ML18038A031
| ML18038A031 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 07/26/1985 |
| From: | Mangan C NIAGARA MOHAWK POWER CORP. |
| To: | Butler W Office of Nuclear Reactor Regulation |
| References | |
| (NMP2L-0753), (NMP2L-753), NUDOCS 8507300461 | |
| Download: ML18038A031 (108) | |
Text
REGULATOR NFORMATION DISTRIBUTION TEM (RIDS)
AL 'SION NBR;8507300461 DOC ~ DATE: 85/07/26 NOTARIZED:
NO DOCKET' FA.IL:50-410 Nine Mile Point Nuclear Station< Unit 2~ Niagara Moha 05000410 AUTH,NAME AUTHOR AFFILIATION MANGAN~C.V.
Niagara Mohawk, Power Corp'ECIP
~ NAME 'ECIPIENT AFFILIATION BUTLERgN.
Licensing Branch 2
SUBJECT:
Forwards updated infore secondary containment bypass'.
leakage discussed dur ing 850615 meetin'g w/NRC.Info iritended to assist in closing of SER Open Item 6 ~ Info wil l be; incorporated into Amend 21 to
- FSAR, DISTRIBUTION CODE ~
8001D COPIES RECEIYED LTR ENCL g
SIZE TITLE:" Licensing Submittal:
PSAR/FSAR Amdts 8, Related Correspondence'OTES:
RECIPIENT ID CODE/NAME NRR/DL/ADL NRR L82 LA INTERNAL: ACRS 41 ELD/HDS3 IE/DEPER/EPB 36 NRR ROE'pM ~ L NRR/DE/CEB 1 1 NRR/OE/EQB 13 NRR/DE/MEB 18 NRR/DE/SAB 24 NRR/DHF S/HFEB40 NRR/OHFS/PSRB NRR/DS I/AEB 26 NRR/DS I /CPB 10 NRR/OS I/ICSB 16 NRR/DS I/PSB 19 NRR/DS I /RSB 23 RGN1 EXTERNAL; 24X DMB/DSS (AMDTS)
NRC PDR 02 PNL GRUELgR COPIES LTTR ENCL 1
0 1
0 6
6 1
0 1
1 1
1 1
1 2
2 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
3 3
1 1
1 1
1 1
1 RECIPIENT ID CODE/NAME NRR LB2 BC HAUGHEYpM 01 ADM/LFMB IE F ILE" IE/DQAVT/QA821 NRR/DE/AEAB NRR/DE/EHEB NRR/DE/GB 28 NRR/DE/MTEB 17 NRR/DE/SGEB 25 NRR/DHFS/LQB 32 NRR/DL/SSPB NRR/DS I/ASB NRR/OS I/CSB 09 NRR/DS I/METB 12 NRR/D
/RAG 22 04 I
IB BNL(AMDTS ONLY)
LPDR 03 NSIC 05 COPIES LTTR ENCL 1
0 1
1 1
0 1
1 1
1.
1 0
1 2
2' 1
1 1-1 1
1 0
1 1
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1 1
1 0
1 1-1 1
1 TOTAL "NUMBER OF COPIES REQUIRED:
LTTR 52 ENCL
It g
I 77 t I t
ll 7
0 V MIIAIAIRA 0 gCo{g+Qg NIAGARAMOHAWKPOWER CORPORATION/300 ERIE BOULEVARDWEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474-1511 July 26, 1985 (NMP2L 0453)
Mr. Walter Butler, Chief Licensing Branch No.
2 U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Butler:
Re:
Nine Mile Point Unit 2 Docket No; 50-410 Enclosed please find updated information regarding secondary containment bypass leakage.
This information was discussed during a meeting with your staff on June 15, 1985. This submittal is intended to assist in closing SER Open Item InI6.
The information will be incorporated into Amendment 21 of the FSAR.
Very truly yours, C. V.
Ma an Vice President Nuclear Engineering E Licensing JM:mbg xc:
R. A. Gramm, NRC Resident Inspector Project File (2) 8507300461 850726 "
PDR ADOCK 05000410 E
PDR~
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Nine Mile Point'nit 2
FSAR Railcar entrance to the reactor building railroad airlock is through an interlocking double door airlock system.
The railroad airlock is completely within and along the northeast side of the reactor building at. el 261 ft.
One of the interlocked doors is the exterior railcar door at the north end of the railroad airlock, and the other is the interior railcar door at the south end of the railroad airlock.
A smaller door for personnel ingress and egress is incorporated into the design of the interior railcar door.
All three doors must be closed before any one of them can be opened.
The reactor building pressure control function automatically maintains a uniform subatmospheric pressure of 0.25 in W.G.
by monitoring the differential pressure between the reactor building interi:or and the external atmospheric pressure.
The differential pressure is monitored by a differential pressure transmitter.
The signal that., indicates the differential pressure also controls the position of the recirculation dampers in the HVRS supply fan units.
En the event of reactor building isolation, the reactor building pressure control instrumentation regulates the reactor building pressure by controlling the SGTS recirculation flow.
The reactor building pressure control instrumentation is designed to eliminate fluctuations in reactor building pressure caused by such factors as wind gusts.
Reactor building pre sure is indicated and recorded and loss of negative pressure is alarmed in the main control room.
6.2.3.2.3 Bypass Leakage Paths Table 6.2-56 presents a
tabulation of all primary containment process piping penetrations including the potential reactor building bypass leakage paths.
The potential bypass leakage paths are routed through the reactor building and terminate in the radwaste, standby gas treatment, or turbine generator buildings.
No guard pipes are used on penetrations and therefore guard pipes cannot constitute a bypass leakage path.
All process lines that rely on a closed system within the primary containment as a
leakage boundary terminate within the reactor building; therefore, these lines are not considered potential bypass leakage paths.
Bypass leakage is included in the radiological evaluation of design basis events.
This is discussed in Section 15.6.5 '.
Tables 6.2-55a and 6.2.55b
.show the bypass leakage paths considered.
They include four main steam
- lines, two main Amendment 21 6.2-54 September 1985
Nine Mile Point Unit 2 FSAR steam drain
- lines, one reactor water cleanup line, one feedwater line, four post-accident sampling
- lines, four primary containment purge lines, and four drywell floor and equipment vent and drain lines.
