are presented in Table 15.6-2 (References 15.6-2 and 15.6-8).

Results Results of the analysis are found in Table 15.6-3 and are within the 10% of 10CFR100 '

specified in the SRP.

15.6.3 Steam Generator Tube Failure This section is not applicable to the direct cycle SBWR.

15.6.4 Steam System Piping Break Outside Containment ,

This event involves postulating a large steam line pipe break outside containment. It is assumed that the largest steam line, instantaneously and circumferentially breaks at a location downstream of the outermost isolation valve. The plant is designed to immediately detect such an occurrence, initiate isolation of the broken line and actuate ,

the necessary protective features. This postulated event represents the envelope evaluation ofsteam line failures outside containment.

15.6.4.1 Identification of Causes and Frequency Classification

, identification of Causes A main steam line break is postulated without the cause being identified. These lines are designed to high quality engineering codes and standards, and to seismic and erwironment;d requirements. However, for the purpose of evaluating the consequences i of a postulated large steam line rupture, the failure of a main steam line is assumed to l occur. j Decmase in Reactor Coolant inventory- Amendment 1 DRAFT 15.6-3 ,

2/27/94

25A5113 Rev. A SBWR standard s.:!ety Analysis neport 15.6-5 Ramsdell,J.V., Alternatives to Current Procedures Used to Estimate Concentrations in Building Wakes,21st DOE /NRC Nuclear Air Cleaning Conference, pgs 714-729.

15.6-6 "Revred Accident-Source-Terms-for44ght-Water-Nuclear 44wer-Plants"r-SEGW92424 April 40,4992, Advanced Licht Water Reactor Utility Reauirements Document. Volume II.L.

Electric Power Research Institute.

15.6-7 C.C. Lin, Anticipated Chemical Behavior ofIodine under LOCA Conditions, January,1981, NEDO-25370.

Lifte N. R. Horton. W. A. Williams. I. W. Holtzclaw. Anahtical Methods for Evaluatine the Radiolocical Aspects of the General Electric Boiline Water Reactor. March 1969. APED-5756.

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25A3113 Rev. A l SBWR standant sarety Analysis Report the level of the r efueling floor to ground level through the refueling floor maintenance hatch.

15.7.5.2 Radiological Analysis The largest size of BWR fuel cask is conservatively assumed to be dropped approximately 24 meters from the refueling floor level to ground level on transport from the decontamination pit out of the reactor building. It is conservatively assumed that all fuel rods are damaged and the fission gases in the fuel rod gap space are released to the reactor building and then to the environment over a two hour period.

Table 15.7-8 provides the assumpdons for this analysis andTable 15.7-9 the radiological consequences. As can be seen from Table 15.7-9, the radiological releases are well within guidelines.

15.7.6 COL License Information None.

15.7.7 References 15.7-1 D. Nguyen, et.al., Radiological Accident Evaluation -The CONAC03 Code, December 1981 (NEDO-21143-1).

15.7-2 N.R. I f orton, W.A. Williams, and LWrjX Holtzclaw, Analytical Methods for Evah2ating the Radiological Aspects of General Electric Boiling Water Reactors March 19761969 (APED-5756).

15.7 8 Radioactivo Release - Amendment 1 DRAFT S 219 4

RAI Number: 470.12 Question: l 470.12 In the response to Part (5) of Question 470.16 above regarding iodine spiking, state the chemical forms ofiodine assumed to spike and state the reasons for not considering spiking of other nucades such as cesium.- ;

GE Response: !

It is assumed that the above questions refers to question 470.11 and not the 470,16 ,

stated. The chemical form assumed is 12 which is typically found in fuel rod gaps under normal conditions. Cesium is not typically found in spiking ,

activity or in the fuel rod gap since the release of cesium from the fuel rod matrix requires temperatures and conditions far above normal operating conditions or those experienced during a depressuritation event.

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RAI Number: 470.'13 i;,.

Question:

In the titles of SSAR Tables 15.G-1 through 15.6-3, change " Instrument" to "Small," so that the titles will read "Small Line 15reak Accident Parameters." :

4' t GE Response; .i 4

See attached revised SSAR Tables 15.41 through 15.6-3. l s

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2SAS113 Rsv. A SBWR standard sakty Anar rsis aepois Table 15.6-1 Instrument S_mA Line Break Accident Parameters l Data and assumptions used to estimate source terms i

A. Power level 2040 MWt B. Mass of fluid released 13,000 Kg (28,600 lbm)

C. Mass of Fluid flashed to steam 57000 2229 Kg (44000 5000 lbm)

D. Mass of fluid in reactor 246,600 Kg (544,000 lbm)

E. Duration of accident 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months i F. Number of bundles in core 732 G. Number of rods per bundle 60 H. Time from SCRAM to lodine Spike 0.0 sec 11 Data and assumptions used to estimate activity released i

A. lodine water concentration Subsection 15.6.4.5, casa 2 l B. todine Spiking 1-131 7.8E+4 MBq/ bundle (2.1 Ci/ bundle) 1-132 1.2E+5 (3.2) 1-133 1.9E+5 (5.1) 1-134 2.0E+5 (5.4) 1 1-135 1.8E+5 (4.8)

C. lodine plateout fraction 50% )

D. Reactor Building Flow rate 200%/ hour E. SGTS Filter Efficiency none assumed F. Time from SCRAM to lodine Spike 0.0 sec ill Dispersion and Dose Data i

A. Meteorology Table 15.G 3 j B. Site Boundary distances Table 15.6-3 l C. Method of Dose Calculation Reference 15.6-2 D. Dose conversion Assumptions RG 1.109,and Fed Guide Report 11 E. Activity Inventory / releases Table 15.6-2 F. Dose Evaluations Table 15.6-3 l

Decrease In Reactor Coolant inventory - Amendownt 1 DRAFT 16.6-23 2s2Sv94

. 1 25A5113 Rev. A SBWR standard sarery Anarrsts Report Table 15.6-2 RBA SLBA lsotropic inventory ISOTOPE 30-sec 10. min 30-min 1-hour 2-hour 4-hour 6-hour 8-hour A. REACTOR BUILDING INVENTORY IN MBq 1-131 2.2E+1 2.5E+1 1.2E+3 1.2E+3 1.2E+3 7.5E+2 0. O.

1-132 2.1 E+2 2.4E+2 1.8E+3 1.8E+3 1.8E+3 1.1 E+3 0. O.

1-133 1.5E+2 1.7E+2 2.8E +3 2.8E+3 2.8E+3 1.8E+3 0. O.

1-134 4.2E+2 4.8E+2 3.0E+3 3.0E+3 3.0E+3 1.9E+3 0. O.

1-135 2.2 E+2 2.5E+2 2.7E+3 2.7E+3 2.7E+3 1.7E+3 0. O.

TOTAL 1.0E+3 1.2 E+3 1.2E+4 1.2 E+4 1.2E+4 7.3E+3 0. O.

B. REACTOR BUILDING INVENTORY IN CURIES l-131 5.9E-4 6.8E-4 3.2E-2 3.2E-2 3.2E-2 2.0E-2 0. O.

I-132 5.8E-3 6.6E-3 4.9E-2 4.0E-2 4.9E-2 3.1 E-2 0. O.

1-133 4.1 E-3 4.6E-3 7.6E-2 7.6E-2 7.6E-2 4.8E-2 0. O.

1-134 1.1 E-2 1.3E-2 8.2E-2 8.2E-2 8.2E-2 5.2E-2 0. O.

