ML20056F255
| ML20056F255 | |
| Person / Time | |
|---|---|
| Issue date: | 06/09/1993 |
| From: | Joshua Wilson Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| PROJECT-669A NUDOCS 9308260234 | |
| Download: ML20056F255 (19) | |
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UNITED STATES W
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((lh].fij NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D C. 20555-0001
%w f June 9, 1993 Project No. 669 APPLICANT:
Electric Power Research Institute (EPRI)
PROJECT:
ADVANCED LIGHT WATER REACTOR (ALWR) UTILITY REQUIREMENTS DOCUMENT FOR PASSIVE PLANT DESIGNS
SUBJECT:
SUMMARY
OF MEETING BETWEEN THE NUCLEAR REGULATORY COMMISSION (NRC) STAFF AND EPRI HELD ON MAY 18 AND 19, 1993, IN ROCKVILLE, MARYLAND, CONCERNING SEVERE ACCIDENT SOURCE TERM AND DRAFT SAFETY EVALUATION REPORT (DSER) OPEN ISSUES A public meeting was held on May 18 and 19, 1993, at the NRC headquarters in Rockville, Maryland, to discuss design basis and severe accident source terms and outstanding open issues from the staff's DSER on passive plant designs.
In the opening remarks, the NRC staff stressed that this was a noticed public meeting between the NRC staff and EPRI, not an advisory committee meeting, as defined by the newly-promulgated Federal Advisory Committee Act of 1993. A list of attendees and their affiliation is provided in Enclosure 1.
The handouts used in the meeting are provided in Enclosure 2.
Tuesday. May 18. 1993 During the morning session, the staff discussed the details of EPRI's physically-based severe accident source term and how it is being used by Westinghouse in its dose calculations for the AP600 passive plant design. The staff stated that it intends to review applications for final design approval / design certification (FDA/DC) against the criteria in draft NUREG-1465.
If some of the criteria in draft NUREG-1465 change when it is issued as a final document, the staff would review applications for FDA/DC against the new criteria.
The differences between the Westinghouse assumptions for release fraction, species composition, and timing of release, compared to the values used in the staff's draft NUREG-1465, were discussed in the context of a mini-sensitivity study conducted by Westinghouse. The staff agreed to consider vendor-specific release time intervals in its safety evaluation.
However, the staff noted that the policy decision to permit such actions would require Commission approval.
The staff indicated that the AP600 design with the fission-product release timing and release durations given in NUREG-1465 did not meet the EAB 2-hour thyroid dose limits. Westinghouse agreed with the staff's preliminary calculation but proposed its own fission-product release timing for the AP600 design to show compliance with the regulatory requirements without a containment spray system.
The staff stated that NRC will review the fission-product release timing proposed by hg' I
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- June 9, 1993 In order to demonstrate the feasibility of the passive pressurized-water reactor (PWR) design to meet the limits of 10 CFR Part 100 without a containment spray system, EPRI agreed to submit dose calculations for a PWR passive plant using the staff's draft NUREG-1465 values for release fraction i
and isotopic composition, combined with a design-specific timing of releases.
This submittal is expected at the end of the first week of June.
For the passive boiling-water reactor (BWR) design, EPRI has included requirements in the utility requirements document (URD) for its concept of a
" safety envelope" which provides secondary building holdup in order to lower the amount of radioactive materials released in the event of a severe accident. The staff stated that any credit taken in the Chapter 15 design basis accident (DBA) for secondary building holdup would have to be justified by testing, and that surveillance requirements would have to be proposed by the vendor to ensure that performance did not become degraded.
Also during the Tuesday morning session of the meeting, the staff evaluated EPRI's submittal dated April 23, 1993, which contained changes to the URD concerning calculations and methodology for determining natural aerosol removal inside containment, based on recent industry research.
For the passive PWR plant designs, reliance is placed on natural aerosol removal, rather than on a containment spray system, to reduce the amount of radioactive materials released from containment following a DBA. The staff will review the individual applications by passive vendors to determine the extent to which this mechanism is credited in the Chapter 15 DBA.
During the afternoon session, the staff evaluated EPRI's responses to the open issues identified in the DSER on the passive Requirements Document.
In order to resolve staff comments, EPRI committed to submit a number of changes to provide additional clarification and consistency of terminology. This submittal is expected by the end of May.
EPRI and the staff have resolved the major differences for all source term-related issues in principle.
- However, there still exist a few minor differences (iodine species for BWR, amount of gap activity, and magnitude of low-and non-volatile fission products released) between EPRI and the NRC. The staff believes that these can be resolved by the issuance of the final NUREG-1465. The staff believes that it can close out all source term-related open issues in the EPRI passive DSER, once EPM has submitted the agreed-on changes.
These changes will be included in the "thcoming SECY paper for Commission approval.
Wednesday. May 19. 1993 The meeting reconvened on Wednesday morning to discuss EPRI's submittal dated May 3, 1993, which contained URD changes related to use of the revised source term in the calculation of radiation doses for compliance with EPA's rotective action guidelines (PAG) for emergency preparedness.
EPRI committed to submit additional URD changes to clarify the differences in EPRI's requirements for PWRs and BWRs concerning the credit to be allowed for secondary building holdup in DBA and PAG calculations. The staff will review the May 3, 1993, submittal to determine if EPRI's requirements for PAG, using EPRI's physically-based source term and median values for meteorology, will
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- June 9, 1993 lead to a best-estimate dose. The issue of whether a best-estimate value is the appropriate value to use for the PAG dose will be determined later, after EPRI submits its white paper on source term.
EPRI expects to submit this report before the end of the year.
Staff from the Plant Systems Branch and EPRI discussed the staff's current position concerning environmental qualification (EQ) and equipment survivability (ES) regarding mechanical and electrical equipment.
The staff's position is that equipment and components must be qualified for accident environments if credit is taken for them in the Chapter 15 DBA analyses. With regard to ES on the other hand, equipment and components used for severe accident mitigation, but for which no credit is taken in the Chapter 15 DBA analyses, must be demonstrated to survive long enough to perform their intended function.
