ML20203C301

Forwards RAI Re Open Items on AP600.FSER Identifies Three Open Items Needing Resolution by W Before Staff Can Complete Review of SSAR Section

ML20203C301
Person / Time
Site:	05200003
Issue date:	11/06/1997
From:	Quay T NRC (Affiliation Not Assigned)
To:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 9712150274
Download: ML20203C301 (12)

Text

--

November 6,'1997 Mr. Nicholas J. Liparulo, Manager' Nuclear Safety and Regulatory Activities Noclear and Advanced Technology Analysis Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA -15230

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RELATED TO OPEN ITEMS ON THE -

AP600

Dear Mr. Liparulo:

As a result of its review of the June 1992 application for design certification of the AP600, the staff has prepared Section 15.3 of the final safety evaluation report (FSER) on the AP600. The FSER identifies 3 open items needing resolution by Westinghouse before the staff can complete its review of this SSAR section. These open items should be resolved in conjunction with the -

resolution of Q470.41F, which was forwarded to you on October 7,1997. The open items are identified in the enclosure with tracking numbers associated with them.

You have requested that portions of the information submitted in the June 1992 application for design certification be exempt from mandatory public disclosure. While the staff has not completed its review of your request in accordance with the requirements of 10 CFR 2.790, that-portion of the submitted information is being withheld from public disclosure pending the staffs final determination. The staff concludes that the enclosure does not contain those portions of the information for which exemption is sought.- However, the staff will withhold this letter from public disclosure for 30 calendar days from the date of this letter to allow Westinghouse the opportunity to verify the staffs conclusions, if, after that time, you do not request that all or portions of the information 'n the enclosures be withheld from public disclosure in accordance with 10 CFR 2,790, this.atter will be placed in the NRC's Public Document Room.

if you have any questions regarding this matter, you can contact the Projecf Manager, Thomas J. Kenyon, at (301) 415-1120.

Sincerely,

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Theodore R. Quay, Director Standardization Project Directorate

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Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003

Enclosure:

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e-9 Mr. Nicholas J. Uparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc:

Mr. B. A. McIntyre Advanced Plant Safety & Licensing Ms. Cindy L. Haag Advanced Plant Safety & Licensing Westinghouse Electric Corporation Westinghouse Electric Corporation Energy Systems Business Unit Energy Systems Business Unit P.O. Box 355 Box 355 Pittsburgh, PA 15230 Pittsburgh, PA 157.30 Enclosure to be distributed to the following addressees after the result of the, proprietary evaluation is received from Westinghouse:

Mr. Russ Bell Ms. Lynn Connor Senior Project Manager, Programs DOC-Search Associates Nuclear Energy Institute Post Office Box 34 1776 i Street, NW Cabin John, MD 20818 Suite 300 Washington, DC 20006-3706 Mr. Robert H. Buchholz GE Nuclear Energy Dr. Craig D. Sawyer, Manager 175 Curtner Avenue, MC-781 Advanced Reactor Programs San Jose, CA 95125

'GE Nuclear Energy 175 Curtner Avenue, MC-754 Mr. Sterling Franks San Jose, CA 95125 U.S. Department of Energy NE-50 Barton Z. Cowan, Esq.

19901 Germantown Road Eckert Seamans Cherin & Mellott Germantown, MD 20374 600 Grant Street 42nd Floor Pittsburgh, PA 15219 Mr. Charles Thompson, Nuclear Engineer AP600 Certification Mr. Frank A. Ross NE 50 U.S. Department of Energy, NE-42 19901 Germantown Road Office of LWR Safety and Technology Germantown, MD 20874 19901 Germantown Road Germantown, MD 20874 Mr. Ed Rodwell, Manager PWR Design Certification Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303