1-135 5.9E-3 6.8E-3 7.3E-2 7.3E-2 7.3E-2 4.6E-2 0. O.

TOTAL 2.8E-2 3.2E-2 3.1 E-1 3.1 E-1 3.1 E-1 2.0E-1 0. O.

C. ISOTOPIC RELEASE TO ENVIRONMENT in MBq I-131 3.0E+1 1.0E+3 9.8E+4 2.5E+5 5.4E+5 9.6E+5 1.1 E+6 1.1 E+6 l-132 2.9E+2 9.9 E+3 1.6E+5 3.8E+5 8.3E+5 1.5E+6 1.7E+6 1.7E+6 l-133 2.1 E+2 7.0E+3 2.4E+5 5.9E+5 1.3E+6 2.3E+6 2.6E+6 2.6E+6 l-134 5.8E+2 1.9E+4 2.7E+5 6.5E+ 5 1.4E+ 6 2.5E+6 2.8E+6 2.8E+6 l-135 3.0E+2 1.0E+4 2.3E+ 5 5.7E+5 1.2E+6 2.2E+6 2.5E+6 2.5E+6 TOTAL 1.4E+3 4.8E+4 1.0E+6 2.4E+ 6 5.3E+6 9.4E+6 1.1 E+7 1.1 E+7 D. lSOTOPIC RELEASE TO ENVIRONMENT IN Ci 1-131 8.2E-4 2.8E-2 2.6E+0 6.6E+0 1.5E+1 2.6E+1 3.0E+1 3.0E+1 1-132 7.9 E-3 2.7E-1 4.3E+0 1.0E+1 2.3E+1 4.0E+1 4.6E+1 4.6E+1 1-133 5.6E-3 1.9E-1 6.5E+0 1.6E+1 3.5E+1 62E+1 7.1 E+1 7.1E+1 1-134 1.6E-2 5.3E-1 7.3E+0 1.7E+1 3.8E+1 6.7E+1 7.7E+1 7.7E+1 1S.6-24 Decrease in Reactor Coolant Inventory- Amendment 1 DRAFT 2/2SR4 ,

.-25A5113 Rev. A SBWR snandard sentry Analysis Report Table 15.6-2 lLSA SLBA Isotropic inventory (Continued)

ISOTOPE 30-sec 10-min 30-min 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 2-hour 4-hour 6-hour 8-hour I-135 8.2E 2.8E-1 6.3E+0 1.5E+1 3.4E+1 6.0E+1 6.8E+1 6.8E+1 TOTAL 3.8E-2 1.3E+0 2.7 E+1 6.6E+1 1.4E+2 2.5E+2 2.9E+2 2.9E+2 Decaease in Reactor Coolant inventory- Arnendment 1 DRAFT 15.6-25 225,94

,o 25AS113 Rev. A I

SBWR Standard Safety Analysis Report i Table 15.6-3 ILBA SLBA Results- Meteorology and Dose Results Meteorology Distance Thyroid Dose Whole Body Dose (sec/m3) (m) (Sv) (Sv) 1.8E-3 max' 3.0E-1 2.1 E-3 1.0E-3 800 1.6E-1 1;2E-4

" max" = maximum meteorology to meet 10% of 10CFR100 limits.

Table 15.6-4 Sequence of Events for Steam Line Break Outside Containment Time (sec) Event 0 Guillotine break of one main steam line outside containment.

0.5 High steamline flow signalinitiates closure of MSIVs

< 1.0 Reactor begins scram.

<2.0 Partial closure (15%) of MSIV initiates isolation condensers.

<5 MSIVs fully cf osed.

10 Reactor low water Level 2 is reached. Isolation condensers receive second initiation signal.

32 Isolation condensers in full operation. Water level stabilized.

435 SRVs open on high vessel pressure (if isolation condensers are not available). The SRVs open and close to maintain vessel absolute pressure at approximately 8.5 MPa (1215 psia).

3540 Reactor low water Level 1 is reached (if isolation condensers are not available).

ADS timer initiated.

3550 ADS timer timed out. ADS actuation sequence initiated. GDCS timer initiated.

3700 GDCS timer timed out. GDCS injection valves open.

3880 Vessel pressure decreases below shutoff head of GDCS. GDCS reflooding flow into the vessel begins.

The core remains covered throughout the transient and no core heatup occurs.

1S.6-26 Decrease In Reactor Coolant inventory- Amendment 1 DRAFT 2N5M

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RAI Number: 470.14 .;

- Question:

SSAR Section 15.6.4 describes a main steam pipe break accident outside containment and SSAR Table 15.6-5 lists the major assumptions and parameters used in the radiological consequence assessment of this accident. Provide the technical bases, complete with applicable references, for the following ,

parameters used: ,

(1) air exchange rate of 6000 per day in the steam tunnel (provide the free air volume of the steam tunnel and break location)

(2) 12,000 kg of steam mass released (3) 2,400 kg of water mass released (4) iodine concentration in the reactor coolant based on offgas release rates of 0.2 Ci/sec and 0.05 Ci/sec.

GE Response:

The technical basis for the main steamline break accident is found in GE Report NEDO-21143-1 (reference 15.6-2), which discusses the basis for and algorithms used in evaluating the accident. The SSAR is the technical .

documentation for the evaluation. !

(1) The air exchange rate of 6000 is an assumption used to describe rapidly venting fission products from the steam tunnel, it is not considered important as to where the break is in the steam tunnel since the tunnel vents to the lower '

floors of the turbine building and then througl 3grates into the upper volume _of the turbine building. Since the volume of steam is subject to extensive condensation along the release pathway, representing the release by a large turn over rate is reasonable. The steam tunnel volume in the reactor building is approximately 950 m3.

(2/3) The masses of steam and water released are calculated by GE transient ';

analysis codes for the specific case of the SilWR MSL break with MSIVs closing '

within 5 seconds plus the steam line vohune at the time of the accident.

(4) The iodine concentration of 0.2 mci /gm is the standard BWR technical specification (WASH-123) for normal operations. This value is referred to as Case 2. The maximum technical specification of 4.0 mci /gm is referred to as Case 1 and assumes significant fuel failures with iodine spiking. The noble gas ,

olTgas release rate of 0.05 Ci/second is a conservative estimate of that provided in NUREG-0016 and ANS Standard 18-1. The maximum noble gas limit of 0.2 ,

Ci/sec is derived from SRP 11-2 as equivalent to 100 mci /sec/MWt for a 2,000 MWt plant. ,

RAI Number: 470.15 Question: ,

-)

i SSAR Section 15.6.5.5 states that except where noted, the reactor accident source terms used are consistent with those speciHed in the EPRI Passive Requirements Document. Itemi/c and list in a table the exceptions taken.

GE Response:

There are no exceptions in the February 1993 submittal. See attached revised Table 15.6-9.

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. l 25AS113 Rev. A l SBWR standantsarery Analysis Repott l 15.6.5.3 Core and System Performance MathematicalModel The analytical methods and associated assumptions which are used in evaluating the consequences of this accident are considered to provide conservative assessment of the expected consequences of this improbable event.

The details of these calculations, theirjustification, and bases for the models are developed in Sections 6.3, 7.3, 7.6, 8.3.

Input Parameters and initial Condition input parameters and initial conditions used for the analysis of this event are presented in Table 6.3-1.

Results Results of this event are presented in detailin Section 6.3. The temperature and pressure transients resulting as a consequence of this accident are insufficient to cause perforation of the fuel cladding. Therefore, no fuel damage results from this accident.