The staf f and EPRI reached a common understanding of the general approach to applying the new source term to EQ and ES analyses:
for EQ, the staff will likely utilize the in-vessel portion of the
=
NUREG-1465 source term; and for ES, the staff will likely utilize source term phase (s) beyond the in-vessel phase.
Finally, the staff described the schedule and scope of the Commission paper that it is preparing on source term. The staff intends to address the issues of radionuclide attenuation, containment bypass and control room habitability in the same Commission paper with physically-based source term.
This SECY paper is currently scheduled to be sent to the Commission in June.
(Original signed by)
James H. Wilson, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
DISTRIBUTION w/ enclosures:
As stated Central file PDST R/F DCrutchfield PDR JHWilson PShea cc w/ enclosures:
See next page plSTRIBUTION w/o enclosures:
TMurley/FMiraglia WRussell, 12G18 RBorchardt ACRS (11)
JMoore, 15B18 RHasselberg TKenyon JNWilson EJordan, MNBB 3701 GGrant, 17G21 LCunningham,10D4 TEssig REmch, llE22 JLee, 10D4 JRaval, 801 Cli, 8D1 H_ Walker, BD1 LSoffer, NLS324
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LA:PDST:ADART PM:PDST:ADAR PRPB NAME:
PShea f l/ ' W JHWilson:tz, LCunnkham TEssig N h
,DATE:
06/N/d3 06/ 3 /93 N 06/Y/93 06/6 /93 0FFICIAL RECORD COPY: MTGSM518.JHW
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ALWR Utility Steering Committee EPRI Project No. 669 cc:
Mr. E. E. Kintner Chairman Utility Steering Committee l'
Bradley Hill Road Post Office Box 682 Norwich, Vermont 05055 Mr. John Trotter Nuclear Power Division Electric Power Research Institute Post Office Box 10412 l
Palo Alto, California 94303 s
Mr. Brian A. McIntyre, Manager Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Post Office Box 355 Pittsburgh, Pennsylvania 15230 Mr. Joseph Quirk GE Nuclear Energy Mail Code 782 General Electric Company 175 Curtner Avenue San Jose, California 95125' Mr. Stan Ritterbusch Combustion Engineering 1000 Prospect Hill Road j
Post Office Box 500 Windsor, Connecticut 06095 l
Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.
20585 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Room 8002 Washington, D.C.
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w 1e LIST OF ATTENDEES AT MEETING WITH EPRI HELD IN ROCKVILLE, MARYLAND ON MAY 18 AND 19, 1993 l
Name Affiliation J. H. Wilson NRC T. Kenyon NRC T. Essig NRC l
R. Emch NRC i
J. Lee NRC l
C. Li NRC J. Raval NRC H. Walker NRC L. Soffer NRC J. Trotter EPRI D. Leaver Polestar J. Li TENERA J. Metcalf SWEC i
A. Sturdis Westinghouse i
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- 15. ACCIDENT ANALYSES Revision: 0 Effective: 06/26/92 15.6.5.3 Radiological Consequences spike.
The noble gas source term is listed in l
Table 11.1-2 and the maximum iodine spike source term Although the analysis of the core response during a is listed in Table 15A-1.
LOCA (See Subsection 15.6.5.4) shows that core Since the AP600employsleak-before-break method-integrity is maintained, for the evaluation of the radio-ology for lines larger than three inches, the postulated logical consequences of the accident it is assumed that LOCA assumed for the modeling of primary coolant major core degradation and melting occur. Thus, the blowdown into the contamment is a three-inch line break LOCA radiological consequences analysis, which mstead of a large LOCA (double-ecded guillotine considered as a design basis accident analysis, contains rupture of the cold leg pipe). As the reactor coolant a source term that is associated with the severe accident enters the contam-t. the noble gas and iodine activity which is the subject of probabilistic risk assessment in the coolant is assumed to be released to the contain-analyses. ne core degradation source term depends on ment atmosphere.
multiple failures that prevent the cooling of the reactor j
core.
15.6.5.3.1.2 Core Release ne dose calculations take into account the release of activity by way of the contamment purge line prior to The release of activity from the reactor core is its isolation near the begamng of the accident and the assumed to follow the model described in the " Passive release of activity resulting from contamment leakage.
ALWR Source Term
- du-t (Reference 18). Using Purge of the contamment for hydrogen control is not an this model, there is no release assumed from the core intended mode of operation and is not considered in the until one hour into the LOCA. At one ho'ur, core dose analysis. While the normal residual heat removal degradation is assumed to initiate and continues for four system is capable of post-LOCA cooling, it is not a bours. At this time vessel failure is assumed to occur, safety-related system and may not be available following releasing most of the cons into the reactor cavity where the accident. If it is operable, it is used only if the it is covered by a pool of water. Additional releases of source term is not far above the normal shutdown activity to the contamment atmosphere are assumed to pnmary coolant source term, there is no recirculation continue until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the accident.
leakage release path to be modeled. It is assuned that At the end of the release period, all of the core core cooling is accomplished by the passive core cooling noble gases are assumed to be released to the contain-l system which does not pass coolant outside of contain-ment atmosphere as well as significant fractions of the ment.
core iodines, cesiums, and telluriums. Table 15.6.51 lists the fractions of the various nuclide classes that are 15.6.5.3.1 Source Term released from the core to the contamment atmosphere.
ne release of activity to the contamment consists 15.6.5.3.1.3 lodine Form of two parts. He initial release is the actmty contained in the reactor coolant system. His is followed by the ne iodine form is consistet with the
- Passive release of core activity.