. - ~ -. - -..-- -. ~. -. -

OPEN ITEMS ON THE AP600 RADIOLOGIOAL CONSEQUENCES OF DESIGN BASIS ACCIDENTS l

In Chapter 15 of the SSAR, Westinghouse performed radiological consequence assessments of the following seven reactor design basis accidents (DBAs) using the bounding set of atmospheric relative concentration (dispersion) values (or X/Q values) provided in Table 15A 5 of Appendix 15A to the SSAR. These X/Q values determine the required minimum distances to the exclusion area boundary (EAB) and the low-population zone (LPZ) for a given site in order to provide reasonable assurance that the radiological consequences of a DBA will be within the dose limits specified in 10 CFR 50.34. The analyzed DBAs are:

(1)- Main steamline failure outside containment (SSAR Section 15.1.5)

(2)

Reactor coolant pump shaft seizure (locked rotor) (SSAR Section 15.3.3)

(3)

Control element assembly ejection (SSAR Section 15.4.8)

(4)

Failurs of smalllines carrying primary coolant outside containment (SSAR Section 15.6.2)

(5)

Steam generator tube rupture (SSAR Section 15 A.3)

(6)

Loss-of-coolant accident (SSAR Section 15.6.5), and (7)- Fuel handling accident (SSAR Section 15.7.4)

- In Chapter 15 of the SSAR, Westinghouse concluded that the AP600 design will provide reason-able assurance that the radiological consequences resulting from any of the above DBAs will be within the dose criteria specified in 10 CFR 50.34 (25 rem TEDE) and control room operator dose crite'ia specified in GDC ig in Appendix A 1o.10 CFR Part 50 (as applied to the AP600 design:

5 rem TEDE). Westinghouse reached this conclusion (1) using the reactor accident source terms provided in NUREG 1465, " Accident source Terms for Light Water Nuclear Power Plants" with two exceptions (release timing of gap activities and release fractions of low-volatils radionuclides), (2) relying on natural deposition of fission-product aerosol within the containment, (3) controlling the pH of the water in the containment to prevent iodine evolution, and (4) using a bounding set of hypothetical X/Q values.

The x/Q values are the relative atmospheric concentrations of radiological releases at the receptor point in terms of the rate of radioactivity release. In lieu of site-specific meteorological data, Westinghouse provided a bounding set of X/Q values for the AP600 design using the meteorological data which is representative of an 80 to goth percentile of U.S. operating nuclear power plant sites. Should the resolution of the open issues discussed in this enclosure continue to show that calculated radiological consequences exceed the dose limits, then the X/Q values (see Q470.41F, dated October 7, igg 7) and/or the AP600 containment design leak rate may need to be reevaluated.

Westinghouse states that site specific meteorology information would be provided by the COL applicant and that if the site specific meteorolom; parameters exceed the bounding X/Q values provided i.1 the AP600 SSAR Table 15A-5 ' ie COL applicant will address how the radiological Enclosure f

2 consequences resulting from the design-basis accidents continue to meet the dose limits given in

.10 CFR 50.34 and control room operator dose limits given in GDC 19 (as applied to the AP600 2

design: 5 rem TEDE) using site-specific X/Q values.

ACCIDENT SOURCE TERME in SECY 94-302, " Source Term-Related Technical and Licensing issues Relating to Evolutionary and Passive Light Water Reactor Designs," dated December 19,1994, the staff proposed to use only the " coolant,"" gap, and "early in vessel" releases from NUREG-1465 for the radiological consequence assessments of DBAs for the passive ALWR designs. These source terms encompass a broad ranga of accident scenarios, including significant levels of core damage with the core remaining in Sa vessel. These would be the most severe scenarios from which the plant could be expected to retum to a safe shutdown condition.