Post-accident tracking instrumentation and control is assured. Continued long-tenn core cooling is demonstrated. Radiological impact is minimized and within limits.

Continued operator control and surveillance is examined and provided.

15.6.5.4 Barrier Performance The structural design basis for the containment is to maintain its integrity and experience normal stresses after the instantaneous rupture of any pdmary system piping within the structure, while also accommodating the dynamic effects of the pipe break and an SSE. Therefore, any postulated LOCA does not result in exceeding the con tainment design limit (see Sections 3.8.2.3,3.6, and 6.2 for details and results of the analpes).

15.6.5.5 Radiological Consequences Two specific analyses are provided for the evaluation of the radiological consequences of a design basis Loss of Coolant Accident (LOCA), one for offsite dose evaluations and the second for control room dose evaluations. Both analyses are based upon assumptions provided by the ALWR URD h+-SEGV42-147 (Reference 15.64). except-wherenoted,The analysis is based upon a process flow diagram shown in Figure 15.6-2 and accident parameters specified in Table 15.6-9.

Radionuclide Releases and Pathways Although a LOCA in the SBWR does not result in any fuel damage as discussed in Subsection 15.6.5.3, a case with core melt with the corium deposited in the lower drywell is analyzed as specified by Reference 15.6-6. Radionuclide releases are based upon Table 44LIk7 of Volume HL Chaoter 5 of Reference 15.6-6 in that it is assumed Decrease in Reactor Coolant Inventory- Amendment 1 DRAFT 153 9 2MSN4

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25A5113 Rev. A

_ SBWR standardsafety Analysis Report l q

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T=b!: 15.S-9 Lee: ef C !:nt Are! dent Per mete e Table 15.6-9 Loss-of-Coolant Accident Parameters '

l. Data and Assumptions used to estimate source terms. .

A. Power Level 2040 MWt i

B. Fraction of Core Inventory Released Early Late Noble Gas 80% 20% 5 lodine 30% 20%

Cesium 23% 18 % I Tellurium 6% 2%Rq Strontium, Barium 0.3% q Ruthenium 0.03% 0.3%

Others 0.003 %

C. lodine Chemical Species Particulate 99.85%

Organic 0.15 %

D. Suppression Pool Decontamination Factors Gap Release Noble Gas 1 Particulates 1 Organic lodines 1 .

11. Data and Assumptions used to estimate activity released A. Core Release Timing Early 1000 to 8,200 seconds Late 8,200 to 83,800 seconds ,

B. In Containment Removal Factors Nobles Gas 0 '

lodine Particulate 0.6/hr ,

Organic 0 All others 0.6/hr C. Submergence Scrubbing Factors -)

Ex-Vessel Release I I

lodine 1 l Decrease in Reactor CoolantInventory- Amendment 1 DRAFT 15.6-33 2/25,94

RAI Number: 470.16 .!

Question:

For each release phase (gap release, early in-vessel, ex-vessel, and late in-vessel) and for each nuclide listed in draft NUREG-1465, provide a comparison table ;

listing the S11WR, EPRI Passive Requirements Document, and draft NUREG- ;

1465 values for fission-product release fractions into the containment following l a DilA.

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- GE Response:

The table below presents the data requested for the I.OCA DIIA analysis which ,

does not consider either ex-vessel or late in-vessel releases.

Earlv Releases 1. ate Releases SilWR/ NURE SBWR/ NURE i

- Elemental Group EPRI G-1465 EPRI G-1465 -

Noble Gas S0% 5% 20 % 95 % l lodine 30 % 5% 20 % 22 % ;

Cesium 23 % 5% 18% 15 %

Tellurium 6% 3% 11 %

Strontium, llarium 0.3% _g 3%

Ruthenium. 0.3% 0.7%

Others l 0.003% 0.2% ' J 9

l RAI Number: 470.17 .

Question:

The primary containment leakage rate (0.475 percent) and hypass leakage ritte ,

i (0.025 percent) provided in SSAR Section 15.6.5.5 should be expressed as those values per weight percents per day rather than in volume percents per day.

GE Response:

The specification (br the containment is 0.5% by volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 55 psig ;

(SSAR Table 6.2-1) and for the safety envelope 25% by volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 6 mm differential pressure (SSAR Section 6.2.3.1) i I

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i RAI Number: 470.18 Question: -

SSAR Section 15.6.5.5 states fission-product release timing and duration's for each release phase. List all SIlWR accident sequences which are considered to- l

- significantly impact the source term and identify the controlling accident scenario and sequence for fuel rod failure (gap release) and fuel melt (early in- !

vessel release). (The accident scenarios should consider but not be limited to , ,

break and non-break of the reactor coolant system, small-break LOCA, large- [

break LOCA, and the leak before break.)

GE Response:

Those design basis accidents required by Regulatory Guide 1.70 and considered for design of the SilWR are given in SSAR section 6.3.

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RAI Number: 470.19 Question:

SSAR Section 15.6.5.5 describes and SSAR Table 15.6-9 lists the amount of -

organic iodide source term as 0.15 percent of the core iodine inventory. This ,

amount is inconsistent with a value developed by the staff and currently undergoing management review. In draft NUREG-1465, the Nuclear.

Regulatory Conunission' (NRC) did not evaluate' the formation of organic '

iodide in the containment following a DilA. The staff realizes, however, that organic iodide can be produced by the reaction of fission product iodine with' organic materials present in the containment. The NRC estimates that no more-than 5 percent of the airborne elemental iodine will be converted into organic species. This amount of organic iodide would correspond to about 0.25 percent of the core iodine inventory (i.e.,5 percent of.5 percent is 0.25 percent). Final

. NUREG-1465 will address this issue. In the interim, GE should reassess the radiological consequences and resubmit for staff review the resulting offsite and control room operator doses using 0.25 percent organic iodide.' Ifit desires, GE ,

n'ay retain the consequence analysis based on the 0.15 percent organic iodide Ihr reference purposes,in addition to the analysis requested herein.

GE Response:

The analysis provided in SSAR Section 15.6.5.5 is done in accordance with the ;

EPRI guidelines (see Reference 15.6-6) which is conservative with respect to'- ;

NURGE-1465 with respect to organic iodide. Thlc 15.6.5.5 anaysis is especially. :i conservative for the BWR when considering that the amount of organic iodide for the BWR is only ~20% of that for a comparable PWR as discussed in Reference 15.6-6. Accordingly, an analysis using .15 is still conservative to a ;

llWR assessment of .25 and the llWR factor of 1/5 in Reference 15.f>6. :

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RAI Number: 470.20 Question:

SSAR Section 15,6,5.5 states that the SBWR design is capable ofinjecting ,

buffering agents into the pool water through various systems to maintain the pH j of the pool water above 7.0. Describe in detail the (1) type and amounts of -

chemicals to be used, (2) provisions designed for chemical injection, (3) '

volum'e of pool water in the containment, (4) provisions designed for pool water mixing, (5) pool water sampling and analysis provisions, (6) pH monitoring, and (7) chemical storage facilities and their locations in buildings.

GE Response: f (1) The specific buffering agents are left to the discretion of the COL applicant IlulTering agents may be added to the skimmer surge tank of the Fuel ;

(2) and Auxiliary Pools Cooling System (FAPCS) as described in Section 9.1.3.

t (3) The volume of pool water in the containment is given in Table 6.2-4.

(4) The pool water will be mixed by the action of the suppression pool filter sys t e m. ;

(5),(6) pH monitoring should not be necessary if sufficient quantities of buffering agents are added to the containment water to assure pH above 8.0 regardless of acidic precipitants. -

l (7) Chemical storage facilities will be the responsibility of the COL .I applicant.

