ML20134D893

AP600 Radioactive Waste Mgt Suppl Info

ML20134D893
Person / Time
Site:	05200003
Issue date:	09/30/1996
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20134D892	List: ML20134D893 NSD-NRC-96-4858, Forwards Suppl Info on AP600 Radioactive Waste Mgt Sys Design
References
NUDOCS 9610310083
Download: ML20134D893 (45)

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AP600 Radioactive Waste Management Supplemental Information l

September 19%

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l Westinghouse Electric Corporation Energy System Business Unit P.O. Box 355 Pittsburgh, PA 15230-0355 019% Westinghouse Electric Corporation All Rights Reserved

=^ tmw wnf:1tMB3096 9610310083 961023 PDR A

ADOCK 05200003 l PDR

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All TABLE OF CONTENTS t

t i 1 INTRODUCTION . . . . .. .................................. ....... 1-1 Question 460.4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.4(R2)-1 Question 460.5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.5(R2)-1 ,

Question 460.7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.7(R2)-1 l Question 460.10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.10(R2)-1  ;

Question 460.11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.11 (R2)-1 l Question 460.12 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.12(R1)-1 Question 460.13 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.13(RI)-1 l Question 460.15 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.15(R3)-1

' Question 460.16 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.16(R1)-1 i Question 460.18 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.18(R1 )-1 Question 460.20 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.20(R1)-1 .l' Question 460.21 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.21(RI)-1 l Question 460.25 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.25(R1)-1 l

- Question 460.26 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460.26-1 l

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iv LIST OF TABLES Table 1-1 RAI Responses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -2 Table 1-2 Cross-Reference to Additional Information in AP600 SSAR . . . . . . . . . . 1-3.

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1 AP600 Radioactive Waste Management Supplemental Information Revision 0, September 1996 m:\3252w.wpf:1b4N3096 i

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l 1 INTRODUCTION Radioactive Waste Management for the AP600 passive plant design is discussed in Chapter 11 of the AP600 Standard Safety Analysis Report (SSAR). Requests for Additional Information (RAI) were provided to Westinghouse as part of the Nuclear Regulatory Commission (NRC) review of the AP600 Radioactive Waste Management systems and  ;

processes. Westinghouse responses to these RAIs were ut!hred in the preparation of the NRC's November 1994 Draft Safety Evaluation Report (DSER). The purpose of this report is

! to consolidate, into a single reference, those RAI responses utilized by NRC staff as a basis l for acceptance as documented in the November 1994 DSER. The RAI responses included in this report are identified in Table 1-1. A cross-reference to additional information in the AP600 SSAR is provided in Table 1-2.

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Table 1-1 RAI Responses 460.4 Rev 2 460.5 Rev 2 460.7 Rev 2 l 460.10 Rev 2 460.11 Rev 2 460.12 Rev1 460.13 Rev1 460.15 Rev 3 460.16 Rev 1 460.18 Rev1 460.20 Rev1 460.21 Rev1 460.25 Rev1 460.26 Rev1 l

I AP600 Radioactive Waste Management Supplemental Information Revision 0, September 1996 m:\3252w.wpf;1b-1001%

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Table 1-2 Cross Reference to Additional Information in AP600 SSAR 460.4R2 t.) COL responsibility per subsection 11.5.7 b) Site boundary dispersion factor in Table 15A-5, nuclide dose conversion factors in Table 15-A-4, source terms in Table 11.1-2, gaseous releases in Tables 11.3-3 and 11.3-4 c) Prevention of hydrogen ignition in 11.3.1.2.3.1 ,

d) See SSAR subsections 3.6.1, 9.4.3.1.2 and 9.4.7 '

e) See SSAR subsections 3.6.1, 9.4.3.2.1.2, 9.4.3.2.3.2, and 9.4.7 460.5R2 Summary of solid waste volumes in SSAR subsection 11.4.2.1 460.7R2 See SSAR subsection 11.5.2.3.1 and Table 11.S-1 460.10R2 a) SSAR Tables 11.2-6 and 11.3-1 are GALE code inputs and design parameters, respectively, b) Basis for question deleted faom SSAR Table 11.2-6. Table 11.3-3 revised per updated GALE code analysis c) See SSAR Table 11.3-2 1

460.11R2 a) See SSAR subsections 11.4.2.3.1 and 11.4.2.4.1 l b) See SSAR subsections 11.4.2.3.2 and 11.4.2.3.3,11.5.2.3.2 I c) See SSAR subsection 11.4.2.4.1 d) See SSAR subsection 11.4.2.1 and Table 11.4-1 e) EPRI requirement not applicable to AP600 I f) Storage space duration description added to SSAR subsection 11.4.2.1 460.12R1 See SSAR subsection 11.5.1.2 460,13R1 See SSAR subsections 9.4.3 and 9.4.7 460.15R3 See SSAR Table 9.3.3-1 460.16R1 See SSAR subsections 11.5.5 460.18R1 11.2a) See SSAR subsection 11.2.4.2 11.2b) See SSAR subsection 11.5.7 11.2c) See SSAR subsection 11.2.3.6 and App.1A 11.3a) See SSAR subsection 11.3.4.1 11.3b) See SSAR subsection 11.5.7 11.3c) See SSAR subsection 11.3.3.6 and App.1A 11.4a) See SSAR subsection 11.4.6 11.4b) See SSAR subsection 11.4.6 l 11.4c) See SSAR subsection 11.4.2.1 11.4d) See SSAR subsection 11.4.1.3 11.4e) See SSAR subsection 11.4.5 11.5a) See SSAR subsections 11.5 and 11.5.7 11.5b) See SSAR subsections 9.3.3.1.2.2,11.5.2.3.1 and 11.5.2.3.3 l

AP600 Radioactive Waste Management Supplemental Information Revision 0, September 1996 m:\3252w.wpf:1b.1001%

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1-4 Table 1-2 Cross Reference to Additional Information in AP600 SSAR (Cont'd) I 460.20R1 See SSAR Appendix 1A 460.21R1 See SSAR Tables 11.2-9 and 11.3-4 460.25R1 See SSAR Tables.11.2-8,11.2-9,11.3-3 and 11.3-4 l l

460.26R1 See SSAR subsection 11.1.1.4 )

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J AP600 Radioactive Waste Management Supplemental Information Revision 0, September 19 m m:\3252w.wpf:1t>.1001%

l NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Question 460.4 Provide the following information regarding gaseous radwaste management systems (Section 11.3):

a. A description of release points for airborne effluents (plant vent and turbine building vent). The description l should include information on the height of the release point above grade, its height above and relative location I to adjacent structures, expected temperature of the gaseous effluents, flow rate and size and shape of the flow orifice (note that these parameters are required in conjunction with plant-specific parameters to determine plant- l specific atmospheric dispersion factors).

b. A demonstration of compliance with Branch Technical Position (BTP) ETSB 11.5 " Postulated Radioactive Releases due to a Waste Gas System L.cak or Failure."

c. A discussion of compliance with GDC 3 as it relates to providing protection to gaseous waste handling and treatment systems from the effects of an explosive mixture of hydrogen and oxygen. The discussion should include the provisions' incorporated in the AP600 design to control releases due to hydrogen explosions in the gaseous waste management system. Additionally, it should include the type, number and locations of gas analyzers provided in the design of the gaseous waste management system (for response guidance, see j Acceptance Criterion II.B.6 of Section 11.3 of the SRP). I

d. A discussion of compliance with GDC 60 as it relates to control of releases of radioactive materials to the environment. The discussion should refer to Regulatory Guide (RG) 1.140 to be consistent with Acceptance Criterion II.6.a of Section 11.3 of the SRP (note that reference to the subject guide in Section 9.4 of the SSAR alone is not sufficient). As a minimum, provide a cross reference with Section 9.4 and state clearly whether the design complies with the guide or not).

e. A discussion of compliance with GDC 61 as it relates to radioactivity control in gaseous waste management systems and ventilation systems associated with fuel storage and handling areas.

Response

a. The plant vent is located along the south-east wall of the containment shield building, as shown in SSAR Figure l 1.2-12. SSAR Figure 1.2-12 also shows the adjacent structures to the plant vent, including the roof of the shield building at elevation 307'-6", the roof of the fuel handling building at elevation 180'-0", and the auxiliary l building at elevation 160'-6". The plant vent has a rectangular cross section approximately 7 feet square. The top of the plant vent is approximately at elevation 250'-0" and is about 150 feet above the grade elevation of l 100'-0", which is shown in SSAR Figure 1.2-13.

A summary follows of the systems, airflow rates, and the normal operating temperature range of the connections to the plant vent of systems which handle radioactive gas or air. The HVAC exhaust airflow rates were l compiled from data provided in SSAR Tabbs 9.4.3-1 and 9.4.7-1. The temperature ranges of the exhaust air l from various plant rooms are provided in SSAR Section 9.4.

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NRC REQUEST FOR ADDITIONAL INFORMATION

$. j j Response Revision 2 Nominal Design Maximum Airflow Rate Temperature System (sefm) Range (*F)'

1. Radiologically Cor.irolled Area Ventilation System 84,0002 50-130

2. Containment Air Filtration System 4,000/8,000'70-122 d

3. Solid Radwaste Building Ventilation System 20,000 50-130

4. Health Physics and Hot Machine Shop HVAC System 12,000 65-85 5

5. Gaseous Radwaste System 0.5 --

The turbine vent for the combined discharge of the condenser air removal system and the gland seal system is located on top of the turbine building and terminates at about elevation 235'-0",8 to 10 feet above the roof.

The exhaust is approximately 18 inches in diameter, with either an inverted "J" discharge oriented vertically downward or a cut at 45 degrees oriented vertically upward. The gaseous effluents discharged will vary with '

the operation of the plant and ambient conditions, but they are expected to have the following approximate ,

properties:

Condenser air removal system:

Flow 40 cfm Temperature 100 F Gland seal system j Flow 1200 cfm '

Temperature 145F 2This temperature range is conservative; multiple compartments at different temperatures will tend to reduce the range.