ALWR Source Term
- model ne model shows the iodine to be predommantly in the form of non-volatile 15.6.5.3.1.1 Primary Coolant Release cesium iodide with a small fraction existing as elemental iodine. Additionally the model assumes that a portion ne reactor coolant is assumed to have activity of the elemental iodine reacts with organic materials in levels consistent with operation at the design basis fuel the contamment to form organic iodine compotmds.
defect level of 0.25 percent plus pre-existing iodine 15.6-15 W Westinghouse
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....a ne resulting iodine species split is:
15.6.5.3.2.2 Particulates Particulate 0.97 ne particulate activity includes most of the iodine Elemental 0.0285 as well as the other nongaseous nuclides. Particulata Organic 0.0015 are removed from the contamment am--
-h; by sedimentation and deposition. The removal rate is If the post OCA cooling solution has a pH ofless than dependent on the contamment atmsphere humidity 7.0, it is possible for part of the cesium iodide to be which increases to very high levels following the reacter converted to the elemental iodine form. The passive vessel melt-thmugh which is assumed to occur at five core cooling system provides sufficient sodium hydmx-hours into the accident. h removal coefficient is ide solution to the contamment sump following the conservatively determmed to be:
LOCA to maintain the solution pH at 7.0 or greater.
0 - 5 hrs 0.35 hr-1 15.6.5.3.2 In-Containment Activity 5 - 5.5 hrs 1.3 hr-1 Removal Processes
> 5.5 hrs 0.5 hr-1 h AP600 does not include active systerus such as The removal of particulates is assumed to terminase contsmment sprays or filtration for the removal of when the total inventory released 'to the contammant activity from the contamnvut atmosphere. The contain-ati.66phcrc is reduced to 0.1 percent (a decontsmmanon ment atmosphere is depleted of elemental iodine and of factor of 1000).
particulates as a result of natursl processes within the containment.
15.6.5.3.3 Release Pathways 15.6.5.3.2.1 Bemental lodine The release pathways are the cer k~at purge line and the contniment leakage. De activity releases are Elemental iodine is removed by deposition onto the assumed to be ground level releases.
structural surfaces inside the contimment. The removal Durtng the initial part of the accident, before the of elemental iodine by deposition is modeled by:
contamment is isolated, it is assumed that contamment purge is in operation and that activity is released Ad = KwA/V through this pathway until the purge valves are closed.
No credit is taken for the filters in the purge ethanct where )a is the first-order removal coefficient by line.
surface deposition A is the surface area, V is the h majority of the releases due to the LOCA are conta mment building net free volume, and Kw is a mass the result of contamment leakage. h contamment is transfer coefficicut.
assumed to leak at its design leak rate for the first Using the parameters from Table 15.6.5-2, the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For the r-==i~ tar of the accident the con-d s 1.6 hr-1 h removal of elemental tamment pressure is less than half of the contamment value for A i
iodme is assumnd to termmate when the total elemental design pressure and the contammentleak rate is===med iodine released to the contarnment armsphere is reduced to continue at half the design basis leak rate for the to 0.5 percent (a decontammation factor of 200).
durabon of the accident.
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15.6.5.3.4 Offsite Dose Calculation Models 15.6.5.3.6 Analytical Assumptions and Parameters The offsite dose calculation models are provided in t
Appendix 15A. The models include:
The analytical assumptions and parameters used in Thyroid dose due to the inhalation ofiodmes the radiological consequmees analysis are listed in Table 15.6.5 2.
Whole body doses due to the inhalation of iodines and other non-gaseous nuclides 15.6.5.3.7 Differences from the Guidance Whole body doses due to immersion in a cloud of i
noble gas activity.
of Regulatory Guide 1.4 15.6.5.3.5 Main Control Room Dose Model Regulatory Guide 1.4 provides guidance on the calculation of radiological consequences of a pet ='-d In addition to the detennmation of offsite doses the LOCA. The radiological consequences analysis deviates doses to operators in the main control room due to from the recommendations of Regulatory Guide 1.4 in a number of respects. h significant differences are airborne activity are calculated. If high radiation is discussed below.
j detected in the ventilation ductwork supplying the main control room. the main control room is isolated from the 15.6.5.3.7.1 Timing of Activity Release nuclear isind non-radioactive ventilation system (Sec-tion 9.4). and the main control room emergency hab-from Core i
itability system (Section 6.4) is brought into operation.
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The main control room emergency habitability system Regulatory Guide 1.4 assumes that the core releases
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Pressunzes the main control room with air from pm occur at the very begmning of the accident with no surized air bot:les and prevents inleakage of conemmmar. consideration for delay to initiate core degradation or the ed air. The main control room is accessed by way of a duration of time that is required to release the activity to vestibule style entrance which restricts the volume of b wamment. The core release source term used is contamM='~4 air that can enter the main control room as formukted to conservatively model the core degradation a result ofingress and egress.
pmcess for the AP600 decip Activity meering the main control room is assumed to be uniformly dispersed. No credit is taken for the 15.6.5.3.7.2 lodine Chemical Form removal of suborne activity in the main control room although elemental modine and particufste< would be Regulatory Guide 1.4 assumes that the iodmes removed by deposition and sedunstation.
released from the core are predommantly in the elemen-m e atmlr m dose duhdon models are g
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g pg of provided in Appendix 15A. The models melude:
half of the elemental iodine) is 91 percent elemental, 4
d 5 pt partih h Thyroid dose due to the m. halation ofiodines iodine is primarily in the cesium iodide form. The Whole body doses due to the inhalation of iodmes adysis assurnes 2.85 percent elemed, E15 pt and other non-gaseous nuclik rganic, and 97 percat perdeulate.
Whole body doses due to immersion in a cloud of noble gas activity Skin doses due to immersion in a cloud of noble gas i
i activity.
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15.6.5.3.7.3 Source Term Magnitude 15.6.5.3.8.1 Primary Coolant Source Term Regulatory Guide 1.4 assumes that 100 percent of h source term is bued on operation with the the noble gases, 50 percent of the iodines (25 percent design fuel defect level of 0.25 percent, whereas the after credit is taken for plateout of half the inventory), expected fuel defect level is far lesa. Adiitionally, the and one percent of the remaining nuclides are releued iodine spike is a temporary condition that has a low to contammmt atmosphere. The source term used is probability of existing at the same time as the pocndatM similar with respect to noble gases and iodines (100 accident.
percent of noble gases and 55 percent of iodines) but assumes much larger releases of cesiums and telluriums.