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4 RAI Number: 470.21 Question: ,

Provide calculated plI values of pool water in the primary containment .:

following the controlling accident sequence selected in Question 470.18 above as a function of time over the entine duration of the accident (30 days).

4 GE Response: j The calculated pH values assuming Nitric acid formation as per NUREG/CR-5732 are as follows:

Time (hrs) I pH* I pH *

O S.9" 9.S3 24 8.79 9.70 48 8.59 ] 9.50 72 8.21 l 9. I 1

= 10% of fi sion products are in l

._w etwell

- = 80% of fission products are in wetwell ;

After 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months buffering agents may be added through the~ Fuel and Auxiliary Pools Cooling System (FAPCS) as necessary to keep pH above 8.0.

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H RAI Number: 470.22 Question:

The most important acids in the primary containment following a DllA are nitric acid, produced by irradiation of water and air, and hydrochloric acid, produced by irradiation or heating of electrical cable insulation. State the chemical properties and estimated amounts of electrical cabk insulator used in l the SilWR primary containment. State the calculated amounts of hydrochloric' i acid produced by radiolysis and pyrolysis of electric cable insulators and resulting pool water pH. ,

GE Response: -

The SBWR will not be signiDcantly different from the AllWR except ilnt ,

fewer cables are employed owing to (1) fewer Fine Afotion Control Rod Drives:

in SllWR, (2) two hiain steam lines with theii isolation valves, and generally l fewer components in the drywell than the AllWR. Using the AllWR as a benchmark and assuming that cables were ethylene propylene rubber insulation and a Hypalon jacketing, both of which are chlorinated polymers, ,

and assuming that all the cabling is located in the lower drywell under severe accident conditions (twice desig n basis conditions), the AllWR found 125 gmol of hcl produced in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This combined with the CsOH from the release, and HNO 3production resulted in a pH of 8A after twenty-four hours. We. ,

expect similar results but Ic3s hcl production in the SilWR. Finally, the :

production is a combination of cable types and radiation field. In the above analysis, worst cases were assumed.

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J RAI Number: 470.23 .j Question: )

l SSAR Section 15.6.5.5 describes and SSAR Table 15.6-9 lists the primary j containment aerosol deposition rate of 0.6 per hour. It states that this rate is based on prior analyses of boiling water reactors (llWRs) under similar circumstances. Provide the prior analyses.

GE Response:

The prior analysis was that reported in NUREG-1465, draft, and performed under NRC auspices as part of the study which resulted in the publication of NURECr1150. Since then a SilWR specific analysis has been produced as is outlined in question 470.24.

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RAI Number: 470.24 Question:

The containment aerosol removal rates are plant design specific and will vary depending on, but not be limited to, the containment geometry, containment size, surface area, steam quality, and containment cooling mechanisms.

Provide assumptions and parameters used in the determination of the SBWR '

containment aerosol removal rates and the computer codes used to calculate the aerosol removal rates.

GE Response:

As stated in the reply to 470.23 the acrosol deposition rate of 0.6 per hour was taken from NUREG-1150 and was calculated for Peach Bottom. Many of the geometric features of the SBWR, such as containment size and surface areas are similar to past BWR's such as Peach Bottom. The SBWR does differ' from past BWR's in containment cooling mechanisms and containment humidity.

These differences result in higher aerosol removal coefficients than past BWRs. Preliminary results from the well benchmarked code NAUAHYGROS yield the following aerosol removal rates:

Time (hr) lambda (1/hr) )

0.9-2.0 1.83 2.0-3.4 0.57 ,

> 3.4 1.36

]

As detailed in Reference 1, a draft copy of which is attached, the low pressure core melt (LPE) with DPVs open and without GDCS injection accident was chosen as the base case for the radiological DBA analysis. This accident along j with other low pressure core melt accidents, some with partial GDCS injection and/or depressurization through W /s only, make up the majority of the SBWR core damage frequency. The case used in the radiological analysis was chosen because it has the shortest time to core uncovery (0.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months ) among the more probable core damage events and because it releases fission products directly into the drywell. Assumptions:

X, Cap release begins at 0.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months after accident initiation as calculated by the MAAP code I NUREG-1465 gap and early in-vessel release fractions (and durations)

I EPRI early in-vessel release fractions for low and non-volatile fission products I Reflood prior to time of vessel failure 1 I Thermal hydraulic conditions computed by MAAP-SBWR

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1) Leaver, D.E., Li,J., Sher, R., "SilWR Containnient Natural Acrosol Removal" February 1994.

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RAI Number: 470.25 Question:

l 470.25 Provide the prirnary containment acrosol removal rates as a function of time following the controlling DilA accident sequence considered in Question 470.18 above, over the entire duration of the accident (30 days).

GE ilesponse: ,

See reply to 470.24.

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RAI Number: 470.26 Question:

SSAR Section 15.6.5.5 refers to a llWR Owners Group report (NEDC-31858P, February 1991) for fission-product holdup, deposition, and resuspension rates used for fission-product mitigation in the SllWR main steam lines and condenser. The report is based on the Ilope Creek Nuc1 car Station design using the TID-14844 source term. For each iodine species !br the entire duration of a DllA (30 days), provide specific values used for fission-product (iodine) mitigation in the SllWR design for the following. ;

(1) iodine deposition rates (2) iodine fixation rates (3) iodine resuspension rates (4) main steam line temperatures (5) total integrated iodine release to the main steam lines (6) total integrated iodine release to the condenser total integrated iodine release from the condenser (7)

GE Response:

The llWR Owners group repor t referred to in the question is a generic report containing a specific application of the ItWR Owners Group radiological methodology to the Hope Creek Nuclear Station. The report does assume that the chemical speciation of the iodines is that appearing in TID-14844 as interpreted in Regulatory Guide 1.3. Even so, the application of the report to the SilWR and source terms as applied from the EPRI URD source terms or NUREG-1465 (draft) is direct and conservative.

(1) the deposition rate is govcined by the equation: -i ano 9 x 10"c " , where 1 R = lloltzman constant = L987 cal M/ mole"K l T = pipe temperature in wK j (2) the fixation rate is goserned hv the equation: l

-n ss 1,351 x 10"c 7 , where T = pipe temperature in ~K (3) the resuspension rate is governed by the equation: j 1.246 x 10"'eT, where .j T = pipe temperature in wK

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(4) The main steam line teml ieratures will he supplied as an appendix to j Chapter 15 in a future amendment. l l

(5) The total integrated release to the main steam lines will be supplied as an j appendix to Chapter.15 in a future amendment.. .

(G) The total integrated release to the condenser will be supplied as an appendix to Chapter 15 in a future amendment.

(7) The total integrated release froni the condenser will he supplied as an appendix to Chapter 15 in a future amendment. The data is available as Table ,

15.6-11 .

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RAI Number: 470.27 ;

Question:

SSAR Section 15.6.5.5 (13 age 15 G15, last ;)aragraj)h) states that s})ccific values -

used and the results of the main steam line leakage analysis are given in SSAR Table 15.48. They are not included in this table. Provide these values and the-results.

GE Resi)onse:

See attached revised SSAR Section 15.6.5.5 which references Table 15.49.

through Table 15.415. The steam line ])arameters are given in item H of Table 15.49 ;tnd the results are given in Tables 15.6-10 through 15.415.'