The. analysis used to predict the bypass leakage rates is discussed in Section 6.2.3.2.4.
Single failure is included in the analysis in that failure of one division of electrical power is assumed.
This results in all motor-operated containment isolation valves on that division failing as is (assumed open),
thereby reducing the restrictions to bypass leakage.
This is the worst single failure to consider for this evaluation.
All leakage is conservatively assumed to be across isolation valve seats and to remain within the system piping until released to the environment.
Any leakage escaping across outboard isolation valve stem packing would be
- released to the secondary containment or main steam tunnel.
Any leakage into the secondary containment would be processed by the standby gas treatment system.
Contaminants leaked into the main steam tunnel will be transported to the environment more slowly due to the much larger cross-sectional area of the tunnel and the resulting slower average velocities.
No credit is taken for a reduction in bypass leakage due to water inboard of or trapped between isolation valves.
The isolation valves are assumed to leak containment atmosphere instantaneously following the accident.
No credit is taken for the time required to initially pressurize the volume between the isolation valves.
Leakage transport time to the environment is based on 1/2 of the available horizontal and vertically downward flow piping located between the outboard isolation valve and the environment.
Further conservatism is added to the analysis by the assumption that, all isolation valves in these paths, except the main steam isolation valves (MSIV)and feedwater check
- valves, leak at a rate equal to the maximum permissible, ASME Section XI, Subsection IWV-3426, recommended acceptance level of 7.5 scf/day per inch of nominal valve diameter at functional pressure.
The MSIVs are assumed to leak at 6
SCFH, nearly three times the valve design limit.
Ieakage across check
- valves, except the feedwater check valves, is assumed to be at twice the recommended rate of 7.5 scf/day per inch of nominal valve diameter, as provided for by ASME Section XI, Subsection IWV-3426.
Leakage across the feedwater valves is assumed to be 12 scfh.
Amendment 21 6.2-5@a September 1985
. ~
0 Nine Mile Point Unit 2 FSAR Several process lines are excluded from consideration as bypass leakage paths due to nitrogen seals inherent in the system design and installation arrangement.
These are discussed below.
Nitro en Seals Some systems rely on pressurized nitrogen contained within the system to eliminate bypass leakage.
These nitrogen seals consist of pressurized nitrogen accumulator tanks which apply backpressure to containment isolation valves.
A typical nitrogen seal is shown on Figure 6.2-88.
As the pressurized accumulators are initially at a pressure greater than the maximum postulated containment
- pressure, all leakage is considered to be into-the primary containment as long as the accumulator tank is at a pressure greater than that of the primary containment.
The process lines that eliminate bypass l'eakage by the use of nitrogen or water seals are discussed below and include condensate makeup and drawoff (CNS), reactor core isolation'ooling (RCIC), high pressure core spray (HPCS),
feedwater (FWS),
and instrument air systems (IAS).
CNS While not directly connected to the primary containment, the condensate makeup and drawoff syst: em is used as the alternate fill source to the
Each condensate fillconnection to these systems is isolated by 'means of a normally closed globe valve.
The main supply line into the secondary containment contains a
check valve at the low point which, in case of a pipe break outside the containment, is sealed by a 70-ft leg of water.
Although the condensate makeup and drawoff system is not of seismic design, any line break within the reactor building would provide a
preferential flow path, for containment atmosphere
- leakage, into the reactor building atmosphere.
Under this condition gaseous leakage would be collected by the SGTS and thus not be classified as bypass leakage.
Amendment 21 6.2-54b September 1985
0
0 Nine Mile Point, Unit 2 FSAR RCIC The RCIC path from the primary containment to the condensate storage building is protected from bypass leakage.
When RCIC is taking suction from the condensate storage tank (2CNS-TKlA), the tank static head pressure maintains a
23-psig water seal at valve 2ICS*V28 and/or 2ICS*MOV136 (Figure 6.2-81).
Also, the piping arrangement as shown in Figure 6.2-81 provides a
loop seal with a high point at 2ICS"MOV136.
- Thus, any containment atmosphere leakage through this valve during the period that containment pressures exceed water seal pressure would be trapped at this high point.
If a
LOCA and an SSE take place simultaneously and a
condensate line break
- occurs, 2ICS~MOV129 on the condensate tank line will shut automatically,'reating an additional barrier to bypass leakage.
HFCS The arrangement of the HPCS suction line from condensate storage tank 2CNS-TK1B provides enough static head pressure to keep a 75-ft (32 psig) water seal at the line low point (valve 2CSH*MOV101) in Figure 6.2-83.
- Further, the piping arrangement as shown in Figure 6.2-83 provides two intermediate loop seals with high points at valves 2CSH*MOV118 and 2CSH*V59, ensuring that any containment atmosphere leakage occurring during the 20 min that containment pressures exceed water seal pressure would be trapped between these high points.
If a LOCA and an SSE take place simultaneously and a
condensate line break
- occurs, 2CSH*MOV101 on the condensate storage tank line will shut automatically, creating an additional barrier against bypass leakage.
FWS For loss-of-coolant accidents not involving a feedwater line break, sufficient water exists in the vertical feedwater piping between the containment penetration and the reactor vessel to prevent bypass leakage for at, least 30 days after the accident.
See Figure 6.2-84.
For a
break in feedwater piping inside containment, bypass leakage through this piping is included in the analysis of Section 15.6.
- However, as discussed
- below, a water seal, restored after the break, will effectively prevent escape of containment atmosphere to the environment after 10 min.
Amendment 21 6.2-55 September 1985
0
0 Nine Mile Point Unit 2 FSAR In considering a
break in the feedwater piping within the primary containment, credit can be given to the piping arrangement which provides low stress levels along with pipe whip restraints.
Consequently, it can be stated that the containment penetration is a break exclusion area.
Assuming a break in the feedwater line at the end of the break exclusion region inside the primary'ontainment (see Section 3.6A and Figure 3.6A-20),
sufficient water will remain in the
- line, even after flashing due to initial depressurization, to maintain a vertical water seal on the feedwater isolation valves (Figure 6.2-84).
Water losses due to long-term containment pressure reduction and the associated water vaporization and the backleakage through the two isolation check valves for 30 days will be replenished by reactor water leaking from the break.
Within 10 min after the break the ECCS injection water will reflood the reactor to above the level of the feedwater sparger.