The revised source terms in NUREG-1465 are to be applied conservatively in evaluating design basis accidents in conjunction with conservative assumptions in calculating doses, such as adverse meteorology. Application to severe accidents may use more realistic assumptions. The staff considered the inclusion of the "ex-vessel" and the " late in-vessel" source terms to be unduly conservative for DBA purposes. Such releases would only result from core damage accidents with vessel failure and core-concrete interactions. For passive ALWRs, the estimated frequencies of such scenarios are low enough that they need not be considered credible for the purpose of meeting 10 CFR 50.34. The Commission approved the staff recommended technical positions to use only the coolant, gap, and the early in-vessel releases from NUREG-1465 for the radiological consequence assessments of DBAs for the passive At.WR designs, in its evaluation of the radiological consequences of accidents for the AP600 design, Westing-house has taken the following two departures from the NUREG-1465 source term:

470.42F - Low-Volatile Fission Product Release Fractions Westinghouse proposed the same core release fractions for the low-volatile elements as those outlined in the EPRI document entitled " Passive ALWR Source Term," issued in February 1991; compared to NUREG-1465, this source term is a reduction by a factor of 5 for the barium and strontium group and for the cerium group, and a reduction by a factor of 2 for the lanthanide group. The low-volatile fission product release fractions in NUREG-1465 were based on (1) the results of the expert panel el! citation for NUREG-1150," Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants"; (2) additional research results obtained since the issuance of NUREG 1150; (3) the results of the in-pile severe fuel damage experiments at the Power Burst Facility; ar,d (4) further examination of the Three Mile Island accident.

In NUREG-1465, the staff selected the 75th percentile value for the low-volatile fission product release fractions on the basis that it bounds most of the range of research data values. The NRC considered the EPRI document prior to the issuance of NUREG-1465 in 1995. Westing-house has not provided new or significant information to the NRC that would prompt reconsidera-tion of the technical bases of NL, REG-1465.

In SECY-96-128, " Policy and Key Technicalissues Pertaining to the Westinghouse AP600 Standardized Passive Reactor Design," dated June 12,1996, the staff stated that it planned to use the low-volatile fission product release fractions outlined in NUREG-1465 for its evalu.ation of the AP600 design review, in its staff requirements memorandum of January 15,1997, the

.d i

3-Commission approved the staff position to use the low-volatile fission product release fractions outlined in NUREG-1465. Therefore, the low-volatile fission product release fractions used in AP600 design are not acceptable for the AP600 design certification. It is the staffs position that the release fractions, including those for low-volatile fission products, for the radiological consequence assessment should be based on Table 3.13 of NUREG 1465.

DSER Open item 15.31 stated that Westinghouse should revise its fission product release fractions, in future SSAR revisions, to reflect the staff position, following the resolution of all remaining source term issues. DSER Open item 15.3-1 remains open until Westinghouse revises the low-volatile fis-lon product release fractions in the AP600 SSAR to be consistent with NUREG 1465.

470.43F - Fission Product Release initiation Time from Fuel Gap Westinghouse used a gap fission product release timing different from that provided in NUREG-1465. In an earlier submittal, Westinghouse proposed that there will be no fission product release from the reactor core until 53 minutes into a postulated design basis loss-of coolant-accident (LOCA), and that the gap and in-vessel releases of fission products will continue for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Subsequently, in Amendment 13 to the SSAR, Westinghouse contended that the release of fission products from the fuel to the containment in the gap release phase would be in two stages: (1) the release of gap activity from 5 percent of the fuel rods would begin instanta-neously at the initiation of a DBA, and (2) the release of gap activity from the remaining 95 percent of the fuel rods would begin at the 50-minute mark from the initiation of a DBA.

In NUREG 1465, the staff provided a realistic estimate of the shortest time for fuel rod failure for a large break LOCA (double ended guillotine rupture of the cold leg pipe) of 10 to 30 seconds without approval of leak before-break, and 10 minutes with leak-before break approval. The basis for this timing is documented in NUREG/CR 5787," Timing Analysis of PWR Fuel Pin Failures," dated September 1992. NUREG/CR 5787 used the FRAPCON?, SCADAPI RELAPAPS MOD 3.0, and FRAPT6 computer codes to calculate for the fuel rod failure time. It is the staffs position, consistent with that provided to the Commission, that the release timing for the radiological consequence assessment should be based on Table 3.6 of NUREG-1465. The AP600 design is a leak-before-break design; therefore, the release of gap activity should begin at 10 minutes from the initiation of a DBA.