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25A5113 Rev. A SBWR standard sarery Analysis soport condenser. A secondary path is found along the main steamlines into the turbine though flow through this pathway as described below is a minor fraction of the flow through the drain lines. Consideration of the main steamlines and drain line complex downstream of the reactor building as a mitigative factor in the analysis of LOCA leakage is based upon the following determination.

(1) The main steamlines and dmin lines are high quality lines inspected on a ,

regular schedule.

(2) The main stearrJ nes and drain lines are designed to meet SSE criteria and analyzed to dynamic loading criteria.

(3) The main steamlines and drain lines are enclosed in a shielded corridor which protects them from collateml damage in the event of an SSE. For those portions not enclosed in the steam tunnel complex, an as built inspection is required to verify that no damage could be expected from other components and structures in a SSE.

(4) The main steamlines and dmin lines are required under normal conditions to function to loads at temperature and pressure far exceeding the loads expected from an SSE. This capability inherent in the basic design of these components furnishes a level of toughness and flexibility to assure their sunival under SSE conditions. Alarge data base of experience in the sunival of these types of components under actual earthquake conditions exits which prove this contention. (Reference 15.64) In the case of SBWRfurther margin -

for sunival can be expected since the SBWR lines are designed through dynamic analysis to sunive such events whereas in the case of the actual experience data base, the lines shown to sunive were designed to lesser standards to meet only normally expected loads.

Therefore, based upon the facts above, the main steamlines and drain lines in the SHWR- ,

are used as mitigative components in the analysis ofleakage from the MSIVs. .

The analysis ofleakage from the MSIVs follows the procedures and conditions specified in Reference 15.64. Two flow paths are analyzed for dose contributions. The first pathway through the drain lines is expected to dominate dommated due to the incorporation of a safety related isolation valve on the outboard drain line which will open the line for flow down the drain line under LOCA conditions. The second pathway through the main steamlines into the turbine is expected to carryless than 0.3% of the flow based upon a determination that the maximum leakage past the turbine stop valves with an open drain line would permit only 0.3% flow for the valves to operate within specification. Specific values used and results of the main steamline leakage analysis are given in Tame 4M4 Table 15.6-9 throuch Table 15.6-15.

I Dwease In Reactor Coolant Inventory- Amendment 1 DRAFT 15.6-16 i

1

l RAI Number: 470.28 Question:

Among other things, SSAR Table 15fv9 lists the SilWR condenser parameters used for. fission-product mitigation. Provide the technical bases for the following parameters:

(1) condenser leakage rate of 11 percent per day (2) iodine removal factor of 99.3 percent GE Response:

The technical basis for the condenser leakage rate and iodine removal factor are found in GE Report NEDC-31858 Rev 1,"flWROG Report for increasing MSlV 1.cakage Rate 1.imits and Elimination of 1.cakage Control Systems" (reference 15.fr4). The detailed models are discussed in Appendix C, Section 8.

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I RAI Number: 470.29 .

Question:

State the operability status of the main steam drain valves from the main control room following a DBA. To facilitate the main steam leakage pathway, the '

drain valves should be able to be manually opened from the main control room following a DBA by means of a safety-related power source.

L GE Response: ;

The main steam line drain line air operated valve is supplied with safety-related power off the SBWR battery system and under the 'unlikely circumstance of loss of power fitils open thereby liteilitating the drain line ,

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RAI Number: 470.30 Question:

Among other things, SSAR Table 15.6-9 lists reactor building leakage parameters. Provide the technical bases for the following parameters:

(1) reactor building mixing efliciency of 50 percent (2) reactor building leakage rate of 3600 percent per day GE Response:

(1) The reactor building mixing efficiency of 50 percent is based upon the consideration that an active air system is in inc. This is the normal HVAC system for the restricted areas of the reactor building excluding the safety emelope. such credit is typically accepted based upon the precedence given in Regulatory Guide 1.3. Since the 11VAC is a non-safety related system, the potential for non-operation of the system exists.

In this case, based upon preliminary studies performed by KEMA on stagnant air mixing for small volumes, it was found that 100% mixing occurred for volumes above a given leakage point and that little mixing-occurred below the elevation of the inlet leakage point for the case where the most conservative boundary conditions were considered.11ased upon this and assuming the average penetration at mid point in any room, 50% of the room would experience 100% mixing. In this case, however, the flow from the room through the reactor building (considerieg that most of the restricted areas are below grade) would be less than the leakage rate of 3600 per cent assumed and therefore would result in an overall lower leakage to the environment than that actually analyzed.

(2) The reactor building HVAC turnover rate for the control (non-clean) areas is one and one-half air exchanges per hour. This translates into a turn over rate of 36 air exchanges per day or 3600% per day. During a LOCA event and assmning a loss of power, the IIVAC senicing these areas would not operate, thereby removing all motive force for releases-from these areas considering these areas are primarily below grade.

Therefore, it was considered more conservative that the IIVAC be.

assumed operating during the LOCA to provide such a motive force for releases to the environment.

RAI Number: '470.31 Question:

SSAR Section 15.6.6 describes the feedwater line break accident. (outside containment) and SSAR Table 15.6-17 lists the major parameters used in its ,

radiological consequence assessment. Provide the technical basis for the following parameters used: >

(1) 320,000 kg of condensate released from the break (2) 10,000 kg of condensate flashed to steam from the break (3) two percent carryover factor ofiodines in the condensate to flashed steam GE Response:

(1) The value of 320,000 kg of condensate released from the break is the sum of the water inventories in (1) four low pressure feedwater heaters, (2) two high pressure feedwater heaters, (3) condensate piping, and (4) the condensate ,

hotwell. This is a s,gnificant conservative over calculation of available water inventory since low water level in the condensate hotwell will trip the condensate pump cutting off the supply of water to the break.

(2) The 10,000 kg of condensate flashed to steam from the break is an .

assessment based upon ' applying the energy contained in each of the ;

components in item (1) above to the equation:

i M' = M' (h, - hf ) , where ;

(h,-h,)

hi s= hlass flow rate of fluid flashed to steam ,

hi t= hf ass flow rate of fluid "

h t= enthalpy of the released fluid

.h t= 180.1 Btu /lb at one atmosphere h3= ll50A Bru/lb at one atmosphere (3) The two percent carryover is based upon water to steam carryover fractions seen in Im'R steam separators and is indicative of this physical- .

interface (APED-5756. SSAR Reference 15.6-8). ]

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25A5113 Rsv. A SBWR StandaniSafetyAnalysis Report Table 15.417. A schematic diagram of the leakage path for this accident is shown in Figure 15.45.

Fission Product Release -There is no fuel damage as a consequence of this accident.

In addition, an insignificant quantity of activity (compared to that existing in the main condenser hotwell prior to the occurrence of the break) is released from the contained piping system prior to isolation closure.

The i(> dine concentration assumed is that of the maximum equilibrium reactor water concentration given in Subsection 15.6.4.5, case 2, subject to a 2% carryover ofiodine in the water to steam condensate (Reference 15.48). Noble gas activity in the condensate is negligible since the air ejecton remove all noble gases from the condenser, nssion Product Transport to the Environment -The transport pathway consists of liquid release from the break, carrymer to the turbine building atmosphere due to flashing and partitioning and unfiltered release to the environment through the turbine building ventilation system.

Of the 320,000 kg (705,500 lb) of condensate release from the break,10,000 kg (22,000 lb) flashes to steam.

Taking no credit for holdup, decay or platemut during transport through the turbine building, the release of activity to the environment is presented in Table 15.418.The release is assumed to take place within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the occurrence of the break.