At that point water would flood back into the feedwater piping and then into the intact containment penetration piping (Figure 6.2-84).
This would more than make up for any losses due to leakage out the containment isolation valves.
Thus a continuous water seal is provided to prevent any bypass leakage through the feedwater lines after the initial 10-min refilling period.
Notwithstanding the above, bypass leakage through a ruptured feedwater line is included in the radiological analysis for the entire 30-day period to ensure conservative analysis results.
In addition to the two isolation check
- valves, each feedwater line has a remote-manual gate valve outboard of the isolation valves that may be shut subsequent to a
LOCA anytime the operators determine that feedwater flow is unnecessary or unavailable.
The gate valve provides further back leakage control.
However, this valve is assumed to remain open for the purpose of evaluating bypass leakage.
IAS Instrument air system lines, serving pneumatically operated valves inside the primary containment, are supplied with nitrogen from the nitrogen inerting system while the plant is operating (Figures 6.2-85 and 6.2-86).
Closed outboard penetration isolation valves are maintained pressurized (nitrogen seal) by the nitrogen inerting system at pressures exceeding the maximum containment pressure.
If nitrogen supply to the reactor building is
- lost, pressure on the isolation valves would be maintained by accumulator tanks which would remain pressurized.
In the long
- term, accumulator tanks serving the ADS accumulators and valves may require repressurization due to ADS valve actuation Amendment 21 6.2-55a September 1985
0
Nine Mile Point Unit 2 FSAR losses and leakage.
Post-accident makeup Nq will be provided by two bottled nitrogen connections located outside the reactor building (see Section 1.10, Item II.K.3.28).
6.2.3.2.4 Bypass Leakage Rates Bypass leakage rates as a
function of time after the postulated LOCA are predicted for each path by two
- methods, assuming isothermal flow and isentropic flow.
Table 6.2-55a lists the bypass paths considered and their contributions to the total bypass
- leakage, assuming isothermal flow determined with the.following equation:
m K
(P' P>)/RT
~/~
(6. 2-12)
Where:
P
= Upstream absolute pressure (post-LOCA pressure/
temperature profile per Section 6.2.1)
= Downstream absolute pressure T
= Upstream absolute temperature Gas constant Constant
( de te rmined from the techni c a 1 speci fi-cation of allowable leak rate)
Mass flow rate To quantify the sensitivity of the bypass leakage analysis to the flow model assumption, the bypass calculation was repeated considering the leakage flow to be characterized as isentropic flow through an orifice.
Table 6 '-55b summarizes the isentropic flow results determined with the following equation:
p 2 Y
Pu PD 2 gc Y-RT P
2 PD Y
Pu Y1
.,Y (6. 2-13)
Where:
Amendment 21 6.2-55b September 1985
0
Nine Mile Point Unit 2 FSAR P
= Upstream absolute pressure (post-LOCA pressure/
u temperature profile per Section 6.2.1)
P
= Downstream absolute pressure D
Upstream absolute temperature Gas constant y = Specific heat ratio g
= Conversion constant, c
A = Orifice flow area (to be determined from the technical specification of allowable leak rate)
I m = Mass flow rate The isentropic flow is than the isothermal flow.
generally 20 to'35 percent higher In each case the fractional flow rate is evaluated using the following equation:
Ill f sa pV Where:
(6.2-14) m = Mass flow rate p = Density of containment air and steam mixture (P/RT)
V = Containment volume f = Fractional flow rate The containment bypass leak rate for various paths is calculated based on two closed valves in series or one closed and one open valve depending upon the direct consequence of the postulated'ailure of one emergency diesel generator.
The failure of one diesel generator, combined with a loss of offsite power, would be the worst single failure.
Amendment 21 6.2-55c September 1985
Nine Mile Point Unit 2 FSAR 6.2.3.2.5 Iodine Plateout Considerations The radiological consequences arising from bypass leakage are provided in Section 15.6.5.
The analyses include credits for elemental iodine deposition on the walls of the piping between the outboard isolation valve and the release point.
Details of the iodine deposition analysis can be found in Section 15.6.5.5.3.
6.2.3.2.6 Activity Transport Delay Considerations The leakage of activity from the primary containment. to the environment, is through a portion of piping downstream of the outer isolation valve.
Because of the very low leakage
- rates, there is a considerable transport delay time between the outer isol'ation valve and the release point.
Therefore, the analyses include the credit of the delay time in the dose calculations.
This is
- further, explained in Section 15.6.5.5.3.
Amendment 21 6.2-55d September 1985
0
Bine Nile Point Unit 2 FSA~
TABLE 6.$-55a EVALDATECH OF POTERTEAL BYPASS LEAKAGE PATHS (ESOTHFFNAL FLOP NODEL)
I 21 Line Desgr~pt A'o n 4 Hain steam lines Terai-nation Peqion Turbine Bldg Bypass Leakage Barrier 2-21 I'alves in each line Leak IIate<>>
Tech Spec IIraction/
SCFH( i 45 Dave'R) 0 139x10-4 Conta jument Bvnass Leak Rate (Fraction/Dav) <>>
0-2 hr 0-8 hr 8-74 hr 1-4 dav 4-30 daV
- 0. 540x10-4 0.492x10-~
- 0. 478x10- 4
- 0. 432xl0-3
- 0. 3 05x10 hain steam drain line (in boar d)
Turbine Bldg 1-6h valve
- 1. 875
- 0. 3 44z 1 0-i Hain steam drain line (outboard)
Turbine Bldg 2 II valve
- 0. 625
- 0. 145z10-~
0.211xl 0-+
0.191x10-~
D. 185xl0-~
0.166xlD-~
0.115x10-~
) 21 4 Post accident saaplinq lines Fadvaste, 1-3/4h Tunnel valve in each line
- 0. 2344
- 0. 543x10->
0.217x10-~
0.206970-~
0.277910-
~
0.209x10-'.172910-'ryvell equipment drain line Dryvell equipment vent line Dryvell floor drain line Radvaste Tunnel Radvaste Tunnel aadvaste Tunnel 41 valve 2 II valve 1-6" valve
- 1. 25
- 0. 625
- 1. 875 0 290x10-~
0 422zl0-~
0 381x10-+
0.370x10-~
0.331x10-~
0.229x10-0 I 21
- 0. 145x10-~
0.211zl0-~
0.191x10-~
0.185xl0-~
- 0. 166x>0-~
0.115x10-~
0-435xlO-~
0.633x10-~
0.572x10-~
0.555xl0-~
D.497x10-~
0.344x10-+
l 21 Dryvell floor vent line RL'CU line Radvaste Tunnel Turbine Bldg 1-3h val ve 1-8h valve 2.5 0.579x10-~
0.845x10
~
0.763x10-~
0.739xl0
~
0 663x10
~
0 459x10-~
) 21
- 0. 9375
- 0. 217x10-+
- 0. 317z1 0-i
- 0. 286x10-i
- 0. 277 xl0- i
- 0. 249x10-+
0.172'x10-i
( 21 Feedvater line Turbine Bldg 24h check valves 12 0 278x10-3 0.270z10-3 0 246x10-4 0.239x10-*
0 216x10-~
- 0. 'l53x10-~
) 21 CPS supply line to dryvell Standby 2-14h Gas Trtmt valves Area 4
38 0 102xlO->
0 985x10-~
0.898x10-~
0 873x10" ~
0 789xl0-~
0.557x10-~
21 CPS supply line to dryvell St and by 2-2h Gas Trtmt valves Area
- 0. 625 0 145x 10-i
- 0. 141xl 0-~
0.128z10-~
- 0. 125x10-+
O. 113xl0-+
0.795x10-~
Amendment 21 1 of 2 September 1985
Fine Bile point Jlnit 2 FSAR TABLE 6.2-.5h (Cont)
Line DescriBtion Termi-nationn Recion Bypass Leakage Barrier Leak Rate<>>
Tech Spec Fraction/
SCPH<1
~ i Day< >>
C.