15.6.5.3.8.2 Core Release Source Term 15.6.5.3.7.4 lodina -)ose Conversion The assumed core melt is a major conservatism Factors associated with the analysis. In the event of a postulated LOCA no major core damage is expected. Release of Regulatory Guide 1.4 specifies that the iodine dose activity from the core is limited to a fraction of the core conversion factors gisen in ICRP Publication 2 (Refer. gap activity, ence 18) be used. ICRP Publication 2 has been super-seded by ICRP Publication 30 (Reference 19) and the 15.6.5.3.8.3 Atmospheric Dispersion dose conversion factors from ICRP Publication 30 are Factors used for the analysis.
& armnepheric dispersion factors assumed to be 15.6.5.3.7.5 Dose Calculation present during the coune of the accident are consena-Methodology for Whole tively selected. Actual meteorological conditions are Body and Skin expected to result in significantly higher dispersion of the relenmi activity.
Regulatory Guide 1.4 specifies that the whole body and skin doses be calculated using average dirmtegration 15.6.5.3.9 Releases and Doses energies. Instead, immersion doses are calculated using the dose conversion factors denved from EPA Guidance The releases of activity to the environmmt due to Report No.11 (Reference 20). Additionally, although the initial elease by the contammmt purge and the l
Regulatory Guide 1.4 defines the whole body dose as contamment leakage pathway are listed in the external dose, the whole body dose calculatico is Table 15.6.5-3.
extended to include the equivalent whole body dose h thyroid and whole body doses calculated for the contnbution resulting fron. mhalation of iodines and site boundary and the low population zone boundary are other non-gaseous nuclides.
listed in Table 15.6.5-4.
De doses are within the 10 CFR 100 dose limit guidelines of 300 rem thyroid 15.6.5.3.8 Identification of Conservatisms and 25 rem whole, body.
& thyroid, whole body, and skin doses calenlatM ne LOCA radiological consequences analysis for the main control room personnel due to saborne assumptions include a numbe+ of conservatisms. Some activity entering the main control room are listed in of these consenatisms are discussed below.
Table 15.6.5-4. The doses are within the dose criteria defined in SRP Section 6.4 of 30 rem thyroid,5 rem whole body, and 30 rem skin.
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- 15. ACCIDENT ANALYSES Revision: 0 Effective: 06/26/92
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1 Table 15.6.3 5 RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE Thyroid doses (rem)
Case 1 - Accident initiated iodine spike Site boundary 8.5 Low population zone 1.2 Case 2 - Preaccident spike Site boundary 10.2 Low population zone 1.4 Whole body doses (rem)
Case 1 - Accident initiated iodine spike Site boundary 3.3 E-01 b OPulation zone P
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Table 15.6.5-1 Core Activity Releases to the Containment Atmosphereta)
Nuclide 1-5hr 5 - 24 hr Total Noble gases S.0 E-01 2.0 E-01 1.0 E00 lodmes 3.8 E-01 1.7 E-01 5.5 E-01 Cesiums 3.0 E-01 1.8 E-01 4.8 E-01 Teliuriums 8.0 E 02 3.0 E-02 1.1 E-01 4.0 E-03 Strontiums, Bariums, & Rutheniums 4.0 E43 4.0 E-05 Remamder 4.0 E45 k
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f kRelease rates are assumed to be constant over the identified time intend.
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Table 15.6.5-2 (Sheet 1 of 3)
Assumptions and Parameters Used in Calculating Radiological Consequences of a Loss-of-Coolant Accident j
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Primary coolant ource data
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Noble gas concentration See Table 11.1-2 l
Iodine concentration See Table 15A-1 t
Primary coolant mass (Ib) 3.39 E+05 i
i Contamment purge release data Cont =mment purge flow rate (cfm) 4000 Time to isolate purge line (sec) 15 j
i Reactor coolant flow out of 3 inch break (Ib/sec) 2900 Fraction of reactor coolant iodine that becomes airborne 0.5 Core source data Core activity at shutdown See Table 15A-3 Release of core activity to contamment atmosphere See Table 15.6.5-1 Iodine species distribution Elemental (%)
2.85 l
Organic (%)
0.15 Particulate (%)
97 Contamment leakage release data
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l Contamment volume (ft )
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Table 15.6.5-2 (Sheet 2 of 3) i Assumptions and Parameters Used in Calculating Radiological Consequences of a Loss-of-Coolant Accident Contamment leak rate (% per day) 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.12 1 - 30 days 0.06 l
2 Surface area available for deposition (ft )
1.73 E+05 1
Elemental iodine mass transfer coefficient for deposition (m/hr) 4.9 Elemental iodine deposition removal coefficient (br-1) 1.6 DF limit for elemental iodine removal 200 i
Removal coefficient for particulates (hr-1) 0 - 5 hr 0.35 i
5 - 5.5 hr 1.3
> 5.5 hr 0.5 DF limit for particulates removal 1.0 E + 03 Main control room 3
Volume (ft )
42,260 i
Normal air intake flow (cfm) 1400 Time assumed for MCR isolation (sec) 15 Pressunntion air Cow rate using bottled air (cfm) 20 Filtered air intake N/A l
Unfiltered air inleakage from ingress and egress (cfm) 0.3