DSER Open item 15.3-2 stated that the staffs review of Westinghouse's technical positions regarding fission product release timing and the bounding reactor accident sequences selected for the source term applications was not complete. Subsequently, the staff has completed its review and finds that the gap fission product release timing used in the AP600 design by Westinghouse is not acceptable for the AP600 design certification. * *refore, DSER Open Item 15.3-2 remains open.

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-4L AEROSOL REMOVAL MECHANISMS

. 470.44F - Aerosol Removal Rates An active containment atmosphere cleanup system has r'ot been provided for the AP600 design.

Reliance is placed on natural aerosol removal processes in the containment such as holdup (for decay), sedimentation (for settling), diffusion (for plateout), and leakage (for depletion). In Table 1581 of Appendix 158 to the SSAR, Westinghouse provided aerosol removal coefficients starting at the onset of gap release through the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into a DBA. The values range between 0.43 to 0.72 per hour. The staff requested in RAI Q470.3g that Westinghouse provide the uncertainty distribution for the aerosol removal coefficients because they did not appear to be -

conservative for DBA purposes. This information was not provided in Westinghouse's June 26, 1997 response.

In its evaluation of aerosol removal coefficients, the staff considered two natural processes for removing aerosols from the containment atmosphere over the entire period of an accident s

(30 days):. (1) sedimentetion mechanism of gravitational settling including aerosol agglomeration, and (2) diffusion mechanisms of diffusiophoresis and thermophoresis..in consideration of these two natural processes for removing aerosols from the containment atmosphere, the staff performed quantitative analyses of uncertainties to predict the aerosol removal rates. The uncertainty analyses were performed using Monte Cario methods.

in its evaluation of aerosol removal rates, the staff used (1) the containment geometry (volume, l:

upward facing surface area etc.) and thermal-hydraulic parameters provided by Westinghouse and (2) fission product release timing, fractions, e d release rates as provided in NUREG-1465.

The principal uncertainties in aerosol properties and aerosol behavior considered in the staffs i-analyses included (1) aerosol size distribution, (2) aerosol void fraction (aerosol particle shape factors), (3) non radioactive aerosols, and (4) chemical forms of radionuclides. The staff estimated aerosol removal rates at several confidence leveh (i.e.,10, 50, go, and g5 percent confidence levels).

Tb AP600 design relies on natural circulation currents enhanced by the Passive Containment l

Cooling System (PCCS) to inhibit stratification of the containment atmosphere. The physical mechanisms of natural circulation mixing that occur in the AP600 are discussed in Appendix 15A of the SSAR. Steam generated by decay heat can vent into the containment atmosphere in the l

form of a jet plume through the postulated break or the fourth stage of the ADS. The interaction of the plume with the ambient atmosphere can be described in terms of entrainment flow induced by the plume. Entrainmant flow results in the mixing of ambient atmosphere with the steam flow

_in the plume. The plume will rise to the cont $ ment dome where the steam will be condensed on the inner surface of the containment shell, and the resulting cooler, denser air will fall to the L

operating deck.

Westinghouse provided an estimate of the degree of mixing by calculating volumetric flow rates of gas entrained by a rising buoyant plume associated with steam generated by decay heat. The calculations were made on the basis of a steam production rate corresponding to decay heat at i hour and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the accidenti Entrainment flow rates were es!culated using equations i

presented in an article by Peterson in Volume 37, Supplement 1, of the Intemational Joumal of

- Heat and Mass Transfer, titled, " Scaling and Analysis of Mixing in large Stratified Volumes."- In the. Westinghouse estimate, no credit was taken for cold plumes falling from the containment dome which cause further circulation above the operating deck. Westinghouse estimated the i

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circulation time constant at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to be 490 seconds and at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be 670 seconds.