Containment Bypass Leak Pate gPractionfnay) <~i I
21 0-2 br 0-8 hr 8-24 hr 1-4 d~a 4-30 day CPS supply line to supp.
chamber<<>
St and by 2-12<<
Gas Trtat valves Area 3.75 0.523x10-i 0
- 0. 462z10-i 0.450z10-i
chamber~~>
Standby 2-2<<
Gas Trtat valves Area
- 0. 625 0.871x10-5 0.845z10-5 0.770x10-a 0.749x10-5 0.677x10-=
0.478x10-~
21
<>>Std Conditions:
14.7 psia and 68DP
<>>Fraction/Day is defined as fraction of dryvell voluae leakage/(1ay per line under test conditions.
<>>Test Conditions:
Air medium; 40 psig and 800P
<'>The leak rate is based on ASLE Section ZI (Subsection I%V-3426) applied to each valve, except for rain steaa and feedv, ater lines.
~>>Fraction/day is defined as fraction of dryvell volume leakage+ay under LOCA conditions.
<<>Leak rate is defined as a fraction of entire primary containment volume under.
LOCA conditions.
Aaendaent 21 2 of 2
Septeaber 1985
l
Nine Nile Point Unit 2 FSAP TABLE 6.2-5Kb EVALUA ION OF PCTEETZAY. BTPASS LEAFAGE PATLS (ISEHTPO>IC PLOP NODEL)
Line Descr irt ion Terai nation Peci.on Bypass Leakage Barrier Leak Sate< >>
Tech Spe" Fraction/
SCPNCi i)
Davci>
l.
Conta irment Bvnass Leak Rat e fFraction/Day'l <a >
0-8 hr S-54 hr 1-4 dav 4-30 daV I
21 4
Hain steam lines Turbine Bldg 2-21 v valves in each line
- 0. 139 xl0-a
- b. 651 x 1 0- a 0.606x10-v 0.591x10-a 0.550x10-a 0 411x10-3 Naia steam drain line (in boar d)
Turbine Bldq 1-6" valve 1
875 0.435z10"~
0.663x10-~
0.615z10-~
0.597z10-~
0.564x10-~
0.442x10-~
Hain steam Brain line (outloard) 4 Post acciBent sampling lines Turbine Bldg
>advaste Tunnel 1
21I val ve 1-3/4" valve in each line
- 0. 625
- 0. 2344 0 145z10-+
- 0. 543z10-5
% 221x10-i 0.205x10-i
- 0. 307z10-i
- 0. 298 x10-i
- 0. 2 82x10-i 0 147z10 0 221z10 Dryvell equipment drain line Dryvell eguipment vent line Dryvell floor drain line Padvaste Tunnel Padvaste wunnel Padvaste Tunnel 4 II valve 1
2 II valve 1
6n valve 1
25
- 0. 625 1 875
- 0. 290z10-~
4 442x10-~
0.410z10-~
0.398z10-~
0 376z10-+
- 0. 294x10-+
- 0. 145x10-~
~0.221x1 0-~
- 0. 205x10- ~
- 0. 199 z10- ~
- 0. 188x10-~
0.1 47x10-~
0 435x10
~
g 663x10-~
0 615z10-~
0.597z10-~
0 564x10
~
0.442x10-~
I Dryvell floor vent line Badvaste Tunnel 3 II valve
- 0. 9375
- 0. 217 x10-~
'6. 331x10-~
- 0. 307z10-~
- 0. 298 x10-~
- 0. 282x10-~
- 0. 221x10-~
PMCU line Turbine B'g 1-8" valve 2.5
- 0. 579z10-~
@.883x10-~
0.820z10-~
0.796 z10-~
0.751x10-~
0.589z10-~
)
21 Peedvater line Turbine Bldg 2
24 II check valves 12
- 0. 278 x10 t) 326z10->
0 303z10->
0.296zl0->
0 275xl0 0 206x10 CPS supply line to dryvell CPS supply line to dryvell Standby 2-14" Gas Trtmt valves Area Standby 2-2" Gas Trtat valves Area 4
38
- 0. 625 0 102x10
- 0. 145x10-i
- 10. 119x1 0-a
- 0. 111z10-*
- 0. 1 08 xl 0- 3
- 0. 100 x10-a
- 0. 7 51x10-i D 170x10
~
0 158z10-~
0 154x10-~
0 143x10
+
0.107x10 21 Amendment 21 df 2 September 1985
line Pile Point Uhit 2 FSAR TABLE 6. 2-55b (Cont)
Line CPS supply line<<>
to supp.
chaaber TPrai-nation Re@ j,on Bypass Leakage Barrier Stand by 2-12" Gas Trtat valves Area Leak ~ate~>>
Tech Spe" Fraction/
SCFH<~
~>
Dav< >>
L 0 523x10-i
- 3. 75 Containaent Bynass Leak Hate (FractiongDaI) C~>
O-S hr 8-24 hr 1-4 dav 4-30 daI
.'0. 612x10-~
0.570x10-~
- 0. 556x10- ~
- 0. 517xl0-~
- 0. 3 87x10-~
I2i 21 CPS supply line< ~>
to supp chaaber St andby 2-2" Gas Trtat valves Area
- 0. 625 0.871x10-a 0.102x10-~
0.950x10-4 0.926xlO-a 0.861x10-a 0.644xl0-4
<>>Std Conditions:
14.7 psia and 68oF
~>>Fraction/Day is defined as fraction of drywell volune leakage/day per line under test conditions.
