ML20046C973
| ML20046C973 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 08/09/1993 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ET-NRC-93-0297, ET-NRC-93-297, NUDOCS 9308130131 | |
| Download: ML20046C973 (22) | |
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Westinghouse Energy Systems Box 355 Pittsburgh Pemsylvania 15230-0355 Electric Corporation ET-NRC.93-3936 NSRA-APSI 93-0297 Docket No.: STN-52-003 August 9,1993 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 A'ITENTION: R.W.BORCHARDT
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600
Dear Mr. Borchardt:
Enclosed are three copies of the Westinghouse responses to NRC requests for additional information ,
on the AP600 from your letter of April 13,1993. This transmittal completes the responses to the April 13,1993 letter. A listing of the NRC requests for additional information responded to in this letter is contained in Attachment A. Attachment B is a complete listing of the questions associated with the April 13,1993 letter and the corresponding Westinghouse letters that provided our response.
In addition to responses to ocr April 13,1993 letter, the enclosure includes revisions of responses previously transmitted.
These sesponses are also provided as electronic files in Wordperfect 5.1 format with Mr. Hasselberg's copy.
If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334.
W 8 g Nicholas J. Liparulo, Manager Nuclear Safety & Regulatory Activities lnja Enclosure cc: B. A. McIntyre - Westinghouse e F. Hasselberg - NRR m,. 130053 9308130131 930809 _6 6004 !
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. ET-NRC-93-3936 ATTACHMENT A -
AP600 RAI RESPONSES SUBMITTED AUGUST 9,1993 RAI No. Issue 471.004 : Rad piping area maps -
471.005 : Transport of equipment to lower radiation areas 471.006 i Location, number, rationale for ARMS 471.014 : Very high radiation areas 471.020 : Cleanup / exposure from ADS 4th stage actuation 720.052R01: Offsite consequences (Resised Response) 720.053R01: Severe accident mitigation design alternatives o
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Prirted: 08/10/93 ATTACHMENT B CROSS REFERENCE OF WESTINGHOUSE RAI RESPONSE TRANSMITTALS TO NRC LETTER OF AUGUST 9,1993 Question issue NRC Westinghouse No. Letter Transmittal Date l
220.021 Wind force & vertical velocey profiles 04/1193 05/1493 i 220.022 Mssile velocities 04/13/93 05/1493 -l 220.023 Compliance with SRP Secton 3 5.3 04/13/93 05/1493 l 471.004 Rad piping area maps 04/1193 O&D9/93 )
471.005 Transport of equipment to lower radiator, areas 04/13 S3 08/09.93 )
471.006 Locatm, number, rationale for ARMS 04'1193 08/09,93 {
471.007 Steam generator manway ease of entry 04/1193 05/2893 l 471.008 Overfow lines into waste collecten system 04/1193 05/2&S3 l 471.009 Floor drain system 04/13/93 07/01 S 3 l 471.010 Lighting in high radiaton areas 04/13/93 07/16/93 j 471.011 Ventilaten during niaintenance a outages 04/1193 07/01/93 ;
471.012 Limitations on cobalt impwity content 04/13G3 05/28/93 l 471.013 Head closure system 04/1193 06/17S3 1 471.014 Very high radiaten areas 04/13/93 08/09 %3 l 471.015 Tank venting to building ventilation system 04/13/93 06/17/93 471.016 Containment monitoring activities 04/1193 05/28/93 ,
471.017 Radiation compartment wall & flaur coatings 04/1193 022&S3 1 471.018 BAS /CAS connections with other systems 04/1193 07/1693 1 471.019 Event / condition requinng additioral shield walls 04/1193 07/1693 1 471.020 Cleanup / exposure from ADS 4th stage actuaton 04/13S3 08/09/93 1 480.004 HWRF Test Data 04/1193 05/28/93 ;
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NRC REQUEST FOR ADDITIONAL INFORMATION iE *i
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Question 471.4 Section 12.1.1.1 of the SSAR states that pipes contaiuing radioactive fleids or radioactive sources are adequately shielded and properly routed to minimize exposure to warsonnel. Indicate and describe on area maps all n.dioactive ;
horizontal pipe chases and areas that contain radioactwe piping that personnel could come in contact with. ,
Response
This is an interim response to the referenced question.
The radioactive pipe chases cxyataining radioactive piping in the nuclear island and the radwaste building and areas that personnel could come in contact with will be identified by a cross-hatched pattern on the mom number drawings.