2 Assumes a maxirrum airtlow rate from the fuel handling area associated with refueling operations; normal operation will have a lower flow.

3 Operates intermittently. One or two redundant systems may be operated.

l ' Includes 20 F estimated temperature rise due to thmugh e*haust fan motor heat.

i l ' Intermittent operation.

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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 l Note that any of this information may be changed in the site specific Offsite Dose Calculation Manual (ODCM) l which will be submitted by the Combined License Applicant.

b. As stated in SSAR Subsection 15.7.1, the SRP no longer includes this event as part of the review process and no analysis is provided in the SSAR.

An analysis has been performed in keeping with the guidance of Branch Technical Position ETSB 11-5. The analytical assumptions are as follows:

1. The accident consists of a failure of the gaseous waste management system in such a way that the charcoal delay beds are bypassed and activity is released directly to the environment. The duration of this release is one hour.

2. The site boundary atmospheric dispersion factor is as given in Table 15A-5 (1.0E-3 sec/m').

3. The nuclide dose conversion factors are as given in Table 15A-4.

4. The primary coolant source term is that associated with operation at the design basis fuel defect level of 0.25 percent. This source term is provided in Table 11.1-2. Consistent with ETSB 11-5, the beyond-design-basis case of one percent fuel defects is also considered. With one percent fuel defects, the source term is four times that associated with operation with 0.25 percent fuel defects.

5. The full letdown flow of 100 gpm is being processed by the gas stripper, and the stripped gases are released.

6. There is a 30 minute decay period to address the time for the release to reach the exclusion boundary.

7. The normal release of noble gas activity from the gaseous radwas'e system is added to the calculated accident releases.

The resulting activity releases follow for the case assuming the design fuel defect level of 0.25 percent:

Isotope Release (Ci)

Kr-85m 1.2El Kr-85 4.5El Kr-87 5.5E0 Kr-88 1.9El Xe-131m 1.2El Xe-133m 1.lE2 Xe-133 1.8E3 i

Xe-135m 5.0E-1

, Xe-135 5.0El Xe-138 7.9E-1 W westinghouse 4mm2N l

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l NRC REQUEST FOR ADDITIONAL. INFORMATION Response Revision 2 For the case assuming 0.25 percent fuel defects (design basis fuel defect level), the whole body dose at the site boundary is 0.0224 rem which is below the 0.5 rem dose limit identified in ETSB 11-5.

For the case assuming the beyond-design-basis case of 1.0 percent fuel defects, the above releases would be increased by a factor of four and the whole body dose at the site boundary is 0.09 rem which is below the 0.5 rem dose limit identified in ETSB 11-5.

c. The AP600 gaseous radwaste system (WGS) normally contains a mixture of hydrogen and nitrogen gases, bearing small quantities of radioactive gases. Depending on the plant operation, the mixture will be mostly nitrogen during purging and mostly hydrogen when degassing reactor coolant. The WGS is designed to prevent an explosive mixture of hydrogen and oxygen.

Provisions made to prevent a flammable or explosive mixture, as discussed in SS AR section i 1.3.1.2.3.1, include the following:

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- The WGS operates at a pressure slightly above atmospheric, which prevents the ingress of oxygen, which l could occur in a sub-atmospheric system. l

- A continuous purge flow of nitrogen is provided at the outlet of the WGS in order to prevent back-leakage l of air through the discharge check valves.

- Redundant oxygen analyzers are provided for continuous sampling in a side stream taken off the process flow path. These analyzers create alarm messages both locally and in the main control room upon high oxygen )

level. Setpoints are specified to allow adequate time for operator action. The setpoint for alarm is adjustable, l but is generally at about concentration. This is approximately one fourth of the lower flammability limit for mixtures of hydrogen and oxygen. The side stream is returned to the process flow after sampling.

- A hydrogen analyzer is provided for direct measurement of hydrogen concentration in the sampling side stream. No specific alarm setpoint is assigned to this analyzer because of the broad range of expected hydrogen concentrations. By using the hydrogen analyzer reading and a flammability chart, operator can j assess the flammability potential of the gases during an upset situation which permits oxygen into the system. i

- A hydrogen monitor is provided to sample the ambient environment of the charcoal bed vault. This analyzer creates alarm messages both locally and in the main control room upon high hydrogen level. The setpoint for alarm is adjustable, but is generally at about 1% concentration. This is approximately one fourth of the lower flammability limit for mixtures of hydrogen and oxygen.

- The gaseous radwaste system is of welded construction. The piping and components are metallic conductors.

The entire system is electrically at the same potentia.1 which eliminates the buildup of static electricity and sparking.

- The gaseous radwaste system throttling and isolation valves are packless metal diaphram type which eliminate leakage in or out of the system through the stem seals.

l - liigh oxygen alarm automatically stops the degasifier and injects nitrogen for dilution.

These provisions satisfy GDC 3 and Acceptance Criterion II.B.6 of Section 11.3 of the SRP.

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1 NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 l

d. A brief discussion of compliance with GDC 60 is provided in SSAR Section 3.1. Relative to the gaseous radwaste system, GDC 60 specifies that means should be included to suitably control the release of radioactive  !

materials in gaseous effluents and that sufficient holdup capacity shall be provided, particularly where I unfavorable site environmental conditions can be expected to impose limitations on the release of such effluents I to the environment.

l Suitable control of releases from the gaseous radwaste system is provided by the radiation monitoring system (discussed in SSAR Section 11.5), which automatically terminates releases from the gaseous radwaste system when a high-activity setpoint is exceeded in the system discharge line.

Sufficient holdup time is provided by the gaseous radwaste system since the system can be isolated at any time.

The system is not nonnally in operation but is opera J as necessary when reductions in the reactor coolant )

system noble gas inventory are made. It is not expecteu that any alteration in the system operation will be i l necessary because of adverse meteorological conditions since anticipated operation in the system provides a l minimum of 61 days holdup of xenon isotopes and 2.2 days holdup of krypton isotopes. After this level of radioactive decay, all nuclides except for Kr-85 would be reduced to extremely small amounts, and the release of Kr-85 is not of consequence.

l In addition to the gaseous radwaste system, the exhaust air from several ventilation systems also may contain airborne radioactivity. Provisions to control radioactive releases from these systems to the environment are j described belo v.

1. The radiologically controlled area ventilation system (VAS) serves the fuel handling area (which encloses the spent fuel pool), radiologically controlled areas of the auxiliary building and the annex building. These areas are normally maintained at a slightly negative ambient air pressure with respect to adjacent clean plant areas and the environment to provide controlled release and monitoring of airborne effluents at the plant vent.

compliance with 10 CFR 20 effluent concent ation limits and 10 CFR 50 Apper. dix I dose guidelines for offsite eleases has been conservatively evalu&d assuming that the exhaust air discharged by the VAS to the plaat vent is unfiltered. Analysis indicates that no HEPA or charcoal filtration is required in order to keep normal plant releases within the specified limits. Therefore, the criteria set forth in RG 1.140 do not l specifically apply to the VAS exhaust air system. See SSAR Appendix 1 A for a discussion of conformance with Regulatory Guides.

The exhaust air upstream of the plant vent is monitored for airborne radioactivity (discussed in SS AR Section 11.5). In the event that abnormal airborne radioactivity is detected, the exhaust and supply air is automatically isolated to terminate unfiltered releases to the plant vent. As discussed in SSAR Subsection l 9.4 3.2, the exhaust ducts from the fuel handling area and auxiliary building areaa that have the greatest potential for a significant release of airborne radioactivity are sized to hold up the exhaust air, allowing the isolation dampers to close before the abnormal airborne radioactivity is exhausted. The exhaust from the 4%I high radiation area is then realigned to the containment air filtration system (VFS). The VFS exhaust h n and fans operate to maintain a slightly negative air pressure within the isolated area with respect to a-4"dN 3 Westinghouse i

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the surrounding clean plant areas to prevent unfiltered releases to the environment. The VFS exhaust air is HEPA and charcoal filtered before it is discharged from the plant vent.

2. The containment air filtration system provides HEPA and charcoal filtration of air exhausted from the containment during normal plant operation. Compliance with 10 CFR 20 effluent concentration limits and 10 CFR 50 Appendix I dose guidelines for offsite releases has been evaluated assuming that the exhaust filters remove 99 percent of the particulate contaminants and 90 percent of the radioiodides, based on the guidance provided in RG 1.140. The containment exhaust filters meet RG 1.140 guidelines to the extent practical. See SSAR Appendix 1 A for a discussion of conformance with Regulatory Guides. Most of the design standards referenced by RG 1.140 have been revised since the issuance of Revision 1 of the Regulatory Guide in October 1979. The filtration systems are designed in accordance with the updated design standards such as ASME N509-1989 and ASME N510-1989 and will provide decontamination efficiencies comparable to filtration systems designed in accordance with previous standards. The SSAR also includes the following key performance requirements, based on RG 1.140 guidelines, as a basis for assuming i RG 1.140 decontamination efficiencies: l l

l SSAR Table 9.4.7-1 states that the charcoal absorber provides an overall bed depth of 4 inches with an air residence time of 0.5 second. Thi.c is consistent with RG 1.140, Position C.3.g and Table 2 to provide an air residence time rated for a 90 percent decontamination efficiency.

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• SSAR Subsection 9.4.7.2.2 describes the electric heater provided to maintain the relative humidity of the incoming air so that it does not exceed 70 percent relative humidity. This is consistent with RG 1.140, Position C.3.a.

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SSAR Subsection 9.4.7.4 states that charcoal adsorbent is batch tested. This is perfocmed in accordance with ASM3 N509. ASME N509-1989, Section 5.2.3.2, provides adsorbent requirements for non-ESF absorbers by referencing ASME/ ANSI AG-1-1988, Section FF, which provides qualification and batch test requirements for new bulk charcoal. It is assumed that charcoal manufactured in accordance with ASME/ ANSI AG-1-1988 will be equivalent to charcoal meeting the criteria set forth in RG 1.140. Table 1.