<>>Test Conditions:
Air aediua; 40 psig and BO~F
<<>The leak rate is based on ASBB Section XI (Subsection IQV-3426) applied to each valve, except for aain stean lines.
<>>FraCtian/Day iS define" aS fraCtiOn Of dryvell VOluae leakage/ay under LOCA COnditinnS.
<~>Leak rate is defined as a fraction of entire priaary containaent volune under IOCA conditions.
Aaendnent 21 2'of 2 Septeaber 1985
Pm>>
trntioa Syste ~
~~bigtdii()R CDC ot neg.
, gtifr
%ISIS>>
Ro I al feedvatet 55 line A to RPT FSAR Arrange Site scot tlXi dial liSRIEi" vater 24 6 2-70 Sh 3
Loca'<104 of ra)vc Inside/
or<side Fr)aery Conte(a-
~ >>4 1 Oats(dc la side Length ot ripe - Coa ta(sara<
to cats(de Iso 1atios
'rbiwa 2 ~ -1 ~
Type TCSC
< ~ I Po tent 141 byp>>ss Leakage RA&
Isa ERE)IEC KVK tst 2FCS>>lOT2)l b2) FO)ll 2FVS ~T12l 022 fbldl 2 ZEE Seine CLeck Sviog Check Oper RX()I AOT Ptoccss Process Spring Open (test oaly) 5/A Opeo Closed Closed
- Closed Closed 1 irr< ~ )
XB1319 EEI5~4
~(E)E sot>>41 Post-EEARRIZ 2ECVR55u "'LRLitn
~~HIRZ 1$014 tion FO>>>>r 510Oal Loft<<4)
<>>)
Closate Ti>>c
<4
~ )
rove<'oorcc
<))
S/l Rer>>tse floe bc)cree floe The t 1 4 4 R/1 C4kcs for oae valve volrre to pass throagh the valve sloe all>> Point Oait 2
TSAS TAbLE 6 2-56 (Cont) 11 ~ 32 I 21 V~ tet 0
6 2 70 SL 0otside 16 ~
4>>
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0
ine Mile Point Unit 2. FSA 15.6.5 Loss-of-Coolant Accidents (LOCA) (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary)
Inside Primary Containment This event involves the postulation of a spectrum of piping breaks inside containment varying in
- size, type, and location.
The break type includes steam and/or liquid process system lines.
This event is also assumed to be coincident with a safe shutdown earthquake (SSE).
The event has been analyzed quantitatively in Sections 6.2, 6.3, 7.1, 7.3, and 8.3.
Therefore, the following discussion provides information not presepted in the subject sections.
All other information is covered by cross-referencing.
The postulated event represents the envelope evaluation for liquid or steam line failures inside containment.
15.6.5.1 Identification of Causes and Frequency Classification 15.6.5.1.1 Identification of Causes There are no realistic, identifiable events which would result in a
pipe break inside the containment of the magnitude required to cause a
LOCA coincident with SSE plus SACF criteria requirements.
The subject piping is designed to high quality and strict industry code and standard criteria and severe seismic and environmental conditions.
- However, since such an accident provides an upper limit estimate to the resultant effects for this category of pipe breaks, it is evaluated without the causes being identified.
15.6.5.1.2 Frequency, Classification This event is categorized as a limiting fault.
15.6.5.2 Sequence of Events and Systems Operation 15.6.5.2.1 Sequence of Events The sequence of events associated with this accident is
.;shown in Table 6.3-2 for core system performance and Table 6.2-8 for barrier (containment) performance.
Following the pipe break and scram, the low-low water level or high drywell pressure signal initiates HPCS and RCIC systems at time 0, plus approximately 30 sec.
- 15. 6-9
N Mile Point Unit 2 FSAR The low-low-low water level or high drywell pressure signal initiates MSIV closure, and both the LPCS and LPCI systems at time 0, plus approximately 40 sec.
Since automatic actuation and operation of the ECCS is a
system design basis, no operator actions are required for the accident.
15.6.5.2.2 Systems Operations
'Accidents that could result in the release of radioactive fission products directly into the containment are the results of postulated nuclear system primary coolant pressure, boundary pipe breaks.
Possibilities for all pipe break sizes and locations are examined in Sections 6.2 and 6.3, including the severance of small process system
- lines, the main steam lines upstream of the flow restrictors, and the recirculation loop pipelines.
The most severe nuclear system effects and the greatest release of radioactive material to the containment, result from a
complete circumferential break of one of the two recirculation loop pipelines.
The minimum required functions of any reactor and plant protection system are discussed in Sections 6.2, 6.3,.7.3, 7.6, and 8.3, and Appendix 15'5.6.5.2.3 Effect of Single Failures and Operator Errors Single failures and operator errors have been considered in the analysis of.the entire spectrum of primary system breaks.
The consequences of a LOCA with considerations for single failures are shown to be fully accommodated without the loss of any required safety function (see Appendix 15A for further details).
15.6.5.3 Core and System Performance
- 15. 6. 5. 3. 1 Mathematical Model The analytical methods and associated assumptions which are used in evaluating the consequences of this accident provide conservative assessment of the expected consequences of this very improbable event.
The details of these calculations, their justification, and bases for the rr(odels are developed in Sections 6.3, 7.3, 7.6, 8.3, and Appendix 15A.
15.6.5.3.2 Input Parameters and Initial Conditions Input parameters and initial conditions used for the analysis of this event are given in Table 6.3-1.