The final response to the RAI will also pmvide a description of the piping routed in these pipe chases and areas.
The final response to the question, including drawings, will be submitted by October 29,1993 in conjunction with the radiation zone maps.
SSAR Revision: NONE 471.4-1 W-Westinghouse i
NRC REQUEST FOR ADDITIONAL INFORMATION e m.:
h Question 471.5 Sections 12.1.2.2 and 12.1.23 of the SSAR state that facility design considerations (nulularization) to minimize the amount of personnel time spent in a high radiation area include the transportation of equipment or components rquiring service to a lower radiation area. List and describe the location of all major cx>mponents and equipment that art dr *gned to be transported to a lower radiation area during servicing.
Response
he motor / impeller assemblies of the four reactor coolant pumps (RCP) represent the largest components specifically designed to be transponed to a lower-radiation area during servicing. A specially designed pump cart /maintenanoe stand is used to remove the RCP motor / impeller assemblies from me RCP casings. After an RCP assembly is removed from the RCP casing, both the RCP and the pump cart are sin;@ancously lifted from the steam generator compartment by the polar crane and set down on the operating deck adjacmt to the main equipment hatch. The pump cart transports the RCP the short distance from the containment to the arecx 11 building for maintenance.
ihe annex 11 building provides large areas immediately outside the containment at the operating deck level, for taajor equipment assembly / disassembly, decontamination, testing, and inspection. His ckwe coupling of the annex 11 building to the nuclear island allows many maintenance activities to be performed close % the containment and the radiologically controlled areas of the auxiliary building.
The AP600 containment contains a 22-foot-diameter main equipment hatch and a personnel airlock at the operating deck level and a 16-foot-diameter maintenance hatch and a personnel airlock at grade level. These two containment hatches significantly enhance accessibility to the containment during outages and reduce the potential for congestion at the containment entrances. These containment hatches, loalted at the two different levels, allow activities oxurring atsove the operating deck to be unaffected by activities occurring below the operating deck.
Pathways throughout the plant are sized to accommodate equipment maintenarice 9d removal from any area within the plant. A majority of the equipment transported to the annex 11 buildiry f(x maintenaccr are subassemblics of larger components. Examples include valves and valve components, purg motors and internals, instrumentation, small tanks and heat exchangers, IIVAC fans and motors, and other corr pone '"ed in the radiokigically controlled areas of the AP600 nuclear island, the radwaste building, and the r ,M Ni 1 SSAR Revision: NONE i
471.5-1 W Westinghouse l
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NRC REQUEST FOR ADDITIONAL INFORMATION
%! tilii II '!il Question 471.6 Section 11.5.6.1 of the SSAR describes the design objectives of area mdiation monitors (ARMS) and Table 11.5-2 of the SSAR describes the locations of the ARMS throughout the plant. Describe the rational for the kications and ,
numbers (which seem low compared with industry nonns) of the ARMS that are required. Also, compare this rational with Criteria 4.2 (detector k> cations) of ANSI /ANS Standard HPSSC-6.8.1-1981 (Location and Design Criteria for Area Radiation Monitoring Systems for Light Water Nuclear Reactors), and justify any deviations.
Response
ANSI /ANS Standard HPSSC-6.8.1-1981 is not a regulatory requirement. However, the AP600 ARM location criteria are derived from ANSI /ANS Standard HPSSC-6.8.1-1981. The criteria are as follows:
Area monitors shall be located in areas that are normally accessible and where changes in normal plant operating conditions can cause significimt increases in exposure rates above those expected for the areas.
[HPSSC-6.8.1, 6 4.2.1]
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Area monitors shall be k>cated in areas that are normally or occasionally accessible where significant increases in exposure rates might occur because of operational transients or maintenance activities. [HPSSC-6.8.1, f 4.2.2]
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Area monitors shall be kicated to best measure the increase in exposure rates within a specific area and to avoid shielding of the detector by equipment or stmetural materials. [HPSSC-6.8.1, s 4.23]
In the selection of area monitors, consideration shall be given to the environmental conditions under which the monitor will be operating in order to ensure proper monitor operation. [HPSSC-6.8.1, f 4.2.41 Area monitors shall be located to provide access so that minimal maintenance equipment is required and to ,
provide an uncluttered area near the detector and local pmcessing electronics to allow for field alignment and calibration. [HPSSC-6.8.1, f 4.2.5)
A comparison of the monitors currently listed in SSAR Table 11.5-2 with those listed in ANSI /ANS Standard HPSSC-6.8.1-1981 follows, with an exphination for all differences and changes due to design evolution.