• SSAR Subsection 9.4.7.4 states that both the HEPA filters and charcoal absorbers will be field-tested to verify that the bypass leakage rate does not exceed 0.05 percent in accordance with ASME N510-1989.

The bypass leakage rate of 0.05 percent is consistent with RG 1.140, Positions C.5.c and C.5.d.

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• SSAR Subsection 9.4.7.4 also states that periodic testing in accordance with ASME N510-1989 will be performed. This standard encompasses several Ley field-testing requirements that are also referenced in RG 1.140, including the following:

- Table 1 of ASME N510-1989 requires that representative samples of used charcoal adsorbent be periodically tested after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation (and adjustment made based on operating j history) or when inadvertently exposed to organic solvents. This should be comparable to RG 1.140, l Table 2 criteria, which recommend that tests be conducted at least once every 18 months or when 4m4m2N W

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NRC REQUEST FOR ADDITIONAL INFORMATION Respouse Revision 2 exposed to fumes, chemicals, or foreign materials that could adversely affect the charcoal. The objective is to have an effective testing schedule that allows the charcoal to be tested and replaced when it does not meet the required efficiency.

ASME N510-1989, Section 8.6.2, requires that airflow distribution tests across HEPA filters and charcoal absorbers meet an acceptance criterion of plus or minus 20 percent of the design airflow velocity. This is consistent with RG 1.140, Position C.5.b.

SSAR Subsection 9.4.7.4 states that the exhaust filtration units are designed in accordance with ASME N509-1989. Design compliance with ASME N509-1989 should be comparable to ANSI N509-1976, which was used as a reference for the guidelines in RG 1.140.

Compliance with the above criteria should provide assurance that the airborne effluents discharged from the containment will be controlled by filtration consistent with the guidance provided in RG 1.140.

In the event of abnormal containment airborne radioactivity, the containment high-range radiation monitors will automatically isolate the VFS supply and exhaust air containment isolation valves. Additionally, dedicated radiation monitors in the VFS exhaust lines will also provide an alarm signal if abnoanal releases to the environment are detected in the exhaust lines.

3. The solid radwaste building ventilation system (VRS) maintains the solid radwaste building at a slightly negative ambient air pressure with respect to the surrounding clean plant areas to provide controlled release

and monitoring of airborne effluents at the plant vent. The airborne radiological releases from this system are based on a conservative assumption that the exhaust air is unfi!tered since this area normally contributes a small fraction of the normal overall airborne releases. Analysis indicates that no HEPA or charcoal l

filtration is required in order to keep normal plant releases within 10 CFR 20 effluent concentration limits and 10 CFR 50 Appendix I dose guidelines. However, as noted in SSAR Subsection 9.4.8, the exhaust air from this area is provided with HEPA filtration in accordance with RG 1.140, which adds design margin for the normal calculated releases since no filtration credit, based on RG 1.140 design criteria, is assumed.

if abnormal airborne radioactivity is detected in the exhaust air, an alarm in the MCR is initiated for operator action.

, 4. The health physics and hot machine shop HVAC (VHS) maintains the areas that it serves at a slightly l negative ambient air pressure with respect to the surrounding clean plant areas to provide controlled release and monitoring of airborne effluents at the plant vent. The airborne radiological releases from this system are based on a conservative assumption that the exhaust air is unfiltered since thi., area normally contributes a small fraction of the normal overall airborne releases. Analysis indicates that no HEPA or charcoal filtration is required in order to keep normal plant releases within 10 CFR 20 maximum permissible concentration limits and 10 CFR 50 Appendix I dose guidelines. However, as noted in SSAR Subsection l 9.4.11, the exhaust air from this area is provided with HEPA filtration in accordance with RG 1.140, which j adds design margin for the normal calculated releases since no filtration credit, based on RG 1.140 design l criteria, is assumed.

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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 If that abnormal airborne radioactivity is detected in the exhaust air, an alarm in the MCR is initiated for operator action.

c. A brief discussion of compliance with GDC 61 is provided in SSAR Subsection 3.1.6. The fuel storage and handling areas for the AP600 include the fuel handling area of the auxiliary building, which encloses the spent fuel pool, and the containment building that encloses the reactor cavity. Based on the calculated radiological releases resulting from a design basis fuel handling accident (SSAR Subsection 15.7.4) in either area, there is no need to provide safety-related isolation or filtration systems to maintain plant safety.

As discussed in item (d), the ventilation systems serving these plant areas incorporate specific design features to mitigate the potential release of abnormal (non-DBA) airborne radioactivity from these areas. In addition to automatic isolation of the fuel handling area or containment purge valves on a high radiation signal, the isolation l

dampers or valves can be manually controlled from the main control room. The fuel handling area isolation dampers and containment isolation valves are provided with remote position indication to verify proper damper blade or valve disk position during isolation. During abnormal airborne radiological conditions, the containment l

! purge valves can be manually opened to override a high radiation signal, through administrative procedures, to allow cleanup of the containment atmosphere (SSAR Subsection 9.4.7.2.34).

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The supply and exhaust lines that penetrate the fuel handling and containment areas include isolation subsystems.

The VAS maintains the fuel handling area at a slightly negative air pressure during isolation, is redundant l

(SSAR Subsection 9.4.3.2.3.2) and can be manually connected to the onsite diesel generators if there is a loss l

l of offsite power (SSAR Subsection 9.4.7.2.34). As shown in SSAR Subsections 9.4.3.2.2 and 9.4.7.2.2, the isolation dampers and isolation valves are controlled by pneumatic operators that fail in a closed position on loss l

of electric power or air pressure.

SSAR Subsection 9.4.7.4 states that the VFS exhaust air filtration units are designed and tested in accordance with ASME N510. Instrumentation is provided as necessary for the periodic inspection and verification of system l

airnow rates, air temperatures, and filter pressure drops.

SSAR Revision: NONE 460.4(R2)-8 W-Westinghouse 1

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NRC REQUEST FOR ADDITIONAL INFORMATION u -- u Response Revision 2 l

l If that abnormal airborne radioactivity is detected in the exhaust air, an alarm in the MCR is initiated for l operator action.

c. A brief discussion of compliance with GDC 61 is provided in SSAR Subsectior. 3.1.6. The fuel storage and handling areas for the AP600 include the fuel handling area of the auxilir' Suithng, which encloses the spent l fuel pool, and the containment building that encloses the reactor cavit3 s d on the calculated radiological releases resulting from a design basis fuel handling accident (SSAR Suk aon 15.7.4)in either area, there is l no need to provide safety-related isolation or filtration systems to maintain plant safety.

l l As discussed in item (d), the ventilation systems serving these plant areas incorporate specific design features

to mitigate the potential release of abnormal (non-DBA) airborne radioactivity from these areas. In addition to automatic isolation of the fuel handling area or containment purge valves on a high radiation signal, the isolation l l dampers or valves can be manually controlled from the main control room. The fuel handling area isolation dampers and containment isolation valves are provided with remote position indication to verify proper damper blade or valve disk position during isolation. During abnormal airborne radiological conditions, the containment purge valves can be manually opened to override a high radiation signal, through administrative procedures, to l allow cleanup of the containment atmosphere (SSAR Subsection 9.4.7.2.34).

l The supply and exhaust lines that penetrate the fuel handling and containment areas include isolation subsystems.

l The VAS maintains the fuel handling area at a slightly negative air pressure during isolation, is redundant l l (SSAR Subsection 9.4.3.2.3.2) and can be manually connected to the onsite diesel generators if there is a loss l l of offsite power (SSAR Subsection 9.4.7.2.34). As shown in SSAR Subsections 9.4.3.2.2 and 9.4.7.2.2, the I l l isolation dampers and isolation valves are controlled by pneumatic operators that fail in a closed position on loss ,

of electric power or air pressure. l I

SSAR Subsection 9.4.7.4 states that the VFS exhaust air filtration units are designed and tested in accordance l l with ASME N5!0. Instrumentation is provided as necessary for the periodic inspection and verification of system I airflow rates, air temperatures, and filter pressure drops.

SSAR Revision: NONE l

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l 460.4(R2)-8 W-Westinghouse t

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i NRC REQUEST FOR ADDITIONAL INFORMATION P ii.

Response Revision 2

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l Question 460.5 I

Provide the following information regarding solid radwastes (Section 11.4):

a. Estimates of solid waste volumes expected to be shipped annually for wet solid wastes and dry solid wastes separately.

b. A discussion of compliance with Position 111.1 of BTP ETSB 11-3 regarding the storage capacity for accumulated filter sludges.

c. A discussion of compliance with Position III.2 of BTP ETSB 11-3 regarding storage volume for solidified wastes (both wet and dry solid wastes) available in the plant.

I Response: (Revision 2) j i

a. Estimates of solid waste volumes expected to be generated and shipped annually are separately provided below )

in Tables 1.A. and 2.A for wet solid wastes and dry solid wastes. These volumes are consistent with SSAR j l Table 11.4-1. In Table 11.4-1 the volume of wastes to be shipped accounts for volume reduction of some wastes l and addition of the shipping / disposal container. The processing and packaging factors used in SSAR Table

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l 11.41 are conservative and may have over-estimated the annual disposal quantities. Below each estimate of i annual disposal quantities are the details of the inputs to each category of generation (Tables 1.B and 2.B).

These waste generation estimates are based on an 18 month refueling cycle. For a 24 month refueling cycle, the annual waste generation is less.

