NSD-NRC-98-5586, Forwards Addl Changes for Chapter 15 of AP600 SSAR
| ML20203H683 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 02/25/1998 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD-NRC-98-5586, NUDOCS 9803030290 | |
| Download: ML20203H683 (5) | |
Text
- --
- 3%
Wes1?nghouse Energy Systems B"*F ""** "#3"3" Electric Corporation DCP/NRCl269 NSD-NRC-98 5586 Docket No.: 52-003 February 25,1998 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T.R. QUAY SUlljECT:
ADDITIONAL CilANGES TO Tilti AP600 STANDARD SAFETY ANALYSIS REPORT (SSAR) CllAPTER 15 ACCIDENT ANALYSES
Dear Mr. Quay:
Attached are markup pages for Chapter 15 of the AP600 Standard Safety Analysis Report. The changes to Table 15.4-4 reflect the final analysis with the adjusted containraent leak rate. The changes to Table 15.A-5 (Sheet 2 of 2) provide the x/Q values for the plant vent in addition to the containment ground level release.
Please contact Ms, Susan V. Fanto (412)374-4028, if you have any questions concerning this material.
g Ilrian A. McIntyre, Man 'er Advanced Plant hafety aad Licensing jml Attachment ec:
T. J. Kenyon, NRC (w/ Attachment)
N. J.1.iparuto, Westinghouse (w/o Attachment)
/j 9803030290 980225
- =
^== os2ogoa.
Ew/
1.11.1 l.l!I.lli.lli.l!.Il.li
trd iEi 5
- 15. Accidnt Analyses Table 15 4 4 (Sheet 1 of 2)
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A R3D EJECTION ACCIDENT Initial reactor coolant iodine acuvity An assumed iodine spike that has resulted in an increase in the reactor coolant activity to 24 pCvg I
of dose equisalent 1131 (see Appendix 15Al
Reactor coolant noble gas acuvity Design 'sasis ac65ity (see Table 11.12)
Secondary coolant iniual iodine actaity 044 pCUg dcse equivalent 1131 (10% of design basis reactor coolant concentrauons listed in Table 11.12)
Fuel cladding failure
- Fraction of fuel rods assumed to fail 0.15 Fission product rap fractions 0.036 Core melting Fraction of core melung 0.00375 Frection of activity released lodines and cesiums 0.5 Noble gases 1.0 lodine chemical form (%)
Elemental 4.85 Organic 0.15
- Particulate 95.0 Core activity See Table 15A 3 in Appendix 15A I Nuclide data See Table 13A 4 in Appendix 15A Reactor coolant mass (Ib)
J,f 39 E45 I d210 l
a.
The assumption of a pre-exisung iodine spike is a cor.,.rvaGve assumption for the initial reactor coolant activity. However, l
compared to the activity ass M to be released from damaged fuel, it is not significant.
Revision: 17 October 31,1997 15A-50 g Westinghouse
e 15, Aceldent Analyses Table 15.4 4 (Sheet 2 of 2)
PARAMETERS USED IN EVALUATING THE RAI,70 LOGICAL CONSEQUENCES OF A ROD EJECTION ACCIDENT Condenser Not available Durauon of accident (days) 30 l Atmosphenc dispersion (x/Q) factors See Table 15A 5 in Appendis 15A Secondary system release path Primary to secondary leak rate (lehr) 260'"
Seconhry coolant mass (Ib) 2.45E45 (Te4) l Duration of steam release from secondary system Jawak 40-- 14CC 1
Steam released from ser.ondary system (lb) 2.45 E+05 4
Partition coefficient in steam 0.01 generators Containment leakage release path 1
Contai";.ent leak rate (% per day)
-449-C./0 Airbor s activity removal coefficients (hr ')
Elemental iodine 2.0
Organic iodine O
Particulate iodine or cesium 0.1 Decontamination factee limit
- for elemental iodine removal 200 1
Time to reach the decontarmnation factor i
limit for elementaliodine (hr) 2.6 ticiant a
Eqmvaient to 1000 gpd at 5613*F and 223n pata b.