. Occupancy factor 0 - 24 hr 1.0 1 - 4 days 0.6 4 - 30 days 0.4 Breathmg rate (m /sec) 3.47 E-04 f
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- 15. ACCIDENT ANALYSES Revision: 0 Effective: 06/26/92
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Table 15.6.5-2 (Sheet 3 of 3)
Assumptions and Parameters Used in Calculating Radiological Consequences of a Loss-of-Coolant Accident Offsite power Not Available Atmospheric dispersion factors See Table 15A-5 Nuclide dose conversion factors See Table 15A-4 Nuclide decay constants See Table 15A-4 Offsite breathing rate (m3/hr) 0 - 8 hr 3.47 E45 8 - 24 hr 1.75 E44 24 - 720 hr 2.32 E-04 O
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- 15. ACCIDENT ANALYSES
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Table 15.6.5-3 Activity Released to the Environment Due to a Loss of Coolant Accident With Core Melt Containment Purge C'ontamment leakage releases (Ci)
Isotope Release (Ci) 0-2hr 2-8hr 8 - 24 hr 1 -4 days 4-30 days I-130 4.6 E-04 1.6 E00 2.3 E+01 6.4 E00 3.2 E-01 3.2 E43 I-131 5.1 E-02 1.2 E + 02 1.9 E + 03 9.0 E + 02 1.8 E + 02 4.4 E + 02 I-132 8.9 E-02 1.7 E + 02 2.7 E + 03 1.2 E + 03 2.0 E + 02 1.7 E + 02 I-133 8.9 E-02 2.2 E + 02 3.3 E+03 1.2 E+03 9.5 E + 01 6.1 E00 I-134 2.1 E.02 7.0 E + 01 2.0 E + 02 4.7 E-01 I-135 6.0 E-02 1.8 E +02 2.2 E + 03 4.0 E + 02 8.9 E00 1.9 E-03 Xe-131m 3.1 E-03 1.9 E00 7.6 E+01 2.7 E +02 6.1 E+02 2J5 E+03 Xe-133m 2.9 E-02 7.8 E + 01 2.9 E+03 9.5 E+03 1.4 E + 04 8.8 E+03 Xe-133 4.6 E-01 5.4 E + 02 2.1 E+04 7.4 E+04 1.4 E+05 2.9 E +05 Xe-135m 4.8 E-04 1.5 E00 4.9 E-01 Xe-135 1.3 E-02 1.6 E + 024.7 E + 03 8.2 E+03 1.9 E+03 8.4 E00 Xe-138 8.5 E-04 4.6 E00 1.2 E00 Kr-85m 6.9 E-04 5.4 E + 01 1.2 E + 03 1.1 E + 03 5.5 E + 01 7.8 E-04 Kr-85 1.1 E-02 2.7 E00 1.1 E + 02 4.0 E+02 9.7 E+02 8.3 E+03 Kr-87 1.8 E-03 5.5 E + 01 3.9 E +02 2.7 E + 01 2.5 E-03 Kr-88 5.7 E.03 1.3 E + 02 2.1 E+03 9.0 E + 02 1.0 E + 01 Sr-89 N/A 1.2 E00 1.9 E + 01 1.8 E00 8.8 E41 6.3 E00 Sr-90 N/A 9.9 E-02 1.6 E00 1.6 E-01 7.7 E-02 6.6 E-01 Ru-103 N/A 2.0 E00 3.2 E+01 3.2 E00 1.5 E00 1.0 E + 01 Ru-106 N/A 6.6 E 01 1.1 E + 01 1.0 E00 5.1 E-01 4.2 E00 Te-129m N/A 2.1 E00 3.5 E + 01 1.5 E+01 2.4 E00 1.2 E+01 Te-131m N/A 3.6 E00 5.6 E + 01 1.9 E + 01 1.5 E00 2.2 E-01 Te-132 N/A 3.5 E + 01 5.6 E+02 2.2 E + 02 2.7 E+01 2.1 E + 01 Cs-134 N/A 1.5 E+01 2.6 E+02 1.5 E+02 2.0 E + 01 1.1 E + 02 Cs-136 N/A 4.7 E00 7.9 E + 01 4.6 E + 01 5.7 E00 1.6 E+01 Cs-137 N/A 1.0 E+01 1.7 E + 02 1.0 E + 02 1.4 E + 01 7.7 E00
- Release is less than 10-6 curie.
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- 15. ACCIDENT ANALYSES Revision: 0 Effective: 06/26/92 Table 15.6.5-4 Radiological Consequences of a Loss of Coolant Accident with Core Melt Site boundary dose (0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
Hyroid dose (rem)
Contamment purge 3.0 E-02 Contammmt leakage 5.9 E + 01 Total 5.9 E+01 Whole body dose (rem)
Contamment purge 8.2 E44 Contamment leakage 3.6 E00 Total 3.6 E00 Low population zone dose (0 - 30 days)
Hyroid dose (rem)
Containment purge 4.0 E-03 Contamment leakage 1.61 E + 02-Total 1.61 E+02 Whole body dose (rem)
Contamment purge 1.1 E-04 C=tminment leakage 9.9 E00 f
Total 9.9 E00 Main control room operator dose (0 - 30 days)
Hyroid dose (rem)
Contamment purge 3.0 E00 Contamment leakage 2.03 E+01 Total 2.33 E+01 Whole body dose (rem)
Containment purge 1.0 E41 Contamment leakage 1.9 E00 Total 2.0 E00 Skin dose (rem)
Contamment purge 4.0 E-03 Contamment leakage 2.0 E-01 Total 2.0 E 01 r
15.6 59
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- 15. ACCIDENT ANALYSES Revision: 0 r7 ]
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Table 15.6.5-5 AP600 Conditions for the large Break LOCA Superbounded Case l
l 1.
102% core power,10% uniform steam generator tube plugging, and mmimum flow of reactor coolant are assumed 2.
Only safery-related systems operate.
i 3.
A spectrum of breaks considered.
4.
Begmmng of life cycle 1 conditions are analyzed to manmur core stored energy. Fuel average temperature uncertainty applied to hot rod only.
5.
Top skewed power shape with a positive 25 % axial offset.
1 6.
Lower bound containment pressure.
7.
Conservative manmum values of peakug facton, Fq and FAH.
8.
Hot assembly is placed in an isolated location to mmimire blowdown cooling, t
9.
1979 ANS 5.1 decay heat with uncertamry on hot rod only.
- 10. Accumulator flow delivery mrmmized to extend bottom of core recovery time.
- 11. Single failure postulated is the failure of one of the two parallel path, intact loop core makeup tank isolation valves to open upon receipt of signal.
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l APPENDIX 15A EVALUATION MODELS AND PARAMETERS FOR ANALYSIS OF RADIOLOGICAL CONSEQUENCES OF ACCIDENTS i
This appendix contains the parameters and models 15A.1.2 Whole Body Dose Models that form the basis of the radiological consequences analyses for the various postulated accidents.
The whole body dose calculations include two components. One componentis the direct dose resulting 15A.1 OFFSITE DOSE CALCULATION from immersion in the cloud of radioactive noble gases.
i MODELS b 5cCNd component is the whole body dose that is equivalent to the combined organ doses resulting from the ati n f mgascom actMty.