Table 1.A - Annual Wet Solid Waste Generation and Disposal Quantities Summary Waste Generation Disposal Wet Solid Wastes Volume, ft3/vr Volume, ft'/yr Spent Charcoal and Ion Exchange Resins 250 314.8 l Mixed Liquid Wastes 15 17.0 Chemical Wastes 350 19.8 Total Wet Solid Wastes 615 351.6 The generation of these wet solid wastes is summarized as follows:

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NRC REQUEST FOR ADDIT!ONAL INFORMATION l .. ..

Response Revision 2 l

Table 1.B - Annual Wet Solid Waste Generation Quantitles Details Waste Source Waste Generation Rate Annualized Comments Waste Generation l

WLS Deep Bed Filter 10 ft' / 6 months 20 ft' Charcoal i WLS Deep Bed Filter 40 ft' / year 40 ft' l Zeolite l

3 x WLS lon Exchanger 30 ft' ea. / year 90 ft'

, Resin l

2 x CVS Ion Mixed Bed 50 ft' /18 months 33.3 ft' one vessel discharged each Exchanger Resin refueling j 1 x CVS Cation Bed 50 ft' / 36 months 16.7 ft' l Exchanger Resin 2 x SFS lon Exchanger 75 ft' /18 months 50 ft' one vessel discharged each Resin refueling

! Subtotal Wet Solid 250 ft' l Waste (Spent Resins j and Charcoal)

Mixed Liquid Wastes 4 gal / month for 17 15 ft' mainly contaminated lubricating l months + 100 gallons oil l

during refueling outage Chemical Wastes 200 gallmonth for 17 350 ft' themical laboratory and l months + 500 gallons decontamination wastes l during refueling outage l

Total Wet Solid Wastes 615 ft' 460.5(R2)-2 W.-

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I NRC REQUEST FOR ADDITIONAL INFORMATION l Response Revision 2 l

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Table 2.A - Annual Dry Solid Waste Generation and Disposal Quantities Summary )

Waste Generation Expected Disposal Dry Solid Wastes Volume, ft'/yr Volume, ft'/yr l l

Spent Filter Cartridges 3.6 24.5 l WGS Charcoal 8.7 9.8 l l Compactible DAW 4101 872.7 Non-Compactible DAW 225 363.4 Mixed Solid Wastes 5 7.5 l Total Dry Solid Wastes 4343 1277.9 The generation of these dry solid wastes is summarized as follows:

l Table 2.B - Annual Dry Solid Waste Generation Quantitles Details l Waste Source Waste Generation Rate Annualized Comments ,

Waste Generation WSS Resin Fines Filter 1 x 0.5 ft' cartridge / 0.5 ft' 6" dia. x 30" pleated fiberglas, year 25 microns @ 99% efficiency 2 x WLS Filter 4 x 0.3 ft' cartridges /18 0.8 ft' 6" dia. x 18" months CVS Makeup Filter 1 x 0.3 ft' cartridge /18 0.2 ft' 6" dia. x 18" months CVS Reactor Coolant 4 x 0.5 ft' cartridges /18 1.3 ft' 6" dia. x 30" Filter months SFS Filters 4 x 0.3 ft' cartridges /18 0.8 ft' 6" dia. x 18" months WGS Guard Bed 8 ft' / 36 months 2.7 ft' assumes abnormal charcoal Charcoal replacement WGS Delay Bed 60 ft3 /10 years 6 ft' assumes abnormal charcoal Charcoal replacement High Activity Dry 2 ft'/ month for 17 50 ft'

Compactible Waste months + 40 ft' during refueling outage 460.5(R2)-3 t

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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 l

i Low Activity Dry 60 ft'/ month for 17 4000 ft' l Compactible Waste months + 5000 ft' during i refueling outage l Compactible HVAC See Note 1 51 ft' l Filters High Activity Dry Non- 1 ft'/ month for 17 25 ft' Compactible Waste months + 20 ft' during refueling outage Low Activity Dry Non- 6 ft'/ month for 17 200 ft' Compactible Waste months + 200 ft' during i refuehng outage  ;

l Mixed Solid Wastes 0.2 ft'/ month for 17 5 ft' l months + 4 ft' during I refueling outage  !

l Total Dry Solid Wastes 4343 ft' Note 1: Waste HVAC filters are generated as follows -

Containment purge system (VFS) 13.7 ft' 36.7 ft' granular charcoal / 60 months 4 x 12" x 24" x 24" HEPA / 60 mc4I 2 x 12" x 24" x 24" Pre-filter / 60 months 2 x 12" x 24" x 24" Post-filter / 60 months l

Rad. Chem.12b. Exhaust (VAS) 21.3 ft' 2 x 12" x 24" x 24" HEPA /18 month 2 x 12" x 24" x 24" Pre-filter / 6 month l l i

l Rad. Machine Shop (VHS) 16 ft' l 4 x 12" x 24" x 24" HEPA / yr l 51 ft' e m2p4 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION m +

2

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Response Revision 2 Under normal conditions, there are no wastes generated from the secondary-plant cycle. If radioactivity is detected in the steam generator blowdown and reaches a predetermined level (due to primary and s:condary leakage), the blowdown is diverted to the to the liquid radwaste system for processing as described in SSAR Subsections 10.4.8 and 11.2.2.1.5. Thus, blowdown demineralizer resins are not a normal source of radwaste requiring shipping and disposal. Should plant operation continue with leakage from the primary to the secondary side utilizing the blowdown demineralizers, a shipping (disposal) solume of up to 300 ft'/ month could be produced by the steam generater blowdown system, as indicated in SSAR Subsection 11.4.2.1.

b. The AP600 does not include filters that generate sludge-type wastes. Also, tanks and sumps are designed to minimize the formation of sludge deposits, and the particulate matter that can cause sludge deposits is transported to and removed by the cartridge filters in the liquid radwaste system. Therefore, there are no storage provisions for accumulating sludges.

c. Branch Technical Position ETSB 11-3 specifies in Position III.2. that storage areas for solidified wastes should be capable of accommodating at least 30 days of waste generation at normal generation rates and that these storage areas should be indoors. The storage durations of the storage areas are evaluated for four general types of waste and container categories discussed below.

Spent Ion Exchance Resins and Filter Charcoal in Hich-integrity Containers (HICs)

Although the shipping or disposal volume is nearly independent of container size (based on equal filling efficiency, e.g.,90%), the storage duration for the filled HICs is dependent on the number of containers which is indirectly proportional to container size. To be conservative, it is therefore assumed that the spent resins and filter bed charcoal will be dewatered in HICs that will fit into a Type B shipping cask (i.e., the SEG 3-82B, formally HN-200). Normally 250 ft'/yr of spent resin and charcoal is expected to be generated (SSAR Table 11.4-4), with an activity of 950 curies (SSAR Table 11.4-6). This resin can be mixed to produce a uniform specific activity of 3.8 Ci/ft'. A 158 ft' HIC filled to 90% would contain about 540 Curies, well within the cask's capability. About two of these HICs are required per year. The three onsite storage casks can each hold one of these 1;lCs, and the resulting storage duration is about one and a half l years. The spent resin container fill station may also be used for storage until it is necessary to begin filling another HIC. The fill stations and two of the onsite storage casks (reserving one onsite storage cask for high-activity filter drums) provide about one and a half years of storage. These storage times are in addition to the pre-packaging storage times provided by the spent resin tanks as described in SSAR Subsection i1.4.2.2.1.

High Activity Filter Cartridges in Drums As indicated in SSAR Table 11.4-4, packaging of the CVS reactor coolant filter cartridges is expected to l normally generate 3 drums of waste per year. This is based on a generation rate of 4 filter cartridges every 18 months. The high-activity filter storage tube module may be used to store all filter cartridges normally l generated every 18 months. Thus, after a drum is filled with high-activity filters, encapsulated. and sealed, W Westinghouse 1

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l NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 it may remain in the processing cask until it is necessary to begin to package the filter cartridges stored in the storage tube module to clear space for the next batch of spent filters. Therefore, a storage duration of about 17 months is available for high-activity filters using only the processing cask. One of the onsite storage casks could also be used for high activity filter drum storage if necessary.

Other Wastes in Drums Based on SSAR Table 11.4-4, about i1 drums are produced each year containing wastes other than high activity filter and mixed wastes. The Radwaste Building (Proprietary Figure 1.2 29) has a packaged waste storage room that may be used to store both drums and boxes. Using two storage locations for palletized drums stacked three high,24 drums can be stored. This provides about 28 months of storage for the normal expected generation rate. Stacking only two pallets high provides about 18 months of storage. Without stacking about 9 months of storage is available for the normal generation rate.

Mixed wastes are accumulated in drums and are sent to an off site processing facility at an expected rate of about three drums per year.

Wastes in Boxes Based on SSAR Table 11.4-4, about 12 boxes are generated per year. Ten box storage locations are available in the packaged waste storage room. Without stacking and with stacking two and three high, about 1, 2, and 3 years of storage are provided, respectively.

Maximum truck loading is expected to be 28 boxes. Ten storage locations can accumulate a truck load when stacked three high. At the normally expected generation rate,it takes 2 years to produce a tmck load.

In summary, indoor storage is provided for all categories of packaged wastes well in excess of 30 days, based on normally expected waste generation rates.

l SSAR Revision: None l

a s(R2 s W westinghouse I

! NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 -

Question 460.7 Table 2 of Section 11.5 of the SRP, " Process and Effluent Monitoring Instrumentation and Sampling Systems,"

includes a service water system effluent monitor. The staff notes that AP600 design includes an upstream provision for this monitor in the form of component cooling water system monitor. The staff does not consider an upstream provision as an adequate basis for eliminating a downstream provision for this monitor. Therefore, include a service water system monitor or justify its elimination (Sections 11.5).

Response: (Revision 2) l The AP600 service water systems design has been changed to include an effluent radiation monitor. The description l is in section 11.5.2.3.1 of SSAR, Rev. 9.