From AppendA ISB
\\
Revision: 17 T Westingt)Use 15.4 51 October 31,1997 r--
r
O y
s,
- 15. Accident Analysis l
l Table 15A-5 (Sheet 2 of 2)
ATMOSPIIERIC DISPERSION FACTORS (x/Q)
FOR ACCIDENT DOSE ANALYSIS Main control room X/Q Is/m') at ilVAC Intake for the identifled Releese Points
[ Containment l
Elevated Ground LeveC>
l l
Release
Release Points'p Release Points'y" Fuel llanf' ling Fuel Building Containment Secondary Sid Area' Relief PanelY l
0 2 houis 1.2E 3 2.0E 3 2.0E-2 2.0E-3 3.0E-3 l
2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8.0E-4 1.0E 3 1.8E 2 1.5E 3 2.0E-3 1
8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0E-4 5.0E-4 8.0E 3 f.0E-4 1.0E-3 l
1 - 4 days 4.0E-4 '
5.0E 4 7.0E-3 8.0E-4 1.0E 3 1
4 30 days 3.0E-4 4.0E-4 6.0E 3 7.0E-4 9.0E-4 x/Q (s/m') at Control Room Door for the Identified Release Points
- I E!aated Ground Level.)
I Containment Containment Secondary Side Fuel Handling Fuel Building I
Release
Release Polats'I Release Points #8' Area *'
Relief PanetV i
l 0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.0E-4 1.0E 3 2.5E 3 1.0E 3 1.0E-3 8
I 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.0E-4 6.0E-4 2.0E 3 6.0E-4 6.0E-4 l
8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0E-4 3.0E-4 1.0E-3 3.0E 4 3.0E-4 I
i 4 days 9.0E 3.0E-4 9.0E-4 3.0E-4 3.0E-4 l
4 30 days 8.0E-5 3.0E-4 8.0E-4 2.5E-4 2.5E-4 httst 1.
These dispersion factors are to be used I) for the time penod preceding the isolation of the main control room and actuation of the emergency habitability system. 2) for the time after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the compressed air scpply in the emergency habitability system would be exhausted and outside air would be drawn mto the main control room. and 3) for the determination of control room dos s when the non safety ventilation system is assumed to remain operable such that the emergency habitability system is not actuated.
2.
These dispersion factors are to be used when the emergency habitability system is in opetation and the only path for outside air to enter the main control room is that due to ingress / egress.
Dese dispersion factors apply to releases from the plant ventQ 3.
l
- 44. The listed values bound the dispersion factors for releases from the main equipment hatch the stagmg area hatch.and2--
.u g
.-.m_,,
n.u u;,2-0. These dispersion factors would be used for evaluating the doses in the main control room Revision: DRAFT 15A 16 W W85tinR,tt00S8
i
!E
- 15. Accidect Analysis e
for a loss-of coolant accident, for the containment leakage of activity following a rod ejection accident, and for a fuel handling acct nt occumng insid: the containment.
l [5. The listed values bound the dispersion factors for releases from the steam vents, the steam line safety & power-operated relief vahes, and the condenser air removal stack These dispersion factors would be uwd for evaluating the doses in the main control room for a steam generator tube rupture, a main steam line b.eak, a locked reactor coolant pump rotor, and for the secondary side release from a rod ejection accident. Additionally, these dispersion coefficients are conservative for the small line break outside containment l
- 66. he listed values bound the dispersion factors for releases from the fuel starage and hand'ing area. Rese dispersion factors would be used for the fuel handling accident occurring outside containment.
l
- 67. The listed values bound the dispersion factors for releases from the fuel storage area in the event that spent fuel boiling occurs and the fuel building relief panel opens on higk temperature. Dese dispersion factors are to be used for evaluating the impact of releases associated with spent fuel pool boiling.
Revision: DRAFT W Westinghouse 15A-17
_.. _ _