Confirmatory calculations by the staff using the same equations as Westinghouse, but contain-ment atmospheric conditions calculated by the staff, indicate the estimates to be reasonable.

Therefore, the staff concluded that the AP600 containment atmosphere is well-mixed for the purpose of determining the aerosol removal rates.

Based on its calculations, the staff finds that the conservative lower bound aerosol removal rate with 95 percent confidence level to be about 0.25 per hour ranging from 0.1 to 0.53 per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into a DBA (see Table 1). The best-estimate (50 percent confidence level) aerosol removal rates ranged from 0.4 to 0.7, which are about same as the Westinghouse values. The staff concludes that the 95 percent confidence level values for the aerosol removal rates are the appropriate values for DBA dose calculations; these values provide an acceptable level of conservatism. Therefore, the aerosol removal rates stated in AP600 SSAR Table 15B 1 are not acceptable.

DSER Open item 15.3 5 stated that the staff would perform an independent evaluation of the bounding accident sequence and the aerosol behavior and removal rates corresponding to the selected bounding accident sequence in the containment following a DBA. The staff has completed its evaluation as stated above, but this item remains open until Westinghouse adequately addresses the staff's concems.

Radioloaical Consecuences of loss-of Coolant Accidents (LOCAs)

In SSAR Section 15.6.5, Wesiinghouse analyzed a hypothetical design basis LOCA. Westing-house concluded that certain bounding sets of atmospheric relative concentration values specified in Section 2.3 of the SSAR, in conjunction of the use of natural deposition of fission product aerosol within the containment and controlling the pH of the water in the containment w prevent iodine evolution, are sufficient to provide reasonable assurance that the calculated radiological consequences of a postulated design-basis LOCA will be within the relevant dose criteria established in 10 CFR 50.34 and in GDC 19 (as applied to the AP600 design: 5 rem TEDE).

Since no specific site is associated with AP600 plant, Westinghouse defined the site boundaries only in terms of various hypothetical atmospheric relative concentrations (X/Q) values at fixed EAB and LPZ distances. The staff will perform an independent assessment of short-term (less than 30 days) atmospheric dispersion factors for potential accident consequence analyses on a site specific basis for a combined license (COL) applicant who references the AP600 design, if site specific atmospheric dispersion factors are greater than the enveloping values (e.g, poorer dispersion characteristics) used in this evaluation, a COL applicant may have to consider compensatory measures, such as increasing the size of the site, decreasing the containment leak rate, or providing engineered safety feature systems in the AP600 design, to meet the relevant dose limits set forth in 10 CFR 50.34 and GDC 19 (as applied to the AP600 design: 5 rem TEDE).

All of the fission product releases due to the LOCA are the result of containment leakage. The M600 design does not have engineered safety features (ESF) systems outside of the contain-

r. ent; therefore, no leakage from the ESF systems is considered for the radiological conse-quence analyses. The containment was assumed to leak at its design leak rate of 0.12 weight percent per day for the entire duraticn of the accident (30 days). The AP600 design provides neither an ESF filtration (e.g., charcoal adsorbers) nor a safety-related containment spray system.

A4a AEROSOL REMOVAL MECHANISMS 470.44F - Aerosol Removal Rates An active containment atmosphere cleanup system has not been provided for the AP600 design.