SSAR Revision: NONE l

! 460.7(R2)-1

W_

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NRC REQUEST FOR ADDITIONAL INFORMATION

+. u Response Revision 2 Question 460,10 Provide the following information regarding the gaseous radwaste management system (Section i1.3):

a. Tables 11.2-6 and 11.3-1 of the SSAR give different holdup times for Xenon and Krypton a the charcoal delay beds. Clarify the discrepancy between the tables.

b. Describe the basis for the RCS degassing days (17.4 and 1.0 for Xenon and Krypton, respectively) given in Table 11.2-6 of the SSAR. The waste gas system releases given in Table 11.3-3 of the SSAR do not appear to be correct. Confirm the acceptability of this information or correct it, as appropriate.

c. Discuss the provisions for monitoring the individual performance of the equipment within the charcoal delay bed system. Include a list of alarmed process parameters for the delay bed system.

Response

l a. The delay times shown in Table i1.2-6 are based upon the NRC specified isotopic dynamic l adsorption coefficients used in running the PWR-GALE Code. The delay times listed in Table l 11.3-1 are WGS system design performance characteristics based upon calculations using different l isotopic dynamic adsorption coefficients taken from the results of adsorption testing.

l b. Table 11.2-6 has been revised to remove the RCS degassing days information. Table 11.3-3 has l been revised to show the results of recent PWR-GALE Code runs.

c. As show i on Figure 11.3-2, the AP600 gaseous radwaste system contains provisions for continuously monitoring the moisture level at the inlet of the guard bed in order to provide confidence that moisture will not intrude beyond the moisture separator, which may adversely affect system performance.

Monitoring performance of individual components in the gaseous radwaste system is done by collecting and analyzing grab samples. As shown on Figure 11.3-2 sample pumps and connections are provided which allow the collection of grab samples at the inlet and outlet of the guard bed, between the two delay beds, and at the outlet of the second delay bed. A list of gaseous radwaste l system instrumentation and control items is provided in SSAR Table 11.3-2.

SSAR Revision: None 3 Westi'Igh00$0

r l NRC REQUEST FOR ADDITIONAL INFORMATION l WE Response Revision 2

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Ouestion 460.11 1

Provide the following information regarding the solid radwaste management system (Section 11.4):

a. Since solidification and encapsulation are not the same, clarify whether either of the above two options may be l

used for processing spent resins in addition to a third option, namely dewatering the resins (note that encapsulation is not generally used for processing spent resins). Also, clarify whether the AP600 solid radwaste management system design deviates from the EPRI Requirements Document for passive reactor designs. The Requirements Document recommends only dewatering for processing the spent resins.

b. Identify the specific design features provided in the system design to comply with GDCs 60,63 and 64 as they relate to (1) control of release of radioactive materials to the environment from the plant areas where the solid j radwastes are processed, and (2) monitoring radiation levels and leakage.

c. Clarify whether the description and discussion of acceptability of the portable grouting unit that may be used j for processing spent filters is within the COL applicant's scope. If it is within the AP600 design scope, provide j specific details of the unit.

d. The staff is concerned that the projected (Tables 11.4-4 and 11.4-5 of the SSAR) annual solid radwaste volumes to be disposed (1729 CF for the expected case and 3843 CF for the maximum case) are significantly lower than that actually shipped volume for operating PWRs (EPRI NP-5528, February 1988, Volume 2 - Plants Without Evaporators for the Years 1985 and 1986: 9550 CF). The staff recognizes that the projected volume agrees with the value proposed in the EPRI Requirements Document (1750 CF per year). The EPRI-proposed value depends on following what EPRI regards as sound design and operating techniques outlined in the document (Paragraph B.l.2.2 of Appendix B of Chapter 12) for reducing the shipment of processed solid waste volume. One of the operating techniques is to avoid solidification and instead use only dewatering for solidifying the wet solid wastes. As stated above, the AP600 design includes solidification as one of the options. The staffis concerned that the storage volume allotted for processed solid wastes may be inadequate ifit is to be based on the projected shipment volumes given in the SSAR tables. Therefore, provide justification for the projected volumes given in the subject SSAR tables or revise the values as appropriate.

c. Clarify why the AP600 design does not include phase separator tanks, as recommended in the EPRI Requirements Document for passive reactor designs.

f. Section i1.4.1.3 of the SSAR identifies the capability to store processed and packaged solid wastes at the site for at least six months to account for possible delay or disruption of offsite shipping of the wastes as one of the design objectives of the solid waste management system. However, there is no description of the on-site storage facility in the SSAR. Provide a description of the facility, and clarify whether it conforms with the recommendations identified for such a facility in Section 5.4 of Chapter 12 of the EPRI Requirements Document for passive reactor designs.

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460.11(R2)-1

NRC REQUEST FOR ADDITIONAL INFORMATION l

t Response Revision 2 Response: (Revision 2) l l

l a. References to resin solidification and encapsulation were deleted from Section 11.4, except as related to space reserved for future or optional solidification facilities. i

b. Relative to solid wastes, General Design Criterion 60, Control of Releases of Radioactive Materials to the Environment, requires that means be provided to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Means are provided in the solid waste management system to handle the applicable categories of solid radwaste as indicated in SSAR Section 11.4. To control the release of radioactive materials to the environment, the areas and components in the radwaste buildings that l l process or house radioactive solid wastes are located in areas with exhaust ventilation that discharges to the l radiologically monitored plant vent as indicated in SSAR Subsection 9.4.8.2.

Sloped floors and floor drains are provided to collect and thereby control the release of radioactive material that could be removed from stored solid waste by water contact.

General Design Criterion 63, Monitoring Fuel and Waste Storage, requires that means be provided to detect conditions that may result in excessive radiation levels and to initiate appropriate actions. For the solid waste management system, the wastes with the most potential for high radiation levels are the spent ion exchange resins and filter cartridges, especially those from the chemical and volume control system ion exchangers and l l filters. The radiation levels of the spent resin tanks can be monitored without entering the rooms. A floor l penetration above the spent resin tanks used for personnel access via a ladder allows the radiation levels in the tank rooms to be monitored by lowering detectors down the outside of the tanks.

As described in SSAR Subsection 11.4.2.3.2, the dose rates of high-activity filter cartridges are measured during l the changeout process when the filter is raised into the high-activity filter transfer cask (but before the bottom cover of the shield cask is secured) using a long-handled radiation probe. The measured dose rate determines the precautions taken during subsequent handling operations. The high-activity filter cartridges can be l transferred into and out of the high-activity filter storage using the high-activity filter transfer cask without direct exposure to personnel. The filters in storage can be monitored with minimal exposure at any time through l sampling ports (normally closed by shielded plugs).

Dry, solid wastes are normally monitored when received at the radwaste building and are then transferred to the appropriate temporary storage location depending on the measured dose rate as described in SSAR Subsection 11.4.2.3.3. Local shielding can be used within the temporary and packaged waste storage areas to segregate the higher dose rate items and thereby minimize the dose rate in the rest of the storage areas. The radiation levels in the temporary and packaged waste storage areas should be relatively low. The areas can be entered for monitoring at any time. These storage areas have doors that can be locked to control access.

Relative to solid radwas'es, Criterion 64, Monitoring Radioactivity Releases, requires that means be provided for monitoring efnuent d3 charge paths and the plant environs for radioactivity that may be released from normal l operations, including anticipated operational occurrences, and from postulated accidents. Airborne effluents f rom

the auxiliary and radwaste buildings are monitored by exhaust radiation monitors, as described in SSAR 460.11(R2)-2 W Westinghouse I

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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 l Subsection 11.5.2.3.2. Drums and boxes containing filters and lower-activity dry wastes are also surveyed and decontaminated. Mobile or portable equipment may be used for clean waste monitoring to verify that wastes segregated and sorted for nonradioactive disposal are nonradioactive. Hand-held survey meters are used to l prevent removal of radioactivity from the radwaste building by personnel. The arrangement of the radwaste building allows the corridors and vehicle access areas to be very low radioactivity areas, thereby minimizing the need for any decontamination operations. Liquid wastes generated from solid radwaste system operations are discharged directly to the liquid radwaste system (WLS) or are collected by the radioactive waste drain system (WRS) and directed to the WLS for subsequent processing and monitored discharge.

c. Encapsulation of spent filters is by mobile or portable equipment.

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d. The AP600 is smaller and simplified relative to current plants, including reductions in the quantity of valves, l pumps, and other components requiring maintenance, the generation of solid wastes for the AP600 should be I within that produced by the best 10 percent of current plants. AP600 uses mobile or portable processing systems for wet and dry solid wastes, allowing the most efficient volume reduction methods and equipment to be used.

The AP600 waste quantities are based on dewatering wet solid wastes and does not include solidification.

Solidification of spent resins would result in no change in the waste volume shipped from the plant. The vinyl ester styrene binder process does not increase the waste volume but Ells the space existing between resin beads I l with binder. Reference to resin solidification has been deleted from the SSAR .

The overall performance of the nuclear power industry in reducing the volume of solid radioactive waste shipments may be observed by evaluating the historical data provided in NUREG/CR-2907'. The average solid 4

radioactive waste shipments for operating PWR power plants from 1978 to 1988 is about 1.06 x 10 m'/yr per ,

MWe-Hr. Thus, the 1978 to 1988 average waste shipments for PWRs is about nine times greater than the l AP6001750 ft'/yr goal. The average solid radwaste shipments for PWRs reduces to about 4.4 x 105 m'/yr per j MWe-Hr from 1985 to 1988, which is about four times the AP6001750 ft'/yr goal. This reduction by over a j factor of two shows the results of early efforts to reduce the volume of solid radioactive waste shipments. j Some plants with aggressive radioactive waste volume reduction programs did much better than the average. ,

For example, between 1985 and 1988, Diablo Canyon I and 2 shipped an average of 1.03 x 10 m'/ year per MWe-Hr. This is about 10 percent less than the AP6001750 ft'/yr goal. l l

With aggressive solid radioactive waste programs, current nuclear power plants are being operated within the AP6001750 ft'/yr goal for solid radioactive waste volume. It is expected that evaluation of shipped radwaste volumes since 1988 would show continued reductions as plants minimize their solid wastes.