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. . l NRC REQUEST FOR ADDITIONAL INFORMATION sa 'un t-AIWX) SS AR IIPS SC-6.8.1 Comments Table 11.5-2 Table 2 Post Accident Sarnple Room Primary Sampling Station Area Containment Area Containment - Personnel Hatch Operating deck @ elevation 135' Arca Containment - Refueling Platform See Note 1 Containnient - Incore Instrument AIWX) has an integrated reactor Area vessel head package that includes the incore instruments (top entry) therefore, there is no incore scal table area that would need area monitoring.
Main Control Room Control Room Only one ARM is necessary in (3 Monitors) the main control room.
Chemistry Laboratory Arca Radiochemistry Laboratory Fuel Handling Area Fuel Storage Arca See Note 2 Cask Handling Area An ARM will be located in the rail car hay area.
Liquid Radwaste Area (Not identified) Common with gaseous radwaste monitor Gaseous Radwaste Area (Not identified) Common with liquid radwaste monitor Solid Radwaste Area Radwaste Conuol Panel Arca One ARM is provided in the radwaste contrul room.
Technical Support Center Area (Not identified)
Radwaste Storage Area Solid Radwaste Storage Area One ARM is provided in the storage area corridor.
Hot Machine Shop An ARM will be krated in this area. I HVAC Filter Arca Not provided - there is no rapid increase in activity anticipated in this area.
W Westinghouse i
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NRC REQUEST FOR ADDITIONAL INFORMATION AP600 SSAR HPSSC-6.8.1 Comm,mts Table 11.5-2 Table 2 RHR Pump Area Not provided - This area is under access control.
RHR IIcat Exchanger Area Not provided - his area is under access control.
Drumming Station Control Panel Not provided - His area is under Area access control.
Equipment Decontamination Areas Not provided - These areas are under access control.
ReFenerative Chemical Waste Not provided - Regeneration is Evaporator Area perfonned offsite.
Upon funher evaluation, an ARM will be heated in the annex 11 staging and storage area.
NOTES: 1. Radiation levels shall be monitored by the permanent containment area radiation monitor and by a portable bridge monitor during refueling operations. The containment area radiation monitor shall be h>cated to best measure the increase in exposure rates for this area and to provide an alarm krally and in the main control room.
2.
Radiation levels shall be monitored by the pennanent fuel handling area radiation monitor and by a portable bridge monitor during refueling operations. De fuel handling area radiation monitor shall be kicated to best measure the increase in exposure rates for this area and to provide an alarm heally and in the main control rootn.
NRC REQUEST FOR ADDITIONAL INFORMATION g.
Subsection 11.5.6.1 will be revised as follows:
Replace the last portion of this section relating to location critena with the following:
Location of area radiation monitors is based on the following criteria-Area monitors shall be located in areas that are nonnally accessible and where changes in nonnal plant operating conditions can cause significant increases in exposure rates above those expected for the areas.
Area monitors shall be located in areas that are nonnally or occasionally accessible where significant increases in exposure rates might occur because of operational transients or maintenance activities.
Area monitors shall be located to best measure the increase in exposure rates within a specific area and to avoid shielding of the detector by equipment or structumi materials. ;
In the selection of area monitors, consideration shall be given to the environmental conditions under which the monitor will be operating in order to ensure proper monitor operation.
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Area monitors shall be located to provide access so that minimal maintenance equipment is required and to provide an uncluttered area near the detector and kwal processing electronics to allow for field alignment and calibration.
The area nadiation monitors are listed in Table 11.5-2. 2 j
Replace Subsections 11.5.6.2 through 11.5.6.11 with the following: J I
I1.5.6.2 Post Accident Sampic Room Monitor i
The post acci lent sample station is where samples are collected and/or analyzed after the ptulated accident. The I post accident sampla room area radiation monitor is located so that its readout is representative of the radiation to l
which the operating persormel are exposed. A kwal readout and alarm module is kicated so that it is visible to the l operating personnel. The radiation data is displayed and stored at the associated local radiation processor and .l transmitted to the main control room. 1 The monitor detector is a gamma-sensitive Geiger-Mueller tube. The monitor is an extended-range monitor that ;
meets the regulatory position of Regulatory Guide 1.97. The monitor range and principal isotopes are listed in Table 11.5-2.