Results-The calculated exposures for the analysis are presented in Table 15.419 and are a small fraction of 10CFR100 guidelines.

15.6.7 COL License Information None.

15.6.8 Fleferences 15.41 F.J. Moody, Maximum Two-Phase Vessel Blowdown from Pipes, ASME Paper Number 65-WA/IIT-1, March 15,1965.

15.42 II.A. Careway, V.D. Nguyen, and P.P. Stancavage, Radiological Accident Evaluation - the CONACO3 Code, December 1981 (NEDO.21143-1).

15.43 II.A. Careway, V.D. Nguyen, and D. G. Weiss, Control Room Accident Exposure Evaluation, CRDOS Program, February,1981 (NEDO-23909A) 15.44 L.S. Ire, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination ofleakage ControlSystems, February,1991 NEDC-31858P.

Decreas in Reactor Coolant Inventory- Amendment 1 DRAFT 15.6-21 2N7/94

I RAI Number: 470.32 Question: .

I Provide the postulated primary coolant leakages from ESF components (valve stems and pump seals) that are located outside of the primary containment to the secondary containment (safety envelope) and to the reactor building-following a DilA.

GE Response: i There are no ESF components required post LOCA to recirculate containment water outside the containment. Therefore, no leakages are assumed from these sources following a DilA. ;

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RAI Number: 470.33 >

Qtiestion: ,

Provide a radiological consequence assessment Ihr the SilWR reactor water cleanup pipe break accident (outside the primary containment). For this assessment, GE may assume that the break is instantaneous, circumferential, 1 and on the downstream side of the outmost containment isolation valve, but on the upstream side of the reactor water cleanup demineralizers.

GE Response:

The reactor water clean up is not a bounding accident fbr the SilWR. An assessment to prove this will be provided the NRC. The SBWR _ clean up water system uses 3 inch lines similar to past llWRs as compared to the AllWR which uses 8 inch lines which causes this accident to be bounding on the A Il W R. .

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RAI Number: 470.34 Question:

Provide a radiological consequence assessment for the ofTgas system failure accident. GE stated in SSAR Section 15.7.1 that the NRC has deleted this SRP section and, therefore, did not provide this analysis. The NRC has transferred (but not deleted) SRP Section 15.7.1, " Waste Gas System Failure," to SRP Section 11.3, " Gaseous Waste Management System Failure," as llranch Technical Position 11-5, " Postulated Radioactive Releases due to a Waste Gas System Leak or Failure," so the assessment must still be provided.

In its assessment, GE should assume an inadvertent bypass of all charcoal beds due to an operator error or system computer error,in addition to the failure of the automated air-operated downstream isolation valve. _GE should also assume that during this accident,. the plant is operating at and continues to operate at the maximum permissible offgas release rate (measured at ofTgas recombiner effluent) as specified in the SilWR technical specifications.

GE Response:

This assessment will be provided in a future amendment consistent with that provided for the AllWR.

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L RAI Number: 470.35 ,

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Question: 1 i

SSAR Section 15.7.4 describes fuel handling accidents, and SSAR Table 15,7-1 :

lists the major parameters assumed in the radiological consequence assessment for the fuel handling accidents. Provide the technical bases for the following parameters used:

(1) 62 hours7.175926e-4 days 0.0172 hours 1.025132e-4 weeks 2.3591e-5 months of minimum decay time prior to fuel movement (2) reactor building release rate of 500 percent per hour Also state the fuel pool decontamination liictors assumed for the fission-products other than noble gases and iodines.

GE Response:

(1) The 62 hours7.175926e-4 days 0.0172 hours 1.025132e-4 weeks 2.3591e-5 months of mininmm decay time prior to fuel movement is based upon an assessment of mandatmy actions required to be performed for a SilWR Refueling.

(2) The reactor building release rate of 500 percent per hour is an assmned release rate to insure that 95% plus of the released radionuclides are purged to ,

the environmen, within two hours. The actual building vent rate is 150% per hour.

(3) No other radionuclides other than iodines and nobles gases are considered in this accident. Noble gases by their nature will evolve from the fuel plenum into the water and finally to the air space from a broken fuel rod.

lodines as an acumed volatile species (Regulatory Guide 1.25) are expected to also evolve from the water to air subject to water partitioning. No other fission products which are radiologically significant and volatile are found in any ,

quantity in the post-shut down plenum gap. (also see reply to question 470.12) l I

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RAI Number: 470.36 Question:

SSAR Section 15.7.5 describes the spent fuel cask drop accident, and SSAR Table 15.7-8 lists the major parameters assumed in the radiological consequence assessment for the accident. Provide the technical bases for the following parameters used: )

-)

cask drop distance of 24 meters )

(1)

(2) 120 days of minimum decay time prior to spent fuel cask movement (3) 2.5 air exchange per hour in the reactor building (4) the maximum capacity (number of spent fuel rods) of spent fuel cask (5) fuci pool decontamination factors assumed for the airborne fission-products GE Response:

(1) The cask drop distance is the maximum drop distance from the ;

refueling floor down to the loading dock in the reactor buihling, j (2) The time to ship udue of 120 days was originally based upon an assessment for spent fuel rods such that 120 days of cool down was needed to assure sufficient time to allow for heat dissipation and rod integrity under cask loaded conditions. The value is also somewhat overly conservative since we know of no spent fuel storage site which will accept spent fuel out of core less than one year. j (3) The 2.5 air exchange per hour is an assumed exchange rate to purge the area of radionuclides and is an over assessment of the ilVAC capability.

(4) The maximum capacity of the spent fuel cask is 18 bundles (Table 15.7-8, item C) for a total of 1080 fuel rods (Table 15.7-8, item D). This is based upon using a VECTRA IF300 shipping cask, certificate number CFC-9001.

-(5) This accident does not assume a drop into the fuel but onto a concrete.

loading dock. As stated in the reply to question 470.37, the crane pathway for spent fuel casks also does not approach any sensitive areas containing new or spent fuel.

RAI Number: 47037 Question:

Proside the radiological consequence assessment for a heavy load drop accident dropping a heavy object onto the fuel in the reactor vessel during fueling and refueling operations. In the assessment, GE should evaluate (1) whether the reactor building cranc system design meets single-failure-proof criteria, and (2) the provisions prosided Ihr prevention ofload unbalancing which could potentially defeat the single-failure-proof criteria. State whether the SilWR reactor building crane follows the guidance provided in NUREG-0554, " Single.

Failure-Proof Cranes for Nuclear Power Plants." for design,' fabrication, installation, and testing.

GE Response:

The SBWR bounding radiological consequem e assessment for a heavy load drop accident onto the fuel in the reactor vess 1 is defined as a fuel assembly dropped from the maxinnun height of the fuel lift over the reactor The radiological consequence assessment of this accident is provided in SSAR Section 15.7A. The calculated exposmes for the heavy load dro:) accident are provided in SSAR Table 15.7-7 and are within the guidelines oE 10CFR100.

2 The reactor building cranc system is not designed to meet the single-failure-proof criteria provided in NUREG-0554. .

The avoidance of the potential fbr safett or radiological consequences of a reactor building crane heavy load drop accident is achieved by the arrangement of the reactor building and by design featmes, in accordance with j NUREG-0612, " Safely Design and 1.oad Path Control of IIcavv 1.oads at Nuclear -!'