Reliance is piwd on natural aerosol removal processes in the containment such G noldup (for decay), sedimentation (for settling), diffusion (for platoout), and leakage (for der',etion). In

? Table 1581 of Appendix 158 to the SSAR, Westinghouse provided aerosol re' novel coefficients starting at the onset of gap rolesse through the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into a DBA..The values range between_0.43 to 0.72 por hour. The staff requested in RAI Q470.3g that Westinghouse provide -

the uncertainty distribution for the aerosol ramoval coefficients because they did not appear to be conservative for DBA purposes.- This information was not provided in Westinghouse's June 26, 1997 response, in its evaluation of aerosol removal coefficients, the staff considered two naturst processes for

. removing aerosols from the containment atmosphere over the entire period of an accident-

- (30 days):. (1) sedimentation mechanism of gravitational sttiling including aerosol agglomeration, and (2) diffusion mechanisms of diffusiophoresis and thermophoresis, in consideration of these two natural processes for removing aerosols from the containment atmosphere, the staff performed quantitative analyses of uncertainties to predict the aerosol removal rates. -The

- uncertainty analyses were performed using Monte Carlo methods.

In its evaluation of aerosol removal rates, the staff used (1) the containment geometry (volume.

- upward facing surface area etc.) and thermal-hydraulic parameters provided by Westinghouse and (2) fission product release timing, fractions, and release rates as provided in NUREG 1465.

The principal uncertainties in aerosol properties and aerosol behavior considered in the staff's analyses included (1) aerosol size distribution, (2) aerosol void fraction (aerosol particle shape

. factors), (3) non radioactive aerosols, and (4) chemical forms of radionuclides. The staff estimated aerosol removal rates at several confidence levels (i.e.,10, 50, 90, and 95 percent

~

confidence levels).

The AP600 design relies on natural circulation currents enhanced by the Passive Containment Cooling System (PCCS) to inhibit stratification of the containment atmosphere, The physical mechanisms of natural circulation mixing that occurin the AP600 are discussed in Appendix 15A of the SSAR.; Steam generated by decay heat can vont into the containment atmosphere in the i

form _of a jet plume through the postulated break or the fourth stage of the ADS. The interaction of the plume with the ambient atmosphere can be described in terms of entrainment flow induced

. by the plume. Entrainment flow results in the mixing of ambient atmosphere with the steam flow-c in the plume. The plume will rise to the containment dome where the steam will be condensed on the inner surface of the containment shell, and the resulting cooler, denser air will fall to the E operating deck.

Westinghouse provided an estimate of the degree of mixing by calculating volumetric flow rates of gas entrained by a rising buoyant plume associated with steam generated by decay heat. The calculations were made on the basis of a steam production rate corresponding to decay heat at n

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the accidentJ Entrainment flow rates were calculated using equations

. presented in an article by Peterson in Volume 37, Supplement 1, of the Intemational Joumal of-

- Heat and Mass Transfer, titled, " Scaling and Analysis of Mixing in Large Stratified Volumes," In the Westinghouse estimate, no credit was taken for cold plumes falling from the containment dome which cause further circulation above the operating deck. Westinghouse estimated the

r, I

circulation time constant at i hour to be 490 seconds and at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be 670 seconds.

Confirmatory calculations by the staff using the same equations as Westinghouse, but contain-ment atmospheric conditions calculated by the staff, indicate the estimates to be reasonable.

Therefore, the staff concluded that the AP600 containment atmosphere is well-mixed for the purpose of determining the aerosol removal rates.

Based on its calculations, the staff finds that the conservative lower bound aerosol removal rate with 95 percent confidence level to be about 0.25 per hour ranging from 0.1 to 0.53 per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into a DBA (see Table 1). The best-estimate (50 percent confidence level) aerosol removal rates ranged from 0.4 to 0.7, which are about same as the Westinghouse values. The staff concludes that the 95 percent confidence level values for the aerosol removal rates are the appropriate values for DBA dose calculations; these values provide an acceptable level of conservatism. Therefore, the aerosol removal rates stated in AP600 SSAR Table 158-1 are not acceptable.

DSER Open item 15.3-5 stated that the staff would perform an independent evaluation of the bounding accident sequence and the aerosol behavior and removal rates corresponding to the selected bounding accident sequence in the containment following a DBA. The staff has completed its evaluation as stated above, but this item remains open until Westinghouse adequately addresses the staff's concems.