471.6-4 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION iie~ "e T jil 11.5.6.3 Nomial-Range Area Monitors Nonnal-range area radiation monitors are located in accordance with the h> cation criteria given in Subsection l 11.5.6.1. A local readout and alann module is located in each area so that it is visible to the operating personnel in that area. The radiation data is displayed and stored in the monitors' associated kwal radiation pmcessor and transmitted to the main control room.
The monitor detectors are gamma-sensitive Geiger-Mueller tubes. The monitors, their ranges, and principal isotopes are listed in Table 11.5-2.
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NRC REQUEST FOR ADDITIONAL INFORMATION r .m Replace Table 11.5-2 with the following:
TaNe 11.5-2 Area Radiation Monitor Detector Parameters Detector Type Ser ice Isotones Nominal Rance RMS-JE-RE008 y Post Accident Sample Room Cs-137 1.0E-1 to 1.0E+7 mR/hr RMS-JE-RE009 y Containment Area - Personnel IIatch Cs-137 1.0E-? to 1.0E+4 mR/hr (Note 1)
RMS-JE-RE010 y Main Control Room Cs-137 1.0E-1 to 1.0E+4 mR/hr RMS JE-RE011 y Chemistry Laboratory Area Cs-137 1.0E-1 to 1.0E+4 mR/hr RMS-JE-RE012 y Fuel Handling Area Cs-137 1.0E-1 to 1.0E+4 mR/hr (Note 2)
RMS-JE-RE013 y Rail Car Bay Area Cs-137 1.0E-1 to 1.0E+4 mR/hr RMS-JE-RE014 y Liquid and Gaseous Radwaste Area Cs-137 1.0E-1 to 1.0E+4 mR/hr RMS-JE-RE015 y Radwaste Control Room Cs-137 1.0E-1 to 1.0E+4 mR/hr RMS-JE-RE016 y Technical Support Center Cs-137 1.0E-1 to 1.0E+4 mR/hr RMS-JE-RE017 y Radwaste Storage Corridor Cs-137 1.0E-1 to 1.0E+4 mR/hr RMS-JE-RE018 y liot Machine Shop Cs-137 1.0E-1 to 1.0E+4 mR/hr RMS-JE-RE019 y Annex 11 Staging & Storage Area Cs-137 1.0E-1 to 1.0E+4 mR/hr NOTES: 1. Radiation levels shall be monitored by the permanent containment area radiation monkor and by a portable bridge monitor during refueling operations. The containment area radiation monitor shall be located to best measure the increase in exposure rates for this area and to provide an alarm locally and in the main control room.
- 2. Radiation levels shall be monitored by the permanent fuel handling area radiation monitor and by a portable bridge monitor during refueling operations. 'Ihe fuel handling area radiation monitor shall be located to best measure the increase in exposure rates for this area and to provide an alarm locally and in the main control room.
471.6-6 3 We5Tingh0Use
I NRC REQUEST FOR ADDITIONALINFORMATION ar" Rin si Question 471.14 Section 123.1.2 of the SSAR states that Radiation Areas and Iligh Radiation Areas will be posted and mntrolled. t No mention is made of Very liigh Radiation Areas as defined in 10 CFR 20.1602. Provide the location of any areas where personnel could be exposed to radiation levels greater than 100 Rads in one hour as stated in the Standard Review aan, r.nd describe any special controls to prevent personnel entry in to these areas.
Also, Section 12.5.4 of the SSAR staics the entrances to liigh Radiation Areas are equipped with audible and/or visible alarms. Clarify this statement. Will these areas also be koked? Discuss how the requirements of 10 CFR 20.1601, " Control of Access to IIigh Radiation Arcas," and 10 CFR 20.1602, " Control of Access to Very Ifigh Radiation Areas," and the guidance of Regulatory Guide 838 (issued in May 1993) will be implemented.
Response
'Ihis is an interim response to the referenced question.
Location of high and very high radiation areas and the type of access mntrols to these areas will be identified for the nuclear island and radwaste building. The response will include a discussion on how the requirements from 10 CFR 20.1601,10 CFR 20.1602, and Regulatory Guide 838 are impicmented. Expected dose rates will be used to determine tbc type of acass control.
A final response to the question will be submitted by Oct(ber 29,1993.
SSAR Revision NONE r
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NRC REQUEST FOR ADDITIONAL INFORMATION um "l'n t .. g .