Radiological consequences analyses include the determination of doses to the thyroid due to inhalation 15 A.1.2.1 Immedon Dose of radioactive iodines and the determination of doses to the whole body.
Assuming a semi-infinite cloud, the utole body 15 A.1.1 Thyroid Dose Model The thyroid doses are calculated using the equation: -
D - EDCF, ER/X/G),
a J
D. - EDCF, ER/BR),(X/0),
where:
a J
Wie Wy dose (rem)
D.
=
where:
le y dose conymon hr for
=
Thyroid dose (rem)
D.
=
isotope i (rem-m'/Ci-s)
Thyroid dose conversion factor (rem DCF
=
Amount of isotope i released during i
R
=
y per curie inhaled) for isotope i b
I (X/Q); - Atmospheric dispersion factor during time penod j (sid)
Breathmg rate during time period j (BR);
15 A.1.2.2 inhalation Dose (m'/s)
(X/Q); = Atmospheric dispersion factor during Based on the modehng provided in Reference 1, the time periodj (5/d) organ-specific doses can be converted to equivalent whole body doses. N whole body doses are caleninted using the equation:
I 15A-1 g g gg h
o
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- 15. ACCIDENT ANALYSES l
Revision: 0 Effective: 06/26/92
]
=
i (IAR), = Integrated activity for isotope i in the
= TDCF, E, R/BR), (X/Q),
main control room dunng time period D
a 7
j (Ci-s/d) where:
Breathing rate during time period j (BR)3
=
Whole body dose (rem)
(d/s)
D.
Fraction of time j that the operator is Whole body dose conversion factor O
=
- DCF, 3
=
(rem per curie inhaled) for isotope i assumed to be present Amount of isotope i released during 15A.2.2 Whole Body Dose Models R,
=
time period j (Ci) ne whole body dose calculations include two Breathing rate during time period j components. He one component is the dir:
dose (BR);
=
(d/s) resulting from immersion in the cloud of radion ave no-ble gases. He second component is the whole body (X/Q); = Atmospheric dispersion factor dunng dose that is equivalent to the combined organ doses time period j (5/d) resulting from the inhalation of nongaseous activity.
15A.2 MAIN CONTROL ROOM DOSE 15A.2.2.1 Immersion Dose MODELS Due to the finite volume of air contained in the main control rms the immersion doses for an operator Radiological consequences analyses include the occupying the main control room is substantially less determination of doses to the thyroid due to inhalation than it is for the case in which a semi-infinite cloud is of radioactive iodines, the determination of doses to the
"==L he finite cloud doses are calculated using the whole body, and the determmation of skin doses.
equation from Murphy and Campe (Reference 2):
15 A.2.1 Thyroid Dose Model I
D = Of 8-- EDCT, E(IAR),0, he thyroid doses are calculated using the equation:
where:
D, = EDCF, E(IAR))BR),0,.
Whole body dose (rem)
D.
Main control room geometry factor l
GF
=
1173 / V.338 O
=
Thyroid dose (rem)
D.
=
lume f the s catrol room (M)
=
Hyroid dose conversion factor (rem
- DCF,
=
Per se Wed) for isotope i
- DCF, Whole bddy dose conversion factor for
=
isotope i (rem-d/Ci-s)
\\
i i
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- 15. ACCIDENT ANALYSES Revision: 0 Effective: 06/26/92 (IAR), = Integrated activity for isotop: i in the since the distance traveled by beta radiation in air is main control room during time period relatively short. The doses are calculated using the j (Ci-s/d) following equation:
Fraction of time period j that the oper-O,
=
ator is assumed to be present D, = EDCF, E(IAR),0j 15 A.2.2.2 inhalation Dose where:
Based on the modeling provided in Reference 1, the D,
= Skin dose (rem) organ-specific doses can be converted to equivalent whole body doses. The whole body doses are calculated
- DCF, Skin dose conversion factor for iso-
=
using the equation:
tope i (rem-d/Ci-s)
(IAR), = Integrated activity for isotope i in the D. - {DCF, WBR),O main control room during time period
/
j (Ci-s/d) where:
O, Fraction of time period j that the oper.
=
Whole body dose (rem) ator is assumed to be present Da
=
The whole body dose conversion factor 15 A.3 GENERAL ANALYSIS
- DCF,
=
(rem per curie inhaled) for isotope i PARAMETERS (IAR), = Integrated activity for isotope i in the 15 A.3.1 Source Terms main control room during time period j (Ci-s/d)
The sources of radioactivity for release are depen-Breathing rate during time period j dent on the specific accident. Activity raay be released (BR)3
=
(gf,)
from the primary coolant, from the secondary coolant, and from the core if the accident involves fuel failures.
Fraction of time period j that the oper-The radiological consequences analyses use conservative O
=
3 ator is assumed to be present design basis source terms.
15 A.2.2.3 Skin Dose Model 15A.3.1.1 Prirnary Coolant Source Term The skin dose due to immersion in a cloud of The design basis primary coolant source terms are listed in Table 11.1-2. These source terms are based on gaseous activity is calculated for the operators in the main control room. A semi-infinite cloud of activity is c ntinuous plant operation with 0.25 percent fuel de-i n9M Despite the finite volume of air contained in facts. The re-mg assumptions used in determming the prtmary coolant source terms are listed in the main control room, the untnersion doses for an Table 11.1-1.
operator occupying the main control room an essen-tially the same as they would be in a semi-infinite cloud
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- 15. ACCIDENT ANALYSES Revision: 0 EEEE Effective: 06/26/92 The radiological consequences analyses for certain secondary side due to pnmary-to-secondary leakage accidents also take into account the phenomenon of enter the steam phase and are discharged via the con-iodine spiking which causes the concentration of ra-denser air removal system.
dioactive iodmes in the primaty coolant to increase sig-mficantly. Table 15A-1 lists the concentrations of 15 A.3.1.3 Core Source Term iodine isotopes associated with a pre-existing iodine spike. Dis is an iodine spike that occurs prior to the Table 15A-31ists the core source terms at shutdown accident and for which the peak primary coolant activity for an assumed three-region equilibrium cycle at end of is reached at the time the accident is assumed to occur.