Power Plants" e

There are ric heavy loads lifted by the reactor building crane that, if dropped in' :

any location along its prescribed path, would cause the failure of any safety' ;)

system, change the configuration of fuel storage racks or damage new or spent lhel. It is acknowledged that heavy loads are moved over the reactor pressure -

vessel during refueling operations. Examples are; the drywell head, reactor .

pressure vessel insulat' ion, reactor pressure vessel head and steam dryers and _ ,

separators, llowever, these components are too large to reach the fuelif dropped. !

An example of accident mitigation by arrangement is, the spent fuel cask is not moved over systems vital for the safe operation of the plant or along a path that, if dropped, could reach new or spent fuel. An exyunple of accident mitigation by. .

design feature is that the small gates between pools are on hinges that do not require the reactor building crane for movement. These small gates are removed periodically for servicing to replace the gate seals. Since this work is -

done when the RPV head is on, there is no potential for damaging the fuel -

inside the core. The one large gate to the equipment pool that is lifted by the reactor building crane is not normally moved during refueling and b is too large to enter the reactor pressure vessel.

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+ <1 The refueling machine is the only crane that is permitted to move fuel, or any other heavy load over the reactor' vessel during fueling and refueling operatio'ns that if dropped could reach the fuel in the reactor pressure vessel. The hoist, mast and grapple design meet the single-failure-proof criteria of NUREG- 0554, from tne double braking system of the cable drum located on the refueling

. mac hine, through the dual cables to the grapple and through dual grapple attachments to the fuel bale. There are }3rovisions provided for prevennon ofload unbalancing. For example, the design mcludes a load balanced duplicate load :

path for the mast gimbals that attach directly to the refueling machine main structure. The first single load path from the main structure of the refueling machine to the fuel is the fuel bale. For this reason it is assumed in the analyzed accident in Section 15.7.4, that the fuel bale fails, and a spent fuel assembly is dropped from its maximum height to the fuel in the core. i i

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25AS113 Rev. A j SBWR standard satory Analysis Report 1

15.6-5 Ramsdell,J.V., Alternatives to Current Procedures Used to Estimate Concentrations in Building Wakes,21st DOE /NRC Nuclear Air Cleaning Conference, pgs 714-729.

15.6-6 "Re=ed Acciden: Scurce Ter= for Light 4 Vater Nuclear Pc=- P!ar.::",

SFAN-9242,7rApri! 10,1992.

Advanced Licht Water Reactor Utility Reauirements Document. Volume III.

Electric Power Research Institute.

15.6-7 C.C. Lin, Anticipated Chemical Behavior ofIodine under LOCA Conditions, January,1981, NEDO-25370, 1f),6-8 N. R. Horton. W. A. Williams.J. W. Holtzclaw. Anahtical Methods for Evaluatine the Radiolocical Asoccts of the General Electric Boiling Water Reactor. March 1969. APED-5756.

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15.6-22 Decrease In Reactor Coolant Inventory - Amendment 1 DRAFT 2i2794

- . -. -. . . .. = . . -. .-

' 25A6113 Rev. A

.SBWR standars sahrey Analysis seport the level of the refueling floor to ground level through the refueling floor maintenance hatch.

i

' 15.7.5.2 Radiological Analysis The largest size of BWR fuel cask is conservatively assumed to be dropped ,

approximately 24 meters from the refueling floor level to ground level on transport "

from the decontamination pit out of the reactor building. It is conservatively assumed that all fuel rods are damaged and the fission gases in the fuel rod gap space are released !

1 to the reactor building and then to the environment over a two hour period.

Table 15.7-8 provides the assumptions for this analysis andTable 15.7-9 the radiological :

consequences. As can be seen from Table 15.7-9, the radiological releases are well ,

within guidelines.

15.7.6 COL License Information None.

15.7.7 References 15.7-1 D. Nguyen, et.al., Radiological Accident Evaluation -The CONAC03 Code, December 1981 (NEDO-21143-1).

15.7-2 N.R. Ilorton, W.A. Williams, and LW,J.,yL Holtzclaw, Analytcal Methods for Evaluating the Radiological Aspects of Geneml Electric Boiling Water '

Reactors March 49M 1959 (APED-5756).

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15.7 4 Radioactive Releene - Amendment t DRAFT M,94

RAI Number

470.12

- Question:

470.12 .In the response to Part G) of Question 470.16 above regarding iodine spiking, state the chemical forms ofiodine assumed to spike and state the

. reasons for not considering spiking of other nuclides such as cesium. ;

GE Response:

It is assumed that the above questions refers to question 470.11 and not the 470.16 stated. The chemical form assumed is 12 which is typically found in fuel rod :

gaps under normal conditions. Cesium is not typically fbund in spiking activity or in the fuel rod gap since the release of cesium from the fuel rod matrix requires temperatures and conditions far above normal operating . ,

conditions or those experienced during a depressurization event.

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i' RAI Number: 470.13

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j. -l Question: i p- 1 l- In the titles of SSAR Tables 15.61 through 15.6-3, change " Instrument" to o

"Small," so that the titles will read "Small 1.me 15reak Accident l'arameters."

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GE Response

l See attached revised SSAR Tables 15.41 through 15.43.

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25A5113 R2v. A SBWR senmtant so Autrsis soport Table 15.6-1 lastrument Small Line Break Accident Parameters i Data and assumptions used to estimate source terms A. Power level 2040 MWt B. Mass of fluid released 13,000 Kg (28,600 lbm)

C. Mass of Fluid flashed to steam 5,000 222Q Kg (44A00 5.QQQ lbm)

D. Mass of fluid in reactor 248,600 Kg (544,000 lbm)

E. Duration of accident 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months F. Number of bundles in core 732 G. Number of rods per bundle 60 H. Time from SCRAM to Iodine Spike 0.0 sec 11 Data and assumptions used to estimate activity released A. todine water concentration Subsection 15.6.4.5, case 2 B. lodine Spiking 1131 7.8E+4 MBq/ bundle (2.1 Ci/ bundle) 1-132 1.2E+5 (3.2) 1-133 1.9E+5 (5.1) 1-134 2.0E+5 (5.4) 1-135 1.8E+5 (4.8)

C. lodine plateout fraction 50 % !

D. Reactor Building Flow rate 200%/ hour E. SGTS Filter Efficisney none assumed l F. Time from SCRAM to lodine Spike 0.0 sec l l

1 lil Dispersion and Dose Data f I

A. Meteorology Table 15.6 3 l B. Site Boundary distances Table 15.6-3 C. Method of Dose Calculation Reference 15.6-2 D. Dose conversion Assumptions RG 1.109,and Fed Guide Report 11 E. Activity Inventory / releases Table 15.6-2 F. Dose Evaluations Table 15.6-3 Decrease in Reactor Coolant inwntory- Arnendenent 1 DRAFT 15.6-23 225AM

I 2SA5113 Rsv. A SBWR standard sanrty Anarrsis aeport Table 15.6-2 ILBA SLBA isotropic inventory ISOTOPE 30-sec 10-min 30-min 1-hour 2-hour 4-hour 6-hour 8-hour A. REACTOR BUILDING INVENTORY IN MBq 1-131 22E+1 2.5E+1 1.2 E+3 1.2E+3 12E+3 7.5E+2 0. O.

1-132 2.1E+2 2.4E+2 1.8E+3 1.8E+3 1.8E+3 1.1 E+3 0. O.

1-133 1.5E+2 1.7E+2 2.8E+3 2.8E+3 2.8E+3 1.8E+3 0. O. ,

1-134 4.2E+2 4.8E+2 3.0E+3 3.0E+3 3.0E+3 1.9E+3 0. O.