Radioloaical Conseauences of Loss-of Coolant Accidents (LOCAs)

In SSAR Gection 15.6.5, Westinghouse analyzed a hypothetical design basis LOCA. Westing-house concluded that certain bounding sets of atmospheric relative concentration values specified in Section 2.3 of the SSAR, in conjunction of the use of natural deposition of fission product aerosol within the containment and controlling the pH of the water in the containment to prevent iodine evolution, are sufficient to provide reasonable assurance that the calculated radiological consequene.es of a postulated design-basis LOCA will be within the relevant dose criteria established in 10 CFR 50.34 and in GDC 19 (as applied to the AP600 design: 5 rom TEDE).

Since no specific site is associated with AP600 plant, Westinghouse defined the site boundaries only in terms of various hypothetical atmospheric relative concentrati;ns (x/Q) values at fixed EAB and LPZ distances. The staff will perform an independent assessment of short term (less than 30 days) atmospheric dispersion factors for potential accident consequence analyses on a site specific basis for a combined license (COL) applicant who references the AP600 design. If site specific atmospheric dispersion factors are greater than the enveloping values (e.g, poorer dispersion characteristics) used in this evaluation, a COL applicant may have to consider compensatory measures, such as increasing the size of the site, decreasing the c:mtainment leak rate, or providing engineered safety'.sature systems in the AP600 design, to meet the relevant dose limits set forth in 10 CFR 30.34 and GDC 19 (as applied to the AP600 design: S rem TEDE).

All of the fission product releases due to the LOCA are the result of containment leakage. The AP600 design does not have engineered safety features (ESF) systems outside of the contain-ment; therefore, no leakage from the ESF systems is considered for the radiological conse-quence ana'yses. The containment was assumed to leak at its design leak rate of 0.12 weight percent per day for the entire duration of the accident (30 days). The AP600 design provides neither an ESF filtration (e.g., charcoal adsorbers) nor a safety-related containment sprs" system.

6 The staff has reviewed the Westinghouse analysis and finds that certain part..cters and assumptions are unacceptable. They are identified above as Q470.42F,470.43F, and 470.44F.

The staff performed independent radiological consequence calculations for a postulated design 1

basis LOCA using NUREG-1465 source term and aerosol removal rates developed by the staff.

The major parameters and assumptions used by the staff, and the resulting radiological consequence analyses are provided in Table 2. The assumptions and estimates of the radiologi-cal consequences to the control room operators following a LOCA are provided in Table 3.

As shown in Table 2, the staff finds that the radiological consequences at the exclusion area boundary bounded by atmospheric relative concentrations proposed by Westinghouse and the control room operator dose exceed the dose criteria set forth in 10 CFR 50.34 and GDC 19 of Appendix A to 10 CFR Part 50 (as applied to the AP600 design: 5 rem TEDE). Therefore, the AP600 design is not acceptable.

DSER Open item 15.3.4-2 stated that staff would review the bounding accident break size (a large-break LOCA followed by gravity injectiot. failure) proposed by Westinghouse during a NRC/ Westinghouse meeting on source terms, in conjunction with Westinghouse technical positions on fission product release timing. Subsequently,in Amendment 13 to the SSAR, Westinghouse followed the NUREG 1465 source term (with the 3 exceptions discussed in this enclosure) and submitted its technical position on fission product release timing. Therefore, DSER Open item 16.3.4-2 is subsumed by these 3 open issues.