For the AP600 there is a large reduction in the number of components (pumps, valves, etc) that can become radioactively contaminated. This will result in a large reduction in the generation of solid radioactive wastes l

NUREG/CR-2907 (BNL-NUREG-51581, Vol. 9) Radioactive Materials Released from Nuclear Power Plants, Annual Report 1988, published July 1991.

pam. 460.11(R2)-3 l

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l NRC REQUEST FOR ADDIY!ONAL INFORMATION

! - m-Response Revision 2 )

t due to maintenance operations. The waste generated during maintenance operations is a large fraction of the l volume of dry, compressible waste and contaminated equipment. For Diablo Canyon in 1988, this waste accounted for about 89 percent of the total radioactive waste shipments. For an AP600 radwaste estimate, this l type of waste accounts for about 83 percent of the total expected annual disposal volume (SSAR Table 11,4-1).

l l l Please see the response to RAI 460.5, Rev. 2, for more detail on the quantities of solid radwaste projected to l be generated and disposed by the AP600.

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e. The AP600 employs only cartridge filters and does not generate sludges that require settling and decanting.

f. There are three locations where packaged wastes may be stored until shipped to a disposal facility: (1) the l spent resin waste disposal container and (2) the packaged waste storage room. The following is a physical  ;

description of each storage area. l l The spent resin waste disposal container (SSAR Subsection i1.4.2.5.2) is in a cell with thick concrete walls (See SSAR Figure 1.2-29). A thick shield cover, with lifting provisions forms the top of the cell.

l The packaged waste storage room (SSAR Subsection 11.4.2.5.2) is a shielded, unobstructed area and has a clear l height for stacking waste boxes or pallets of drums. The mobile systems facility crane is used to handle waste boxes and palletized waste drunis into and out of storage (SSAR Subsection 11.4.2.3.3). Planned positioning of waste containers and portable shielding may be used to minimize the dose rate in the portions of the area periodically accessed by personnel. Mobile racks for hanging lead blankets or shield panels on casters may be l used for flexible response to changing conditions in the storage area. j Although the ALWR Utility Requirements Document (URD) is not a regulatory requirement, the storage facilities are in conformance with URD Chapter 12, Section 5.4-l SSAR Revision: None l

l 460.11(R2)-4 W-Westinghouse

l NRC REQUEST FOR ADDITIONAL INFORMATION ish l Response Revision 1 i- e 1 l Question 460.12 Provide justification for concluding that the exhausts to the envimnment from the personnel areas in the Annex I building, electrical and mechanical equipment rooms in the Annex I and auxiliary buildings, and the diesel generator l rooms will not be radioactive and, therefore, need not be monitored. (Sections 9.4 and 11.5) l l Response:

l SSAR Section 11.5.1.1 has been revised to include a statement that the ventilation exhausts from the non-radioactive l personnel areas of the annex building, the electrical and mechanical rooms in the annex and auxiliary buildings and l the diesel generator room are not monitored. Justification for not monitoring exhausts for radiation from the l personnel areas in the annex building, electrical and mechanical equipment rooms in the annex and auxiliary buildings, and diesel generator rooms follows.

Diesel Generator Rooms The diesel generator rooms are in a stand-alone diesel generator building, which is separate from any other plant building. Outside air is utilized for building ventilation. The building houses only the diesel generators and mechanical and electrical components directly related to diesel generator operation. The building does not store, utilize, or contain any radioactive materials.

The building does not contain any radioactive materials and has no potential for the transfer of radioactive materials from other buildings via piping, ductwork, or building connections. Since any exhaust from the building to the environment has no potential for radioactivity, monitoring is not required.

l Annex Building Personnel Areas and Equipment Rooms and Auxiliary Building Electrical Rooms l The annex building personnel areas occupy the first and second floors of the building. The electrical and mechanical l rooms are on the second and third floors of the annex building. A partial third floor of the annex building contains l an HVAC equipment room. Portions of the annex I building are adjacent to the auxiliary building; however, the majority of the building walls and the entire roof are not adjacent to any other building or plant area.

Interfaces with the adjacent buildings are limited to doorways, airlocks, and ductwork. Doorways from the annex l building open to clean areas of the auxiliary buiMing. Ductwork connecting the annex building and adjacent areas consists entirely of supply air ductwork handling outside air for the fuel handling area, health physics area, containment purge supply, and main control room. The main control room supplemental air filtration unit is in the HVAC equipment room; however, this unit has no radioactive material during normal plant operation.

l The annex building general area HVAC system normally maintains the personnel areas at a slightly positive pressure with respect to adjacent areas.

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T Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION I 1

The auxiliary building contains two electrical penetration rooms and two reactor trip switchgear rooms that are served l by the annex / auxiliary nonradioactive ventilation system. These rooms are on elevations 100' and 117'-6" of the l

auxiliary building adjacent to a controlled access corridor for the main control room, main control room areas, and )

other electrical equipment areas. I The areas do not contain any radioactive materials. The potential for the transfer of radioactive contamination from l adjacent areas is prevented through interfaces with only clean areas and through general pressurization of the annex building. No potentially contaminated ductwork is contained in the areas. Therefore, any exhaust from the areas to the environment has no potential for radioactivity. Monitoring is not required.

SSAR Revision: NONE l

460.12(RI)-2 W

Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Etli t1 Response Revision 1 k- I Question 460.13 Clarify whether the monitors provided in the exhaust ducts of the Annex II building, the fuel handling area of the auxiliary building, and the radiologically-controlled portion of the auxiiiary building automatically facilitate connection of the applicable exhaust (i.e., monitor detects high radiation in the associated exhaust duct) to the containment air filtration system (Sections 9.4.3.1.2 and 11.5).

Response

l The radiation monitors in the exhaust ducts of the annex building, the fuel handling area of the auxiliary building, l l and the radiologically controlled portion of the auxiliary building are described in SSAR Section 11.5.2.3.2. A high radiation signal from these monitors isolates the normal supply and (unfiltered) exhaust ducts to th( affected area and automatically connects the containment air filtration system (VFS) exhaust filters and fans (SS AR Figure 9.4.7-1, l Sheets 1 and 2) to the isolated area. As discussed in SSAR Subsections 9.4.3.2.3.1 and 9.4.3.2.3.2, the VFS exhaust fans are used to maintain a slightly negative air pressure in the isolated zones to prevent unfiltered releases to the !

environment during conditions of high airborne radioactivity in these areas.

SSAR Revision: NONE l

l

[ W85tiflghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 3 Question 460.15 Section 9.3.3,11.5.3, and 11.5.4 of the SSAR provide incomplete information on radiological sampling provisions for process and effluent streams. For example, the sampling provisions for the waste monitor tank contents, the detergent waste monitor tank contents, the steam generator blowdown, and the condenser air removal system have not been identified. Further, there is no reference to tritium measurements. Identify how the sample provisions for the liquid and gaseous process and effluent streams for the AP600 design meet the sampling provisions for such streams identified in Tables 1 and 2 of Section 11.5 of the SRP.

Response (Revision 3) l SSAR Table 9.3.3-2 has been revised to describe the radiological sampling provisions for the monitor tank contents, 1 l the steam generator blowdown and the condenser air removal system. Reference to tritium measurements is included.

Sampling is performed to measure thos: water chemistry and radioactivity characteristics which must be monitored.

The ranges and accuracy of analysis will be appropriate for the water chemistry characteristics being measured.

These monitoring frequencies are selected so there is sufficient time to detect chemistry or radioactivity changes before any adverse affects occur. '

SSAR Tables 11.5-1 and 11.5-2 list the radiation detectors in the AP600. Tables 460.15-1 and 460.15-2 identify other SSAR items which corresponds to the SRP Tables I and 2 sampling and monitoring provision items.

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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 3 Table 460.15 Gaseous Radiological Monitoring and Sampling No. Item SSAR Reference l 1 Waste Gas Holdup System Section 11.5.2.3.1; Table 9.3.3-2, items 27, 28, 29, 30; Table 11.5-1 l 2 Condenser Evacuation System Section 11.5.2.3.3; Table 9.3.3-2, item 31; Table 11.5-1 3 Vent & Stack Release Pt. System Section 11.5.2.3.3; Table 11.5-1 4 Containment Purge System Section 11.5.2.3.2; Table i1.5-1 5 Aux. Bldg. Ventilation System Section 11.5.2.3.2; Table 11.5-1 6 Fuel Storage Area Vent System Section 11.5.2.3.2; Table 11.5-1 7 Radwaste Area Vent. System Section i1.5.2.3.2; Table 11.5-2 l 8 Turb. Gland Seal Cond. Vent included in item 13; Table 9.3.3-2, item 32 System 9 Mech. Vacuum Pump Exhaust Included in item 2 (Hogging) System 10 Evaporator Vent Systems N/A i1 Pre-treatment Liquid Radwaste included in item 5 Tank Vent Gas System 12 Flash Tank and Steam Generator Flash tank - N/A Blowdown Vent System Steam generator blowdown - Section 11.5.2.3.1 l Table 9.3.3-2, item 7 (See also Table 460.15-2 item 16) l 13 Turbine Bldg. Vent System Section 11.5.2.3.3 System description in Section 9.4.9.

l 14 Pressurizer & Boron Recovery Pressurizer - N/A See note 1.

Vent Systems Boron recovery - N/A Notes: 1. Pressurizer is vented to sampling system, reactor coolant drain tank or containment atmosphere.