15.6-10
Nine Mile Point Unit 2 FSAR 15.6.5.3.3 Results Results of this event are given in detail in Section 6.3.
The temperature and pressure transients resulting as a
consequence of this accident are insufficient to cause perforation of the fuel cladding.
Therefore, no fuel damage results from this accident.
Post-accident:
tracking instrumentation and control is assured.
Continued long-term core cooling is demonstrated.
Radiological input is minimized and within limits.
Continued operator control and surveillance is examined and guaranteed.
15.6.5.3.4 Consideration of Uncertainties e
This event was conservatively ana'yzed (see Sections 6.3, 7.3, 7.6, 8.3, and Appendix 15A for detai.ls).
15.6.5.4 Barr'er Performance The design basis for the containment is to maintain its integrity, and experience acceptable stresses after the instantaneous rupture of the largest single primary system piping within the structure, while also accommodating the dynamic effects of the pipe break at the:same time an SSE is also occurring.
Therefore, any postuila:ed LOCA does not result in exceeding the containment
'design limit.
For details and results of the analyses, see Sections 3.8, 3.9, and 6.2.
- 15. >.5.5 Radiological Consequences Three separate radiological analyses are provided for this accident:
The first
- two, isentropic and isothermal (see Section 6.2.3..2),
present two different approaches to the 'design basis analysis.
The results of both approaches are used in determining the adequacy of the plant design to meet 10CFR100 and General Design Criterion 19 guidelines.
2.
The third is based on assumptions considered to provide a
realistic estimate of radiological consequences.
This analysis is referred to as the realistic analysis.
E 15.6.5.5.1 Design Basis Analyses The
- methods, assumptions, and conditions used to evaluate this accident are in accordance with those guidelines set forth in the NRC Standard Review Plan 15.6.5, Revision 2, Amendment 18 15.6-11 March 1985
ii
Nine Mile Point Unit 2 FSAR and Regulatory Guides 1.3 and 1.7.
Specific values of parameters used in this evaluation are presented in Table 15.6-13.
15.6.5 '.2 Fission Product Release from Fuel It is assumed that the reactor is operating at a power level of 3,489 MWt for 1,000 days prior to the accident.
The airborne source immediately available for release from primary containment contains 100 percent of the core noble gas inventory and 25 percent of the core halogen inventory.
The suppression pool source contains no noble gases and 50 percent of.the core halogen inventory.
While not specifically stated in Regulatory Guide 1.3 or Standard Review Plan 15.6.5, the assumed release of 100 percent of the core noble'as activity and 50 percent of'he halogen activity implies fuel damage approaching melt conditions.
Even though this condition is inconsistent with operation of the ECCS system (Section 6.3), it is assumed "applicable for the evaluation of this accident.
The airborne activity available for release from the primary containment at T=O hr post-LOCA is presented in Table 15.6-14.
15.6.5.5.3 Fission Product Transport to the Environment The transport pathways consist of leakage from the primary containment to. the environment through several different mechanisms.
Where applicable, the SGTS filter efficiency for halogen removal is assessed as 99 percent.
The mechanisms for leakage from the primary containment are discussed in the following paragraphs:
Containment leakage The technical specification leak rate of the primary containment and its penetrations (excluding the bypass leakage paths) is 1.1 percent per day for the duration of the accident.
Additional primary containment leakage occurs through one of the traversing incore probes (TIP) at 0.21 percent per day into secondary containment.
All this leakage is to the reactor
- building, and from there to the environment via a 3,500 cfm SGTS.
- Credit, is taken for 50 percent mixing within the reactor building.
This leakage is the same for the isothermal and isentropic approaches to the design basis analysis.
2.
Leakage from ESF components outside primary containment 1
gpm total leakage of suppression pool fluid into the reactor building is assumed to occur for the duration of the accident.
Ten Amendment 21
- 15. 6-12 September 1985
0
'I
Nine Mile Point Unit 2 FSAR percent of the iodines in this leakage become airborne and available for release through the SGTS.
This leakage is the same for the isothermal and isentropic approaches to the design basis analysis.
3.
Reactor building pressurization
- During the time when the.
pressure in the reactor building is greater than negative one-quarter inch water gauge compared to the environment leakage from the primary containment,
No credit is taken for mixing or filtration.
This leakage is the same for the isothermal and isentropic approaches to the design basis analysis.
P,l Bypass leakage The piping paths listed below provide potential routes for post-LOCA primary containment atmosphere to bypass the reactor building and the standby gas treatment system and be released directly to the environment.
1.
Main steam lines (4).
2.
Feedwater line.
3.
Post-accident sampling lines (4).
4.
Main steam drain lines (2).
5.
Reactor water cleanup line.
6.
Drywell equipment drain and vent lines (2).
7.
Drywell floor drain and vent lines (2).
8.
Primary containment purge lines (4).
Section 6.2.3 describes in detail the two methods used to determine the leak rates through the isolation valve(s) for each path.
These two methodologies, one considering an isothermal flow process and the other considering an isentropic flow process, define the two separate approaches to the flow design basis analysis.
Using the leak rate data from Tables 6.2-55a and 6.2-55b, a prerelease holdup time is calculated for each bypass leakage path using the slug-flow method.
The slug-flow method assumes that the Amendment 21 15.6-12a September 1985
Nine Mile Point Unit 2 FSAR leakage front occupies the full cross-sectional area of the pipe as it travels through the line.
As time increases and the isolation valves leak at progressively lower rates due to a lowering of the differential pressure across the valves, the slugs move. more slowly down the pipe.
The holdup time is equal to the time it takes the leakage through the valve to fill the volume of piping between the isolation valve and the release point.
One-half of the actual calculated holdup times are used.
Credit is taken for the deposition of elemental and particulate iodines (which constitute 91 percent and 5 percent, respectively, of the total iodines released from the core per Regulatory Guide 1.3) on the walls of the piping between the isolation valve and the release point for each bypass leakage path.
In
- addition, 4 percent of the iodines released are assumed to be in organic form.
The elemental and particulate iodine plate out mechanism described below is derived from information given in references 6
through ll (Section 15.6.7).