Question 471.20 Section 6.3.2.2.7.6 of the SS AR descrees the operation of the Automatic Depressurization System (ADS) valves such that the fourth stage valves vent directly into each steam generator comparunent on actuation. Provide the expected f requency at which these valves will open. Describe the expected cleanup operations that would be expected after an imidvertent opening of these valves. Provide the expected person. Rem a licensee would receise to cleanup after an inadvertent opening of these valves and the length of time it would take to complete the cleanup.
Response
Several AP600 design features and operating characteristics make the opening of the fourth-stage ADS vahes scry unlikely. These features and characteristics include the following:
- The SS AR chapter 15 analysis shows diat ADS operation does not occur during non-LOCA events or ste2un generator tube ruptures. h is indicated to occur only during an inadvertent ADS event and during LOCAs.
. The ADS is actuated with 2 out of 4 logic. As a result. multiple instrumentation failures are required m order to inadvertently actuate ADS.
Each ADS line has two normally closed valves in series. As a result, multiple valve failures are necessary in order to inadvenently actuate ADS. In addition, the fourth-stage ADS vakes and operators are designed so that they cannot open a: normal RCS pressures.
If ADS is actuated inadvenently or for a very small RCS LOCA, the fourth stage will not be actuated, assuming that the operators stan the normal residuid heat removal pumps. Injection fmn. these pumps increans the backpressure on the CMTs and causes their injection to stop with the CMT level above the fourdestage actuation setpoint.
If the normal residual heat ternoud system is unavailable in an inadvertent ADS esent and the ADS founidstage valves open, the containment will slowly flomi to its maximum level over several days.
Recovery of the nonnal residual heat removal system during this time allows the floodup to be tenninated by closing the ADS fourth-stage valves.
An estimate of the frequency of inadvertent ADS has been made for the AIWK). The results indicate that the frequency ofinadvertent ADS actuation is approximately 2E-3/yr. This frequency is similar to the frequency of very small and small LOCAs used in the AP600 PRA (1.07E-3/yr). As discussed in item 4, the normal residual heat removal system, actuated by the operators, will prevent the fourth-stage vahes from opening during inadvertent ADS events. Th3 frequency of opening the fourtidstage ADS valves will therefore be less than the frequency of ADS actuation. The frequency of opening the fourth-stage ADS valves is estimated to be about 2E-4/yr. The largest l contnbutri to this frequency is a break of the direct vessel injection line (1.2E-4/yr).
NRC REQUEST FOR ADDITIONAL INFORMATION
- w W tr gj The cleanup operations and times and the person-rem exposures from an inadvertent opening of the founh-stage ADS valves are dependent on the specific sequence of events. The AP600 incorporates features to facilitate the cleanup and restart of the plant following ADS actuation. These features are aimed at reducing the consequences of a parthd ADS (no fourth stage) because it is more likely. These features include the following:
- Decontaminable surfaces of the reactor cavity up to the reactor vessel bottom level.
. The majority of electrically operated equipment need for cleanup is located above thxxl level. The containment sump pumps and reactor coolant drain tank and pumps are designed for submersion.
These AP600 features and ch:urteristics reduce the probability of opening the fourth-stage ADS valves and therefore the consequences.
SS AR Resision: NONE 471.20-2 3 Westinghouse
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NRC REQUEST FOR ADDITIONALINFORMATION init nin I 2 m Response Revision 1 Question 720.52 De calculation of offsite dose is said to be based on a 24-hour exposure following the initiation of core damage, assuming no protective actions. Sina, for the release classes modelled, gross containment failure does not occur until after this 24-hour window (if at all), initiation of the 24-hour window at the time of core damage does not appear to represent the most severe dose. (In effect, the calculation only reflects normal containment leakage.)
Consistent with the assumption that no protective actions are taken, an additional assessment should be made of the 24-hour dose that would result if the 24-hour exposure period was initiated: (1) at the time of significant fission product release to the containment (e.g., about 12-15 houts for relcase class "OK"), or (2) at the time of containment failure.
Response
De results presented in the PRA are for an exposure 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> afler core damage (defined in this case as onset of wre melting), which is coincidental with the time of significant fission product release to the containment (often referred to as the " melt release." See figure le22, which shows the fraction of Csl retained in the RCS for the BC1 case). Dis satisfics the exposure period requested in (1). This same exposure interval convention applies to all of the release category cases that were analyzed. De MAAP4 code was run so that the source term was calculated and output for the time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af!cr the core begins melting and releasing fission products to the containment.
%c MACCS code was run over the 24-hour duration of the source term that was input from the MAAP4 calculation.