O

i To'Ae 1 - Aerosol Removal Rates Used by Staff to Evaluate a Loss-of-Coolant Accident for the AP600 j

Time Removal Rates Time Removal Rates thou1)

(hour')

(hours)

(hour')

0 0.040 2.7 0.372 0.1-0.035 2.8 0.385 0.2 0.030 2.9 0.400 0.3 ~

0.035 3.0 0.410 0.4 0.050 3.1 0.422 0.5 0.070 3.2 0.432 0.6 0.085 3.3 0.447 0.7 0.100 3.4 0.460 0.8 0.115 3.5 0.470 0.9 0,130 3.6 0.481

- 1.0 0.147 3.7 0.492 1.1 0.162 3.8 0.505 1.2 0.180 3.9 0.518 1.3 0.195 4.0 0.530 1.4 0.210-5.0 0.400 1.5 0.225 6.0 0.300 1.6 0.240 7.0 0.230 1.7 0.255 8.0 0.200 1.8 0.270 9.0 0.180 1.9 0.280 10.0 0.170 2.0 0.290 11.0 0.160 2.1 0.300 12.0 0.150 2.2 0.315 13.0 0.130 2.3 0.325 14.0 0.120 2.4 0.340 15.0 0.100 2.5 0.350 94.0 0.1b3 2.6 0.360 720 0.100 l

l

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-_.....a..a-T,sbl6 2 Staff Assumptions and Estimates of the Radiolo9 cal 1

Consequences of a Loss-of Coolant Accident for the AP600

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Parameter Yalat Power level, M%'t 1972 Fraction of core mventory relossed, fractions (NUREG 1465), dose conversion factors (FGR 11 & 12) t Noble gases 1.0 lodine 0.4 Cesium 0.3 Tellurium 0.05 Strontium 0.02 Barium 0.02 Ruthenium 0.00?5 Cerium 0.0005 Lanthanum 0.0002 Start time for fissie ', product rslease (NUREG 1465)

Coolant Activity, minutes 0

Gap Activity, minutes 10 i

Enriy in Vessel, minutes 40 i

lodines chemical form fractions (NUREG 1465)

Organic 0.0015 Elemental 0.0485 Particulate 0.95 Primary containn

'sakage, weight percent / day 0.12 Accident duration,. S 30 Primary containment free volume, cubic feet 1.62E+6 Atmospheric dispersion values 0-02 hour EAB, sedm' 1.00E 3 3

0-08 hour LPZ, sedm 1.35E-4 8 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LPZ, sedm' 1.00E-4 104 day LPZ, sedm' 5.40E 5 4 30 day LPZ, sedm' 2.20E 5 Badioloalcal Conseagtacts of Deslan Basis LOCA (rem TEQgi EAB 39, LPZ - 11, Cof, trol Room 8.1 e

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s-i Table 3 Assumptions and Estimates of the Radiological Consequences to Control Room Operators Following a LOCA for the AP600 Control room free volume 3.57E+4 ft' Time of bottled air depleted 72 hr Prior to depletion of bottled air Flow from oompressed air bottles 23 ft' UnflHerod inloaka9e o to 12 hr 2.5 cfm 12 to 72 hr 5.0 cfm After depletion of bottled air Air intake flow 1700cfm FiHer efficiencies N/A Recirculation flow N/A Breathing rate of operators in control room for the course of the accident 3.47E-4 m'/sec Atmosphode dispersion values 0to 2 hr 1.8E.3 sedm' 2 to 8 hr 1.3E 3 sedm' 8 to 24 hr 1,1E 3 sedm' 24 to 72 hr 7.3E-4 sedm' 72 to 96 hr 8.4E-4 sedm' 96 to 720 hr 4.8E-4 sedm' Control room operator occupational factors O to 24 hr 1

24 to 96 hr 0.6 96 to 720 hr 0.4 4

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DISTRIBUTION Letter to Mr. Nicholas J. Lloarulo. Dated: November 6, 1997

~ ' Docket File

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• Enclosure to be held for 30 days

• PUBLIC PDST R/F TQuay-TKenyon WHuffman JSebrosky DScaletti I

JNWilson WDean,0 5 E23 ACRS (ii)

JM3 ore,0-15 B18 MPSeimen,015 818 JLee,010 D4 REmch,010 D4 CMiller,0-10 D4 '

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