The ratio of the elemental and particulate iodine concentration at the release point to the concentration at the isolation valve is:
'RP CXV where:
C C
= Elemental and particulate iodine concen-tration at the release point and the isolation valve, respectively (uCi/cc)
A
= Pipe deposition area d
V
= Deposition velocity for elemental iodines
= 9.0 x 10 exp
[8100/RTs) cm/sec Ts Gas constant
= 1.987 cal./mole-K Absolute temperature
(~K)
A
= Flow cross-sectional area C
Amendment 21 15.6-12b September 1985
Nine Mile Point Unit 2 FSAR V~ = Flow velocity The change in CRP/C1V due to the change in temperature and flow velocity with time is taken into account conservatively.
The temperature of the air/steam mixture and the piping for the main steam
- line, FW
rate of 100~F/day for first two days.
For all other iodine deposition analyses, the temperature of the air/steam mixture and piping is assumed to be 120 F
and 104 F
for the duration of the accident.
These temperature profiles are assumed to be the same for the isothermal and isentropic flow approaches.
No reduction due to deposition is taken for the remaining 4 percent of the core, iodines released that are'ot in elemental or'articulate form.
Therefore, the ratio of the total iodine concentration at the release point to the total iodine concentration at the isolation valve is:
- 0.96
+ l.0
~ 0.04 RP CRP XV IV As shown in Section 6.2.3, the flow rates for each bypass leakage path are different for the isentropic and isothermal cases.
Therefore, the iodine deposition factors are also different.
All pertinent data used to determine the radiological consequences of a
design basis LOCA are presented in Table 15.6-13.
Amendment 21 15.6-12c September 1985
0
Nine Mile Point Unit 2 FSAR 4
~
THIS PAGE INTENTIONALLY'LANVi
~
'mendment 18 15.6-12d March 1985
Nine Mile Point Unit 2 FSAR Fission product releases to the environment based on the foregoing assumptions for both the isothermal and isentropic approaches are given in Tables 15.6-15a and 15.6-15b.
15.6.5.5.4 Results The calculated exposures for the design basis analysis are presented in Tables 15.6-16a and 15.6-16b and are within
'the guidelines of 10CFR100 and General Design Criterion 19.
15.6.5.5.5 Realistic'nalysis The realistic analysis is based on a realistic, but still conservative, assessment of this accident.
Specific values of parameters used in the evaluation are presented in Table 15.6-13.
Fission Product Release from Fuel Since this accident does not result in any fuel damage, the only activity re'leased to tne primary con"a'nment is that activity contained in the reactor
- coolant, plus any additional activity which may. be released zs a
consequence of reactor scram and vessel depressuri"ation.
While there are contained in the various activation and corrosion products reactor coolant, the products of primary Amendment 18
- 15. 6-13 March 1985
0
~
Nine Mile Point Unit 2 FSAR importance are the iodine isotopes I-131 to I-135.
The design reactor coolant iodine activities are normalized to the maximum technical specification' limits.
The coolant concentrations and the normalized concentrations for these isotopes are presented in Table 15.6-17.
Considering that approximately 40 percent of the released liquid flashes to steam, it is conservatively assumed that 40 percent of the released iodine activity is airborne initially.
Amendment 13 15.6-13a August 1984
Nine Mile Point Unit 2 FSAR THIS PAGE INTENTIONALLYBLANK Amendment,13 15.6-13b August 1984
Nine Mile Point Unit 2 FSAR As a consequence of reactor scram and depressurization, ad-ditional iodine activity is released from those rods which experienced cladding perforation during normal operation.
The reactor coolant iodine activities that are normalized to the maximum technical specification'4 'imits are multiplied by SOO to account for iodine spiking'hile no measurements have been obtained during a pressure transient as rapid as the LOCA, it is difficult to predict the actual release rate from the fuel as a consequence of iodine spiking.
It is, therefore, arbitrarily a'ssumed that 100 percent of the spiking source term is released during the time period that 40 percent of the discharged coolant is flashing to steam.
The initial airborne iodine concen-trations resulting are presented in Table 15.6-18.
Since it is also assumed that plate-out and condensation remove 50 percent of the.
airborne iodine
- activity, the resultant activity airborne in the primary containment from spiking is presented in Table 1S.6-18.
Fission Product Trans ort to the Environment A
large reactor coolant: pipe fails and instantaneously releases the entire mass of coolant in the reactor vessel and recirculation system to the primary containment.
The activity airborne in the primary containment and available for release leaks to the reactor building at a constant rate for 30 days.
From 0 to 90 sec post-LOCA, the reactor build-ing cannot maintain a
pressure less than -1/4 in.
W.G.
During this period, the activity leaking from the primary containment is assumed to be released directly to the environment.
The standby gas treatment system (SGTS) begins operation within 25 sec after the LOCA signal.
After 90 sec, the reactor building returns to a pressure less than
-1/4 in.
W.G.
and for the remaining 30 days the activity airborne in the reactor building is removed and filtered by the SGTS and exhausted through the main stack.
The leak rate from the primary containment to the reactor building is 1.1 percent/'day, where SO percent mixing is as-sumed to occur.;
Release from the reactor building to the environment through a
99 percent iodine-efficient SGTS is at a
rate of 3,500 cfm.
The integrated isotopic activity released to t5e environment is presented in Table 15.6-19.
Amendment 13 15.6-14 August 1984
Nine Mile Point Unit 2 FSAR Results The calculated radiological exposures for this event are presented in Table 15.6-20
- and, as
- shown, are a small frac-tion of 10CFR100.
Amendment 10 15.6-14a April 1984
Nine Mile Point Unit 2 FSAR THIS PAGE INTENTIONALLYBLANK Amendment 10 15.6-14b April 1984
Nine Mile Point Unit 2 FSAR 15.6.6.3.3 Consideration of Uncertainties This event was conservatively analyzed and uncertainties were adequately considered (Section 6.3).
15.6.6.4 Barrier Performance Accidents that result in the release of radioactive materials outside the containment are the results of postulated breaches in the RCPB or the steam power-conversion system boundary.
A break spectrum analysis for
.the complete range of reactor conditions indicates that the limiting fault event for breaks outside thy containment is a
complete severance of one of the main steam lines as described in Section 15.6.4.
The feedwater system piping break is less severe than the main steam line break.
Results of analysis of this event can be found in Sections 6.2.3 or 6.2.4.