In the OK and OKP cases presented in the PRA, the plant has reached a safe, stable state. The analysis was terminated at the transient time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage. No further events (i.e., containment failure) will occur that will affect the fission product release and the offsite dose.
He CC release catt gory represents sequences in which the debris is not coolable the basemat is croding, and the containment is being pressurized with noncondensable gases. De analysis and position paper (WCAP-13388) demonstrate that no containment failure is expected for over 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the release of fission products to the mntainment. He dose presented in the PRA represents the dose expected to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without any shcitering or evacuation. De additional time allows for accident management to prevent containment faifurc.
De Cl release utegory has a mntainment that is open to the environment at the initiation of the fission product release to the containment. De offsite dose measured over a 24-hour duration after the fission product release begins is significant, but the frequency of the CI release category (3 x 10* per reactor-year) is very low.
Derefore, the dose results presented in the PRA are representative of the most severe dose expected for caah of the release categorics.
PRA Revision: NONE 720.52(R1)-1 W-Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION 9 nm Response Revision 1 Question 720.53 Additional offsite consequence measures for the AP600 will be needed to permit comparisons to the Commission's quantitative health safety goals, and the assessment of severe accident mitigation design alternatives. Specific measures that will be required for the AP600 are the individual risk of early fatality, individual risk of cancer fatality, the probability of large release, and societal risk in terms of person-rems for a representative site.
Response
Tabic 720.53-1 provides the release category frequency and societal risk for the AP600 plant in terms of person-rems.
'Ihis table was taken directly from the AP600 SAMDA report, which was provided under s':parate cover (See Table IB.6-1, ref. 720.53-1). The large-release frequency is equal to the frequency of the CI :elease category.
Tables 720.53-2 through 720.53-5 provide the individual risks of early fatality and cancer fatality. Each of those risks was calculated using the source terms from the AP600 PRA and the ALWR reference site data (ref 720.53-2) as input to the MACCS code version 1.5.
PRA Revision: NONE 720.53(R1)-1
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NRC REQUEST FOR ADDITIONAL INFORMATION
- mt .y Response Revision 1 Table 720.53-1 AP600 llase Rhk (Whole-Population Dose to a 50-Mile Radius)
Release Frequency Mean Consequence Rhk Category (yr) (person rem) (person-rem-yr-3)
OK 2.5E-7 6.93 1.73E-6 OKP 5.6E-8 13.4 7.50E-7 CC 7.6E-10 9.01 6.85E-9 C1 3.0E-8 114000 3.42E-3 3.42E-3 720.53(RI)-2 3 Westinghouse
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NRC REQUEST FOR ADDITIONALINFORMATION
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Response Revision 1 Table 720.53-2 RC.OK Source Term Averace Individual Risk Distance Mean Median Early Fatality 0.0-0.4 km 0.00E+00 0.00E+00 Early Fatality 0.4-1.2 km 0.00E+00 0.00E400 Early Fatality 1.2-1.6 km 0.00E+00 0.00E+00 Early Fatality 1.6-3.2 km 0.00E+00 0.00E+00 -