1135 2.2E+2 2.5E+2 2.7EA3 2.7E+3 2.7E+3 1.7E+ 3 0. O.

TOTAL 1.0E+3 12E+3 12E+4 1.2E+4 1.2E+4 7.3E+3 0. O.

B. REACTOR BUILDING INVENT 0FiY IN CUElES l-131 5.9E-4 6.8E-4 3.2E-2 32E-2 3.2E-2 2.0E-2 0. O.

1-132 5.8E-3 6.6E-3 4.9E-2 4.9 E-2 4.9E-2 3.1 E-2 0. O.

1-133 4.1 E-3 4.6E-3 7.6E-2 7.6E-2 7.6E-2 4.8E-2 0. O.

1-134 1.1 E-2 1.3E-2 8.2E-2 8.2E-2 8.25-7 5.2E-2 0. O.

1-135 5.9 E-3 6.8E-3 7.3E-2 7.3E-2 7.3E-2 4.6E-2 0. O.

TOTAL 2.8E-2 3.2 E-2 3.1 E-1 3.1 E-1 3.1 E-1 2.0E-1 0. O.

C. ISOTOPIC RELEASE TO ENVIRONMENT in MBq l-131 3.0E+1 1.0E +3 9.8E+4 2.5E+5 5.4E+5 9.66+5 1.1E+6 1.1E+6 l-132 2.9E+2 9.9E+3 1.6E+5 3.8E+5 8.3E+5 1.5E+6 1.7E+ 6 1.7E+6 l-133 2.1 E+2 7.0E+3 2.4E+5 5.9E+5 1.3E+6 2.3E+6 2.6E+6 2.6E+6 l-134 5.8E+2 1.9Ed 2.N+5 6.5E+5 1.4E+6 2.5E+6 2.8E+6 2.8E+6 l-135 3.0E+2 1.0E+4 2.3E+5 5.7E+5 1.2E+6 22E+6 2.5E+6 2.5E+6 TOTAL 1.4E+3 4.8E+4 1.0E+6 2.4E+6 5.3E+6 9.4E+6 1.1E+7 1.1E+7 D. ISOTOPIC RELEASE TO ENVIRONMENT IN Ci 1-131 8.2E-4 2.8E-2 2.6E+0 6.6E+0 1.5E+1 2.6E+1 3.0E+1 3.0E+1 1-132 7.9E-3 2.7E-1 4.3E+0 1.0E+1 2.3E+1 4.0E+1 4.6E+1 4.6E+1 1-133 5.6E-3 1.9 E-1 6.5E+0 1.6E+1 3.5E+1 6.2E+1 7.1E+1 7.1 E+1 1-134 1.6E-2 5.3E-1 7.3E+0 1.7E+1 3.8E+1 6.7E+1 7.7E+1 7.7E+1 16.6-24 Decrease in Reactor Coolant Innntory- Amendment 1 DRAFT 2/2M4

25AS113 Rsv. A 'l SBWR samrantsannyAnarr sisneron Table 15.6-2 ESA SLBA Isotropic inventory (Continued)

ISOTOPE sec 10-min 30-min 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 2-hour 4-hour 6-hour 8-hour i J

1-135 8.2E-3 2.8E-1 6.3E+0 1.5E+1 3.4E+1 6.0E+1 6.8E+1 6.8E+1 ;

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TOTAL 3.8E-2 1.3E+0 2.7E+1 6.6E+1 1.4E+2 2.5E+2 2.9E+2 2.9E+2 ;

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-l' Decrease in Reactor Coolant Innntory- Amendment 1 DRAFT 15.6-25 2/25&nd a M= ~ w ~ M

25A5113 Rav. A SBWR StandaniSafety Analysis Report Table 15.6-3 ILBA SLBA Results - Meteorology and Dose Results Meteorology Distance Thyroid Dose Whole Body Dose (sec/m3) (m) (Sv) (Sv) 1.8E-3 max' 3.0E-1 2.1 E-3 1.0E-3 800 1.6E-1 1.2 E-4

" max" = maximum meteorology to meet 10% of 10CFR100 limits.

Table 15.6-4 Sequence of Events for Steam Line Break Outside Containment Time (sec) Event 0 Guillotine break of one main steam line outside containment.

0.5 High steamline flow signalinitiates closure of MSIVs

< 1.0 Reactor begins scram.

<2.0 Partial closure (15%) of MSIV initiates isolation condensers.

<5 MSIVs fully closed.

10 Reactor low water Level 2 is reached. Isolation condensers receive second initiation signal.

32 lsof ation condensors in full operation. Water level stabilized.

435 SRVs open on high vessel pressure (if isolation condensers are not available). The SRVs open and close to maintain vessel absolute pressure at approximately 8.5 MPa (1215 psia).

3540 Reactor low water Level 1 is reached (if isolation condensors are not available).

ADS timer initiated.

3550 ADS timer timed out. ADS actuation sequence initiated. GDCS timer initiated.

3700 GDCS timer timed out. GDCS injection valves open.

3880 Vessel pressure decreases below shutoff head of GDCS. GDCS reflooding flow into the vessel begins.

The core remains covered throughout the transient and no core heatup occurs.

15.6-26 Decrease in Reactor Coolant Innntory- Amendment 1 DRAFT 22 594

RAI Number: 470.14 Question:

SSAR Section 15.6.4 describes a main steam pipe break accident outside containment and SSAR Table 15.G-5 lists the major assumptions and parameters used in the radiological consequence assessment of this accident. Provide the technical bases, complete with applicable references, for the following parameters used:

(1) air exchange rate of 6000 per day in the steam tunnel (provide the free air volume of the steam tunnel and break location)

(2) 12,000 kg of steam mass released (3) 2,400 kg of water mass released (4) iodine concentration in the reactor coolant based on ofTgas release rates of 0.2 Ci/sec and 0.05 Ci/sec.

GE Response:

The technical basis for the main steamline break accident is found in GE Report NEDO-21143-1 (reference 15.6-2), which discusscs the basis for and algorithms used in evaluating the accident. The SSAR is the technical documentation for the evaluation.

(1) The air exchange rate of 6000 is an assumption used w describe rapidly venting fission products from the steam tunnel. It is not considered important as to where the break is in the steam tunnel since the tunnel vents to the lower floors of the turbine building and then through grates into the upper volume of the turbine building. Since the volume of steam is subject to extensive condensation along the release pathway, representing the release by a large turn over rate is reasonable. The steam tunnel volume in the reactor building is-approximately 950 m3 (2/3) The masses of steam and water released are calculated by GE transient analysis codes for the specific case c,f the SI1WR MSL break with'MSIVs closing within 5 seconds plus the steam line volume at the time of the accident.

(4) The iodine concentration of 0.2 mci /gm is the standard 13WR technical specification (WASH-123) for normal operations. This value is referred to as Case 2. The maximum technical specification of 4.0 mci /gm is referred to as Case 1 and assumes significant fuel failures with iodine spiking. The noble gas offgas release rate of 0.05 Ci/second is a conservative estimate of that provided in NUREG-0016 and ANS Standard 18-1. The maximum noble gas limit of 0.2 Ci/sec is derived from SRP 11-2 as equivalent to 100 mci /sec/MWt for a 2,000 MWt plant.

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RAI Number: '470.15 Question:

SSAR Section 15.6.5.5 states that except where noted, the reactor accident source -

terms used are consistent with those specified in the EPRI Passive Requirements Document. Itemize and list in a table the exceptions taken.

GE Response:

There are no exceptions in the February 1993 submittal.' See attached revised Table 15.69.

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References:

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