life after continuous operation at full core thermal These isotopic concentrations are also defined as 24 power. In addition to iodines and noble gases, the Cilg dose equivalent I-131. He probability of this source terms listed include nuclides that are identified as adverse timing of the iodine spike and accident is small. potentially significant dose contributors in the event of Although it is unlikely for an accident to occur at a degraded core accident. He design basis loss of the same time that an iodine spike is at its maximum coolant accident analysis is not expected to result in reactor coolant concentration, for many accidents it is significant core damage, but the radiological conse-expected that an iodine spike would be initiated by the quences analysis assumes severe core degradation.
accident or by the reactor trip associated with the For accidents involving assumed fuel cladding accident. Table 15A-2 lists the iodine appean.nce rates damage, the activity available for release is limited to (rates at which the various iodine isotopes are trans-the iodmes and noble gases present in the fuel-clad gap.
ferred from the core to the primary coolant by way of For short-lived isotopes of iodine and the noble gases the assumed cladding defects) for both normal operation (short-lived being defined as a half-life of less than one and for the iodine spike. The iodine spike appearance year), the gap fraction is conservatively assnmad to be rates are assumed to be 500 times the normal appear-three percent of the core inventory. For Kr-85 (the only ance rates. He appearance rates are assumed to return long-lived isotope of concern), the gap fraction is to normal after the pumary coolant activity reaches the assuned to be 10 percent of the core inventory.
maximum spike concentrations listed in Table 15A-1.
15A.3.2 Nuclide Parameters 15 A.3.1.2 Secondary Coolant Source Term The radiological consequence analyses consider radioactive decay of the subject nuclides prior to their he secondary coolant iodine source term used in release but no additional decay is assumed after the the radiological consequences analyses is conservatively activity is released to the envtronment. Table 15A-4 assumed to be 10 percent of the design basis pnmary lists the decay constants for the m'clides of concern coolant source term provided in Table 11.12. This is Table 15A-4 also lists the dose conversion factors more conservative than using the design basis secondary for calculown of thyroid dose due to inhafarion of coolant source terms listed in Table 11.1-5.
icdines, for calculation of whole body doses as a result Since the iodine spiking phenomenon is short-lived of inblation of iodmes and other nuclides, and for cal-and there is a high level of conservatism for the as-culation of the whole body and skin doses due to im-sumed secondary coolant iodine concentrations, the ef-mersion in a cloud of noble gases. Dese dose conver-feet ofiodine spiking on the secondary coolant iodine sion factors are based on ICRP Publication 30 (Refer-source terms is not modeled.
ence 3) and on EPA Federal Guidance Report No.11 Here is assumed to be no secondary coolant noble (Reference 4).
gas source term since the noble gases entering the
/
15A-4 W85tingh00S8
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- 15. ACCIDENT ANALYSES
. Revision: 0 i
Effective: 06/26/92 15A.3.3 Atmospheric Dispersion Factors Subsection 2.3.4 lists the short term atmospheric dispersion factors (X/Q) for the nference site, Table 15A-5 reiterates these X/Q values and includes i
the X/Q values for the main control room.
i t
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- 15. ACCIDENT ANALYSES Revision: 0 Effective: 06/26/92 15 A.4 REFERENCES 1.
10 CFR 20, ' Standards for Protection Against Radiation,* including =-h s contained in 56 FR 23360 t
(May 21,1991).
2.
Murphy, K. G., Campe, K. M., " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," paper presented at the 13th AEC Air Cleaning Conference.
P 3.
ICRP Publication 30,
- limits for Intake of Radionuclides by Workers," 1978-1981.
4.
EPA Federal Guidance Report No.11, " Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion " EPA-520/1-88-020, September 1988.
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- 15. ACCIDENT ANALYSES Revision: O Effective: 06/26/92 l
l Table 15A-1 j
Reactor Coolant lodine Concentrations for Maximum Iodine Spike of 24 Ci/g Dose Equivalent 1-131 Nuclide pCi/g I-130 0.16 I-131 18.0 I-132 31.2 1-133 31.2 1-134 7.2 I-135 21.0 l
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- 15. ACCIDENT ANALYSES Revision: 0
- -~ ~r_
Effective: 06/26/92 i
l Table 15A-2 lodine Appearance Rates in the Reactor Coolant Equilibriurn Iodine Spike Appearance Rate Appearance Rate 1
Nucinde (Ci/ min)
(Ci/ min) 1130 1.32 E-03 6.60 E-01 I-131 1.03 E-01 5.15 E+01 I-132 5.81 E-01 2.90 E+02 1-133 2.22 E-01 1.11 E+02 I-134 2.85 E-01 1.42 E +02 i
1-135 2.14 E-01 1.07 E + 02 l
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- 15. ACCIDENT ANALYSES i
Revision: 0 Effective: 06/26/92 M
Table 15A-3 Reactor Core Source Term Nuclide Inventory (Ci)
I-130 8.4 E+05 I-131 5.5 E+07 I-132 8.0 E + 07 I-133 1.1 E + 08 I-134 1.2 E+08 I-135 1.0 E+08 Kr-85m 1.4 E + 07 Kr-85 5.4 E+05 Kr-87 2.7 E + 07 Kr-88 3.8 E + 07 Xe-131m 3.9 E + 05 Xe-133m 1.6 E+07
(
Xe-133 1.1 E + 08 Xe-135m 2.2 E+07 Xe-135 3.6 E+07 Xe-138 9.1 E +07 Sr-89 5.2 E+07 St-90 4.4 E+06 Ru-103 8.9 E+07 Ru-106 2.9 E+07 Te-129m 4.7 E+06 Te-131m 8.4 E + 06 Te-132 7.9 E+07 Cs-134 9.0 E + 06 Cs-136 2.8 E+06 Cs-137 6.1 E+06 The following assumptions apply:
- Core thermal power of 1972 MWt (two percect above the design core power of 1933 MW:)
Three-region equilibrium cycle core at end oflife
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- 15. ACCIDENT ANAL.YSES Revision: 0 j
Effective: 06/26/92 E--
.. e e.