Early Fatality 3.2-4.8 km 0.00E+00 0.00E+00 Cancer Fatality 0-0.4 km 338E-05 3.13E-05 Cancer Fatality 0.4-1.2 km 4.40E-06 4.07E-06 Cancer Fatality 1.2-1.6 km 1.58E-06 139E-06 Cancer Fatality 1.6-3.2 km 5 81E-07 5.58E-07 Cancer Fatality 3.2-4.8 km 1.)8E-07 1.93E-07 Cancer Fatality 4.8-6.4 km 9.71E-08 9.25E-08 Cancer Fatality 6.4-8.1 km 5.53E-08 535E-08 Cancer Fatality 8.1-9.7 km 3.53E-08 3.26E-08 Cancer Fatality 9.7-113 km 2.43E-08 2.22E-08 Cancer Fatality 113-12.9 km 1.74E-03 IIX)E-08 Cancer Fatality 12.9-14.5 km 137E-08 1.21E4)8 Cancer Fatality 14.516.1 km 1.10E-08 1.02E-08 Cancer Fatality 16.1-17.7 km 8.78E-09 8.42E-09 Cancer Fatality 17.7-19.3 km 7.14E-09 7.06E-09 Cancer Fatality 193-20.9 km 5.96E-09 5.85E-09 Cancer Fatality 20.9-22.5 km 5.08E-09 4.84E-09 Cancer Fatality 22.5-24.1 km 4.50E-09 4.09E-09 Cancer Fatality 24.1-25.8 km 3.94E-09 3.53E-09 Cancer Fatality 25.8-27.4 km 3.54E-09 3.22E-09 Cancer Fatality 27.4-29.0 km 331E-09 2.99E-09 Cancer Fatality 29.0-30.6 km 2.91E-09 2.73E-09 Cancer Fatality 30.6-32.2 km 2.67E-09 2.51E-09 Cancer Fatality 32.2-483 km 1A9E-09 1.28E-09 Cancer Fatality 483-64.4 km 7.13E-10 7.24 E-10 Cancer Fatality 64.4-80.5 km 4.40E-10 4.75E-10 W Westinghouse 20.53m3
NRC REQUEST FOR ADDITIONAL INFORMATION mn mt:
1 E Rosponse Revision 1 Tabk 720.53-3 '
OKP REVISED SOURCE TERM Averace Imlividual Risk Distance Mean Median Early Fatality 0.0-0.4 km 0.00E+00 0.00E+00 Early Fatality 0.4-1.2 km 0.00E+00 0.00E+00 Early Fatality 1.2-1.6 km 0.00E+00 0.00E+00 Early Fatality 1.6-3.2 km 0.00E+00 0.00E+00 Early Fatality 3.2-4.8 km 0.00E+00 0.00E+00 Cancer Fatality 0-0.4 km 6.54E-05 634E-05 Cancer Fatality 0.4-1.2 km 8.40E-06 8.21 E-06 Cancer Fatality 1.2-1.6 km 3.09E4 3.02E-06 Cancer Fatality 1.6-3.2 km 1.14E-06 1.03E-06 Cancer Fatality 3.2-4.8 km 4.04E-07 3.66E-07 Cancer Fatality 4.8-6.4 km 1.98E-07 1.71E-07 Cance 7atality 6.4-8.1 km 1.12E-07 1.02E-07 Cancer Fatality 8.1-9.7 km 7.20E-08 7.09E-08 Cancer Fatality 9.7-113 km 5.09E-08 5.08E4)8 Cancer Fatality 113-12.9 km 3.82E-08 3.43E-08 Cancer Fatality 12.9-14.5 km 2.93E-08 3.02E-08 Cancer Fatality 14.5-16.1 km 2.29E-08 2.13E-08 Cancer Fatality 16.1-17.7 km 1.94E-08 1.72E-08 Cancer Fatality 17.7-193 km 1.60E-08 1.24E-08 Cancer Fatality 193-20.9 km 133E-08 1.13E-08 Cancer Fatality 20.9-22.5 km 1.09E-08 930E-09 Cancer Fatality 22.5-24.1 km 9.14E-09 8.69E-09 Cancer Fatality 24.1-25.8 km 7.92E-09 734E-09 -
Cancer Fatality 25.8-27.4 km 6.70E-09 633E-09 Cancer Fatality 27.4-29.0 km S M E-09 5.64E-09 Cancer Fatality 29.0-30.6 km 5.03E-09 5.17E-09 Cancer Fatality 30.6-32.2 km 4.58E-09 4.19E-09 ;
Cancer Fatality 32.2-483 km 2.6]E-09 2.53E-09 Cancer Fatality 483-64.4 krn 1.42E-09 1.23E-09 Cancer Fatality 64.4-80.5 km 9.06E-10 830E-10 72 u sca w 4 W westingtiouse
k' NRC REQUEST FOR ADDITIONALINFORMATION Response Revision 1 Table 720.53-4 CI SOURCE TEllM Averace Individual Risk Distance Mean Median Early Fatality 0.0-0.4 km 5.70E-02 7.06E-02 Early Fatality 0.4-1.2 km 1.19E-03 0.00E+00 Early Fatality 1.2-1.6 km 7.72E-05 0.00E+00 Early Fatality 1.6-3.2 km 131E-07 0.00E+00 Early Fatality 3.2-4.8 km 0.00E+00 0.00E+00 l Cancer Fatality 0-0.4 km 7.46E-02 7.06E-02 l Cancer Fatality 0.4-1.2 km 2.5SE-02 2.24E-02 l
Cancer Fatality 1.2-1.6 km 1.46E-02 1.44E-02 Cancer Fatality 1.6-3.2 km 7.60E-03 738E-03 Cancer Fatality 3.2-4.8 km 3.66E-03 3.05E-03 Cancer Fatality 4.8-6.4 km 2.02E-03 138E-03 Cancer Fatality 6.4-8.1 km 1.19E-03 8.96E-04 Cancer Fatality 8.1-9.7 km 7.50E-04 5.78E-04 Cancer Fatality 9.7-113 km 539E-04 3.93E-04 Cancer Fatality 113-12.9 km 3.91E-04 3.24E-04 Cancer Fatality 12.9-14.5 km 2.77E-04 2.4SE-04 Cancer Fatality 14.5-16.1 km 2.01E-04 1.60E-04 Cancer Fatality 16.1-17.7 km 1.55E-04 1.11E-04 Cancer Fatality 17.7-193 km 1.26E-04 1.09E-04 Cancer Fatality 193-20.9 km 1.03E-04 8.50E-05 !