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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 3 Table 460.15 Liquid Radiological Monitoring and Sampling No. Item SSAR Reference l 1 Liquid Radwaste (Batch) Effluent System Section i1.5.2.3.3; Table 9.3.3-2, items 11,15,16,17,18, 19,20,21 Table 11.5-1 2 Liquid Radwaste (Continuous) Effluent N/A System l 3 Service Water System Section 11.5.2.3.1; Table 9.3.3-2, item 23; Table 11.5-1 l 4 Component Cooling Water System Section 11.5.2.3.1; Table 9.3.3-2, items 12,13,14; Table 11.5-1 l 5 Spent Fuel Pool Treat. Syst. Table 9.3.3-2, item 8 6 Equip. & Floor Drain Collection and Radioactive floor drain system included in item 1.

Treatment Systems Clean floor drain system - System description in Section 9.3.5. See item 20.

7 Phase Separator Decant & Holding Basin N/A Systems 8 Chemical & Regeneration Solution RadlChem. lab waste included in item 1.

Waste Systems Regeneration chemical waste - N/A  ;

9 Laboratory & Sample System Waste Included in item 1. '

Systems 10 Laundry & Decontamination Waste Laundry - N/A Systems Decontamination waste included in item 1.

l 11 Resin Slurry, Solidification & Baling Spent resin - Table 9.3.3-2, item 22; Solidification & Baling Drain Systems - N/A; Table 11.5 12 Radwaste Liquid Tanks (outside the N/A buildings) 13 Storm & Underdrain Water Syst. The storm drain system is site specific.

l 14 Tanks and Sumps inside Reactor Reactor coolant drain tank - Table 9.3.3-2, item 10.

Building Containment sump vents to containment atmosphere.

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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 3 l

No. Item SSAR Reference 15 Boron Recoveiy System Liquid Effluent N/A l 16 Steam Generator Blowdown (Batch) Section 11.5.2.3.1; Table 9.3.3-2, item 7; Table 11.5-1 Liquid Effluent System 17 Steam Generator Blowdown Section 11.5.2.3.1 (Continuous) Liquid Effluent System 18 Secondary Coolant Treat. Waste & See item 20 Turbine Bldg. Drain Systems l

19 Ultrasonic Resin Cleanup Waste Systems N/A 20 Non-Contaminated Waste Water & PWR Section 11.5.2.3.3; Table i1.5-1 Turbine Building Clean Drain System l SSAR Revision: None i l

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NRC REQUEST FOR ADDITIONAL INFORMATION I t *t Response Revision 1 k-Question 460.16 Provide the following information regarding accident monitoring instrumentation (Section i1.5). State if any of these -

items are outside the design scope of the AP600, but are within the design scope of the COL applicant.

a. The recommended range for the noble gas effluent monitor for the condenser air removal system (Revision 3 to RG 1.97) is 10 pCi/cc to 105 pCi/cc. The monitor is not needed if the effluent discharges through a common plant vent (however, this is not the case for the AP600). Table 11.5-4 1 of the SSAR provides a much narrower range, i.e.,10~* pCi/cc to 10 pCi/cc. Why is the range so limited?

b. Describe the calibration frequency and technique for calibrating the monitors.

c. Describe the methods used to ensure representative measurements are taken with appropriate background correction.

d. Describe the location of instrument readout (s) and the methods of recording this information, including the method or procedure for transmitting or disseminating the information or data.

e. Provide assurance of the capability to obtain readings at least every 15 minutes during and following an accident.

f. Describe the procedures or calculation methods to be used for converting instrument readings to release rates per unit time, based on exhaust air How and consideration of radionuclide spectrum distribution as a function of time after shutdown.

g. Describe the sampling system design, including the sampling media, to demonstrate how the design meets the requirements identified in Clarification No. 2 of NUREG-0737, " Clarification of TMI Action Plan Requirements" (page II.F.1-7).

h. Describe the sampling technique to be used under accident conditions to demonstrate how the technique meets the requirements identified in Clarification No. 3 of NUREG-0737 (pages II.F.1-7 and II.F.1-8).

i. Describe the sampling technique to ensure the system capability to collect and analyze or measure representative samples of radioactive iodines and particulates in plant gaseous effluents during and following an accident as identified in Table II.F.12 of NUREG-0737 (page II.F.1-9).

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Response

l a. The condenser air removal efnuent discharges through a vent that is independent of the plant vent. Table 11.5-1 4

l has been revised to show a nominal range from 10 to 10' pCi/cc. As stated in Note 4 of the table, the wide l range is accomplished using two G-M tubes.

b. The radiation monitors are calibrated every refueling outage or following an unsatisfactory functional test or replacement of a major monitor component. Each detector is given an isotopic calibration using decay-corrected sources. The monitor electronics are independently calibrated using simulated (electronic) input signals.  ;

c. Grab samples of the process and efnuent streams are taken and analyzed in the plant chemistry laboratory. The results of the laboratory analysis are used to correlate the readings obtained from the in-situ radiation measurements. The installed radiation monitors are then electronically adjusted.

l d. As stated in SSAR Section 11.5.2.1, the radiation monitoring system uses distributed radiation monitors, where l each radiation monitor consists of one or more radiation detectors and a dedicated radiation processor. Each l radiation processor receives, averages and stores radiation data and transmits alarms and data to the plant control l system (protection and safety monitoring system for safety-related monitors) for control (as required), display l and recording. These alarms include: low (fail), alert, and high. Selected channels have a rate-of-rise alarm.

l Storage of radiation readings is provided.

l l Radiation monitoring data, including alarm status, are provided to AP600 operators via the plant control system l (and the protection and safety monitoring system for Class lE monitors). The information is available in either l counts per minute (c.v.mt rate), microCuries/cc (activity concentration), or R/hr (radiation dose rate).

e. Radiation readings for those monitors that are required post accident are continuous. The monitors are provided with reliable, uninterruptible power, which ensures their availability for postaccident monitoring.

f. The local radiation processor associated with the plant vent monitor accepts analog signal inputs from plant vent effluent flow and temperature. These analog signals are used to calculate concentrations and flow rates at standard conditions. These sigeals are also used by the radiation processor to calculate total process and sample How for an operator-selected period and total discharge for an operator-selected period.

Software is provided to determine the radioactive release rates as a function of time by combining the gross concentrations and flow rate provided by the plant vent monitor with internal!y stored radionuclide spectrum distribution as a function of time for different types of accidents. Decay of the spectra from the times that the accidents begin to the time that the release rates are calculated is also included. Capabilities to manually or automatically enter actual spectra from laboratory analysis of grab samples are also included in the software.

g. The particulate and iodine filters of the extended range plant vent monitor are used as grab sample modules to l provide the capability to collect representative samples of iodines and particulates for onsite analysis during and i following an accident. Filters on the high-activity sample flow path are housed in shielded enclosures designed l for quick removal and replacement of filter media. Filter removal is provided by quick-disconnect couplings.

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Response Revision 1 l After removal, the filter is placed in a shielded cask for transport to the onsite laboratory. The filter enclosure and transport cask are designed and the access routes are selected in accordance with the requirements of NUREG-0737 to keep personnel exposure in sample handling and transport below the GDC 19 limits of 5 rem whole-body exposure and 75 rem to the extremities during the accident. The iodine filter is a fixed silver zeolite filter, hydrated in accordance with the recommendations of Information Notice No. 86-43.

h. SSAR Subsection 11.5.2.3.3 describes the technique used to keep the sample flow isokinetic with the exhaust 1 flow. The radiation processor associated with the plant vent monMor accepts analog signals from plant vent effluent flow and temperature. These signals are used by the radiation processor to control the sample flow to maintain isokinetic extraction at the sample nozzles for vent exhaust flow rate variation of 20 percent. The normal-range detectors are deactivated automatically when the concentrations exceed their normal ranges. Also, the sample flow bypasses the normal range skid, and only a small portion is extracted for the extended range skid. , .*

i. A representative sample of the plant vent effluent is collected isokinetically from an array of nozzles in the plant l

vent. The sample tubing run from the nozzles to the radiation monitor skids and the number of bends and '

elbows are minimized to reduce the losses of iodides and particulates. Heat-tracing is provided to eliminate entrained moisture that could degrade the sample filter media. As described in item g, provisior.s are made to limit the occupational dose to personnel below the GDC 19 limits during sample handling and transport. .

1 The onsite laboratory is fully equipped with the instrumentation and equipment required to perform isotopic analyses from grab samples.

l l SSAR Revision: None i

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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Question 460.18 Because specific compliance with Appendix 1 of 10 CFR Part 50 for gaseous and liquid effluents (which includes offsite dose guidelines and cost-benefit analysis criterion), and the guidelines given in ANSI N13.1 " Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities," Regulatory Guide (RG) 1.21 " Measuring and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," and RG 4.15, " Quality Assurance for Radiological Monitoring Programs (Normal Operation)- Efiluent Streams and the Environment," is not within the scope of the AP600 design, the staff will review individual COL applications referencing the AP600 design to ensure their conformance with these documents. Section 1.8 of the SSAR summarizes the COL and site dependent interfaces, but these are not complete.

Besides the above, the setpoints for terminating instantaneous discharges of liquid waste and processed waste gas

[which will be given in site-specific offsite dose calculation manual (ODCM)] will be reviewed on a site-specific basis. Also, conformance with 10 CFR Parts 61 and 71 for processed solid wastes will be reviewed on a site-specific basis. The COL applicant may be required to prepare an operation and maintenance manual to demonstrate compliance with Section 50.34(f)(2)(xvii) of 10 CFR Paet 50 as it relates to noble gas effluent monitoring and sampling and analyzing plant gaseous effluents for radioiodine and particulates during and following an accident (the above regulation incorporates the guidelines of TMI Action Item II.F.1, Attachments 1 and 2). Also, the staff will review the COL applicant's Process Control Program (PCP) for processing " wet" solid wastes. The staff will also review quality assurance (QA) provisions for liquid, gaseous and solid waste management systems against RG 1.143 guidelines.