Table 15A-4 (Sheet 1 of 2)
Nuclide Parameters A. IODINES Dose Conversion Factors Decay Constant Thyroid Whole Body Isotope (hr-1)
(rem /CI)
(ran/Ci)
I-130 5.59 E-02 7.40 E+04 2.64 E +03 I-131 3.59 E-03 1.07 E+06 3.29 E+04 I132 3.03 E-01 6.29 E+03 6.81 E+02 I-133 3.33 E-02 1.81 E+05 5.85 E+03 1134 7.91 E-01 1.07 E+03
' 1.31 E + 02 I-135 1.05 E-01 3.14 E+04 1.23 E+03 B. NOBLE GASES Dose Conversion Factors Decay Constant Whole Body Skin Isotope (hr-1)
(ran-m3/Ci-sec)
(ran-m3/Ci-sec)
Kr-85m 1.55 E-01 3.07 E-02 5.47 E-02 Kr-85 7.37 E-06 4.84 E-04 4.80 E-02 Kr-87 5.47 E-01 1.46 E-01 3.68 E 01 Kr-88 2.48 E-01 3.70 E-01 1.65 E-01 Xe-131m -
2.41 E-03 1.52 E-03 1.60 E 02 Xe-133m 1.30 E-02 5.53 E-03 3.26 E-02 Xe-133 5.46 E 03 6.24 E-03 1.33 E-02 Xe-135m 2.72 E00 7.75 E-02 3.55 E 02 Xe-135 7.56 E-02 4.82 E 02 7.48 E42 Xe-138 2.93 E00 1.98 E-01 2.14 E41 e
n.
15A 10
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- 15. ACCIDENT ANALYSES Revision: O Effective: 06/26/92 i
Table 15A-4 (Sheet 2 of 2)
Nuclide Parameters C. OTIER NUCLIDES Whole Body Dose Decay Constant Conversion Factor Isotope (hr 1)
(ran/Ci)
Sr-89 5.55 E-04 4.14 E+04 Sr-90 2.81 E-06 1.30 E+06 Ru-103 7.29 E-OS 8.95 E+03 Ru-106 7.83 E-05 4.77 E+06 Te-129m 8.65 E-04 2.39 E+04 Te-131m 2.31 E-02 6.40 E + 03 Te-132 8.89 E 43 9.44 E + 03 Cs-134 3.&4 E45 4.62 E+04 i
Cs-136 2.22 E-03 7.33 E+03 Cs-137 2.63 E-05 3.19 E+04 1
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- 15. ACCIDENT ANALYSES Revision: 0 Effective: 06/26/92
=
1 Table 15A-5 Atmospheric Dispersion Factors (XIQ) for Accident Dose Analysis X/Q (s/m3)
}
f Site boundary O - 2 hr 1.0 E-03 Low population ::one 0 - 8 hr 1.35 E-04 8 - 24 hr 1.00 E-04 i
24 - 96 hr 5.40 E-05 96 - 720 hr 2.20 E 05 i
Main control room 0 - 2 hr 2.20 E-03 2 - 8 hr 1.50 E-03 8 - 24 hr 1.30 E43 24 - 96 hr 8.40 E-04
% - 720 hr 4.80 E-04 i
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VOLUME Ill, CHAPTER 5: ENGINEERED SAFETY SYSTEMS P:mgraph No.
Requirement Rationale Rev.
6.6.5.3 Closed Systems Closed Systems 0
The designer shall evaluate those closed systems penetrating Subject to certain conditions, dosed systems that penetrate O
containment that are not protected by the customary two con-containment meet altomative requirements for containment tainment isolation valves to verify that the system boundary isolation and surveillance testing over the range of licensing will not fal as a direct consequence of an ex-vessel sovere ac.
design tusls cond!! ions. Severe accidents should not pose a cident. If boundary failure cannot be precludod, an attomativo direct hazard to such closed system boundaries in a manner basis for ensuring containment integrity during a severe acci-that would significantly degrade the assurance of contain-dont shall be established.
ment Integrity.
h 6.6.5.4 Equipment Survivability Equipment Survivabi:lty 0
6.6.5.4.1 The design of equipment identified as useful for severe acci-It is important to identify equipment that may be useful for 0
dont mitigation shall provido reasonable assurance that ths severe accident mitigation and provide reasonable assurarK,a equipment can function during severo accident conditions suf-that equipment wlil be avalable, if needed.
ficient to perform its identified function. The design considera-tions for equipment survivability under severe accident condi-tions should include the circumstances of applicable initiating events and tho environment (o.g., pressure, temperature, radia-tion) in which the equipment is expected to function.
6.6.5.4.2 Equipment identified as useful for sovere accident manage-Consideration of the location and protection of equipment 0
ment shall, to the extent practical, be located and arranged useful for severo accident mitigation at the design stage within the plant to enhance its usefulness under severo acci-provides a high degree of assurance that such equipmert wil dent conditions. This includes providing access to equipment be available if nooded, by shleiding it from severo accident radiation sources and ar-O ranging equipment inside containment to provido protection against the severe accident conditions such as hydrogen bum or detonation.
Page 5.6-50 t
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p VOLUME Ill, CHAPTER 5: ENGINEERED SAFETY SYSTEMS I
Paragraph No.
Requirement Rationato Rev.
6.6.5.4 Equipment Survivability (Continued)
Equipment Survivability (Cont!nued) 0 6.6.5.4.3 Equipment identified as useful for the mRigation of severe acci-11is nehher required to provide the same level of reliabity for 2
dents is not required to be qualified for the expected environ-equipment identified as useful for severe accident mitigation ment in accordance with 10CFR50.49 and k is not required to as is required for safety related equipment, nor is R necessary, have redundance and diversity in accordance with 10CFR50, Appendix A, or quality assurance in accordance with Appen-dix B of 10CFR50. Safety related equipment that also serves 3
for severe accident mitigation is required to be quallfled only W
for LDB conditions and provided with qualRy assurance consis-tent with LDB requirements.
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