i Cancer Fatality 20.9-22.5 km 8.76E-05 6.45E-05 Cancer Fatality 22.5-24.1 km 7.56E-05 5.51E-05 Cancer Fatality 24.1-25E km 6.65E-05 5.05E-05 Cancer Fatality 25.8-27.4 km 5.91E-05 4.24E-05 Cancer Fatality 27.4-29.0 km 5.22E-05 3.83E-05 Cancer Fatality 29.0-30.6 km 4.53E-05 3.43E-05 ;
Caneer Fatality 30.6-32.2 km 3.80E-05 3.16E-05 i Cancer Fatality 32.2-483 km 2.43E-05 1.95E-05 Cancer Fatality 483-64.4 km 132E-05 1.06E-05 {
Cancer Fatality 64.4-80.5 km 7.84E-06 5.99E-06 l 20.53ps W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Table 720.53 5 CC SOURCE TERM Average Indivklual Risk Distance Mean Median Early Fatality 0.0-0.4 km 0.00E+00 0.00E+00 Early Fatality 0.4-1.2 km 0.00E+00 0.00E+00 Early Fatality 1.2-1.6 km 0.00E+00 0.00E+00 Early Fatality 1.6-3.2 km 0.00E+00 0.00E+00 Early Fatality 3.2-4.8 km 0.00E+00 0.00E+00 Cancer Fatality 0-0.4 km 430E-05 3.70E-05 Cancer Fatality 0.4-1.2 km 5.57E-06 5.24E-06 Cancer Fatality 1.2-1.6 km 2.00E-06 1.92E-06 Cancer Fatality 1.6-3.2 km 7.28E-07 737E-07 Cancer Fatality 3.2-4.8 km 2.55E-07 233E-07 Cancer Fatality 4.8-6.4 km 1.23E-07 1.08E-07 Cancer Fatality 6.4-8.i km 7.0SE-08 6.81E-08 l
Cancer Fatality 8.1-9.7 km 4.55E-08 430E-08 Cancer Fatality 9.7-11.3 km 3.12E-08 3.03E-08 Cancer Fatality 113-12.9 km 2.25E-08 2.11E-08 Cancer Fatality 12.9-14.5 km 1.72E-08 1.52E-08 Cancer Fatality 14.5-16.1 km 137E-08 1.20E-08 )
Cancer Fatality 16.1-17.7 km 1.14E-08 1.04E-08 Cancer Fatality 17.7-193 km 937E-09 8.43E-09 Cancer Fatality 193-20.9 km 7.76E-09 7.10E-09 Cancer Fatality 20.9-22.5 km 6.53E-09 5.89E-09 Cancer Fatality 22.5-24.1 km 5.75E-09 5.22E-09 Cancer Fatality 24.1-25.8 km 5.08E-09 4.46E-09 l Cancer Fatality 25.8-27.4 km 4.57E-09 4.28E-09 !
Cancer Fatality 27.4-29.0 km 4.16E-09 3.87E-09 Cancer Fatality 29.0-30.6 km 3.69E-09 3.40E-09 Cancer Fatality 30.6-32.2 km 3.42E-09 3.23E-09 ,
Cancer Fatality 32.2-483 km 1.94E-09 2.02E-09 Cancer Fatality 483-64.4 km 9.45E-10 9.52E-10 Cancer Fatality 64.4-80.5 km 5.78E-10 5.72E-10 720.53(R1) 6 T Westinghouse
NRC REQUF.ST FOR ADDITIONAL INFORMATION uni Response Revision 1 II References 720.53-1 AP600 Severe Accident Mitigation Design Alternatives Report. See response to RAI 100.2.
720.53-2 Advanced Light Water Reactor Requirements Document, Volume 111, Appendix A to Chapter 1, "PRA Key Assumptions and Ground Rules," EPRI, Rev. 2, December 1991.
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