Therefore, the staff believes that all the following items should be identified as COL Action Items in Sections 11.2 through 11.5 of the SSAR:

Section 11.2 (Liquid Waste Management System)

The COL applicant should provide:

a. A statement of specific compliance with Appendix I numerical objectives for offsite individual doses via liquid effluents and cost-benefit analysis for population doses via liquid effluents.

b. The basis for set point calculation in the plant-specific ODCM for terminating liquid waste discharge:

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bC i bquid etiluent lima for im*9e in any unresuiled area For the C,,y limit, see 10 CFR Part 20, Sections 20.1001-20.2402, Appendix B, Table 2, Column 2.

c. A statement of compliance with the QA provisions of RG 1.143.

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Section 11.3 (Gaseous Waste Manacement System) i The COL applicant should provide- '

a. A statement of specific compliance with Appendix I numerical objectives for offsite individual doses via ]

gaseous effluents and cost-benefit analysis for population doses via gaseous effluents. i

b. The basis for set point calculation in the plant-specific ODCM for terminating instantaneous GWPS discharge: This should be based on instante.neous dose rates in unrestricted areas due to radioactive materials released via gaseous effluents. The following are the limits: noble gases 500 mrem /yr total body; 3000 mrem /yr skin; others 1500 mrem /yr to any organ. Note: Instantaneous rate here means the above annual dose rates prorated for I hour,

c. A statement of compliance with the QA provisions of RG 1.143 l l

Section 11.4 (Solid Waste Management System) l The COL applicant should provide:

a. A demonstration that the wet waste processing will result in products that comply with 10 CFR 61.56 l b. The ut.blishment and implemer.4stion of a PCP for processing wet solid wastes, i.e., solidifying (if appiedie) using an approved solidification agent and the dewatering processing of spent resins.

c. A discussion of on-site storage oflow-level waste and demonstration that such a facility will meet GL 81-38 guidelines (only if applicable),

d. A demonstration that all radioactive waste shipping packages will meet 10 CFR Part 71.

e. A statement of compliance ws the QA provisions of RG 1.143.

Section 11.5 (Process and Effluent Monitoring and Samr>1ine Systems The COL applicant should provide:

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! a. Sampling details and demonstration of compliance with RGs 1.21 and 4.15, and ANSI N13.1 guidelines.

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b. An operation and maintenance manual to demonstrate coropliance with 10 CFR Part 50, Section 50.34(f)(2)(xvii) with regard to monitoring and sampling of gaseous effluents during and following an accilent.

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Response

Chapter 2 of the SSAR defines the site-related parameters for which the AP600 plant is designed. These parameters envelope most potential sites in the United States. This chapter discusses how the specific interfaces are to be used in the AP600 desigr.. The Combined License applicant is responsible to demonstrate that the selected site meets '

the interface.

For cases where a site characteristic exe is the envelope parameter, it is the responsibility of the Combined License applicant referencing the AP600 to demonstrate that the site characteristic does not exceed the capability of the design. Thus,it is not necessary or appropriate to include in the design cenification of the AP600, requirements and commitments for applicants with sites that do not meet the site characteristics for the standard design. Where the plant design must accommodate site specific issues, these are referenced in the SSAR.

Regarding the specific items which were requested to be assigned for COL action, for each radwaste system, the I information is addressed as follows:

I Liquid Radwaste System  !

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a. The AP600 compliance with 10 CFR 50, Appendix I is discussed in SSAR section i1.2.3.1. The cost-benefit l analysis for population doses via plant effluents after severe accidents is discussed in SSAR Appendix IB. The l l Combined License applicant will provide a site specific cost-benefit analysis to address the requirements of 10 CFR l 50, AprM I, regarding population doses due to liquid effluents as stated in SSAR, Section 11.2.4.2. j

b. Radiation monitor setpoints which terminate liquid waste discha ge are based on limits as defined in 10 CFR 1 l 50, Appendix I as discussed in SSAR section 11.2.3.1. These will be included in the Offsite Dose Calculation l Manual (ODCM) which will be submitted by the combined license applicant as stated in SSAR Section 11.5.7.

l c. Please refer to the SSAR Appendix 1 A for a discussion of compliance to the QA provisions of Regult.:ory l Guide 1.143. The quality assurance program for design, fabrication, procurement, and installation of the liquid l radwaste system is :n accordance with the overall quality assurance program described in SSAR Chapter 17.

Gaseous Radwaste System i

a. The AP600 compliance with 10 CFR 50, Appendix I is discussed in SSAR section 11.3.3.4. The cost-benefit l analysis for population doses via plant effitents after severe accidents is discussed in SSAR Appendix IB. The l Combined License applicant will provide a site specific cost-benefit analysis to address the requirements of 10 CFR l 50, Appendix I, regarding population doses due to gaseous effluents as stated in SSAR, Section 11.3.4.1.

b. Gaseous radiation monitor setpoints which terminate gaseous waste discharge are based on limits as defined in l 10 CFR 50, Appendix I as discussed in SSAR section i1.3.3.4. These will be included in the Offsite Dose l Calculation Max.nal (ODCM) which will be submitted by the combined license applicant as stated in SSAR Section l 1 f.5.7.

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NRC REQUEST FOR ADDITIONAL INFORMATION l

l c. Please refer to the SSAR Appendix 1 A for a discussion of compliance to the QA provisions of Regulatory Guide l 1.143. The quality assurance program for design, fabrication, procurement, and installation of the gaseous radwaste l system is in accordance with the overall quality assurance program described in SSAR Chapter 17.

Solid Waste Management System

a. SSAR section 11.4.1.3 discusses compliance with 10CFR61 and site specific disposal requirements.

l b. Processing of wet solid wastes is discussed in SSAR sections 11.4.2.3.1 and 11.4.2.4.1. The method is to l dewater the spent resin in the a carbon steel container used for shipping and disposal. See the response to RAI l 460.11, Revision 2 for additional detail.

l c. Please refer to the response to RAI 460.5, Revision 2 for a demonstration of the storage facilities for solid radwaste.

d. SSAR section 11.4.1.3 requires compliance with 10CFR71 for packaging of radioactive solid wastes.

l e. Flease refer to the SSAR Appendix I A for a discussion of comoliance to the QA provisions of Regulatory Guide l 1.143. The quality assurance program for design, fabrication, p ocurement, and installation of the solid radwaste l system is in accordance with the overall quality assurance program described in SSAR Chapter 17.

Process and Effluent Monitoring and Sampling Systems

a. Compliance with Regulatory Guide 1.21 is discussed in SSAR sections 11.5.1.2 and i1.5.2.3.3. The radiation monitoring system is supplied and maintained according to a quality assurance program per the requirements of ASME NQA-1-1989 Edition through NQA-lb-1991 Addenda, as outlined in Chapter 17 of the SSAR. This quality program meets the requirements of Reg. Guide 4.15 " Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment". The radiation monitoring system is also specified to be in compliance with ANSI N13.1 " Guide to Sampling Airborne Radioactive Materials".

b. As stated in SSAR sections 9.3.3.1.2.2,11.5.2.3.1 and 11.5.2.3.3, the monitoring and sampling of effluents will be performed according to Reg. Guide 1.97. The 10CFR50 monitoring requirements are reflected in Reg. Guide 1.97.

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NRC REQUEST FOR ADDITIONAL INFORMATION l

T iMponse Revision 1 Question 460.20 There is no systematic discussion of how the liquid, gaseous and solid waste management systems meet each one of the guidelines of RG 1.143. Provide an item-by-item demonstration of compliance with RG 1.143 for a]l of the radwaste management systems.

1 Response: l l

Appendix I A of the SSAR provides a summary of the extent of compliance to the regulatory positions (Section C) l of Reg. Guide 1.143, Rev,1,10D9 " Design Guidance for Radioactive Waste Management Systems, Structures, l l and Components Installed in Light-Water-Cooled Nuclear Power Plants" I

SSAR Revision: None PRA Revision: None l

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Question 460.21 l 1

l Compliance of liquid and gaseous effluent concentrations in unrestricted areas with 10 CFR Part 20 limits is not  ;

based on 1% FF or annual average. Tables 11.2-8 and 11.3-4 of the SSAR should be revised based on 1% FF and I annual average effluent concentrations of radionuclides in unrestricted areas.

Response: ,

l l The determination of compliance of maximum expected liquid and gaseous effluent concentrations in unrestricted l areas with 10 CFR Part 20 limits that is provided in the SSAR is based on the design basis fuel defect level of 0.25 percent and on the annual average atmospheric dispersion factor.

l Altnough the AP600 has as a design basis a maximum fuel defect level of 0.25 percent, Tables 11.2-9 and 11.3-4 l have been revised to include consideration of operation with a maximum defined fuel defect level of 1.0 percent.

The 1.0 percent fuel defect level is a beyond-design-basis condition but is consistent with the guidance of Regulatory Guide 1.70, Revision 3 and with Sections 11.2 and 11.3 of the Standard Review Plan.

I SSAR Revision: NONE l

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Response Revision 1 Question 460.25 The GALE code should be rerun w'th the revised inputs for liquid waste processing and waste gas processing. The GALE output should be checked for secondary concentrations of iodines and other isotopes.

Response

1 The GALE code results from the reanalysis are provided in SSAR Tables 11.2-7,11.2-8,11.2-9,11.3-3 and 11.3-4.

SSAR Revision: NONE 1

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l l NRC REQUEST FOR ADDITIONAL INFORMATION Responso Revision 1 Question 460.26 l

Table 11.1-8 of the SSAR should have an entry for N-16 primary coolant activity. Revise the table accordingly.

Response

Table 11.1-8 of Revision 0 of the SSAR had a listing of N-16 activity. The listing of N-16 activity was removed. l l The GALE code does not con:;ider N-16 because it is not a consideration in regard to effluent releases. Table 11.1-8 has been revised to show coolant activity concentrations calculated by the GALE computer code.

N-16 is a major contributor to the shielding source term for the reactor coolant system. Table 12.2-3 provides N-16 concentrations at various points in the reactor coolant system.

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