15.6.6.5 Radiological Consequences 15.6.6.5.1 Design Basis Analysis The NRC provides no specific regulatory guidelines for the evaluation of this accident; therefore, no specific design basis analysis is presented.
- However, the radiological consequences of this event are enveloped by the results of analyses for the main steam line break (presented in Section 15.6.4.5).
This is considered justified since the feedwater line check valves isolate the reactor from the downstream side of the break at time Q+
seconds after accident initiation.
Therefore, there is no reactor coolant backflow via the feedwater lines connected to the RPV.
The only contribution is from main steam, which must first, pass through 0he turbines, condenser, and other condensate and feedwater system components, which reduce 'sotopic concentrations due to decay and demineralization.
In the main steam line break analysis, a coincident iodine spike is assumed based on a compound spiking sequence giving 4 uCi/gm dose equivalent I-131, as described in Section 15.6.4.5.2.
15.6-17
-I
Nine Mile Point Unit 2 FSAR 15.6.7 References 1.
- Moody, F.
J.
Maximum Two-Phase Vessel Blowdown From Pipes.
ASME Paper Number 65-WA/HT-1, March 15, 1965.
2.
- Nguyen, D.
et. al.
Radiological Accident Evaluation The CONCAC03 Code, NEDO-21143-1, December 1981.
3.
DRAGON 4
- Code, Dose and Radioactivity From Nuclear Facility Gaseous
- Outflows, NU-115, Version 4, Level 1, April 1982.
4.
USNRC Standard Technical Specifications for General Electric Boiling Water Reactors.
NUREG-0123, Rev.
2, Washington, D.C., August 1979.
5.
USNRC Standard Review Plan, Radiological Consequences of a
Small Line Carrying Primary Coolant Outside Containment, 15.6.2, Rev.
2, July 1981.
6.
Report of the Special Committee on Source
- Terms, The American Nuclear Society, September 1984.
8.
- Genco, J.
M., et al.
Fission Product Deposition and Its Enhancement Under Reactor Accident Conditions:
Deposit'on on Primary - System Surfaces, March 1969.
- Kress, T.
S.
and Wright, A. L.
Status of the Validation of the TRAP-MELT Computer Code for the Accident Source Term Reassessment Study (ASTRS),
Oak Ridge National Laboratory (Draft).
9.
NUREG/CR-2713, BMI 2091, Vapor Deposi ion Velocity Measurements and Correlations for Iz and CsI, May 1982.
10.
NUREG/CR-0009, Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vessels, October 1978.
ll. EPRI NP-876, Surface Effects Airborne Radioiodine at Light
- Plants, September 1978.
in the Transport of Water Nuclear Power Amendment 18
- 15. 6-18 March 1985
Hine Nile Point Unit 2 PSAR Data and assumptions used to esti mate radioactive source from postulated accidents TABLE 15 6-13 LOSS-OP-COOLAHT ACCIDEHT - PABANETERS TABULATED POR POSTULATED ACCIDENT AHALTSES Design Basis Assuantions Isentropic Case Isotheraal Case Realisti" Basis Assumptions a.
Pover level b.
Release of activity to containment air c.
Release of activity to suppression pool d.
Iodine fract. ons (1)
Or ganic (2)
Elemental (3)
Particulate, e.
Compu ter code used <>>
f.
Single active failure 2.
Data and assumptions used to estiaate activity released 3,489 Nltt 100% core noble gas inventory 25% core halogen inventory 50% core halogen inventory
- 0. 04 0 91 0.05 Dragon 4
Loss of diesel 3,489 NWt 100% core noble gas inventory 25% core halogen inventory 50% core halogen inventory
- 0. 04 0.91
- 0. 05 Dragon 4
Loss of diesel 3,489 NWt 100% of iodines in coolant flashing to steam H/A 0.0 1
0
- 0. 0 Dragon 4
i2>
a.
b.
Total mass of coolant released Pour main steam lines Bypass leakage rates (fractions of dryvell volume per day)
(main steam tunnel release)
H/A 0-2 hr 0-8 hr 8-24 hr 24-24. 37
- 24. 37-96 96-720 hr hr 0 Octa 0
0 0
0 0
0 5 48-4 4 12-4 H/A 0-2 hr 0-8 hr 8-24 hr 24-31.74 hr 31.74-96 hr 96-720 hr 0 0<2) 0.0 0.0 0.0
- 4. 28-4 3.05-4
- 2. 72+8g 0.0 0.0 0 0 0.0 0
0 0.0 2l Amendment 21 1 of 11 September 1985
Nine Nile Point Unit 2 PSAF TABLE 15. 6-13 (Cont)
Des~in Basis Isentropic Case Assum tions 2
Isothermal case mealisti" Basis Assumotions Iodine concentration ratios 0-8 hr 8-24 hr 24-24.37 hr 24.37-96 hr 96-720 hr 0
0 0.0 0
0 0 273 0.04 0-8 hr 8-24 hr 24-31.74 hr 31.74-96 hr 96-720 hr
- 0. 0
- 0. 0 0.0 O. 255
- 0. 047 0.0 0.0 0
0
- 0. 0 0.0 co mipe inside diameter (actual/design basis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient-pipe inside surface Inboard main steam drain line 25.23 in/25.23 in 508 ft/312 ft 3,355 ft2/2,060 ft2 0-1 day 450oP 1-2 day 450-350op 2-3 day 350-250oF 3-4 dav 250-1200P 4-30 day 1200P 25.23 in/25.?3 in 508 ft/312 ft 3,355.ft2/2,050 ft2 0-1 day 450o P 1-2 da y 450 3 5 0o F 2-3 day 350-2500?
3-4 day 350-120oP 4-3'0 day 1200F N/A N/A N/A Bypass leakage rates (fraction of dryvell volume per day)
(main steam tunnel release) 0-2 hr 0-5-12 hr 5 12-8 hr 8-24 hr 24-96 hr 96-720 hr Q ~ Q(2) 0 0
5 93-5 5 97-5 5 64-5 42-5 0-2 hr 0-5.51 hr 5.51-8 hr 8-24 hr 24-96 hr 96-720 hr Q ~ 0( 2)
- 0. 0
- 5. 44-5
- 5. 55-5 4 97-5 3 44-~
- 0. 0 Q.o
- 0. 0 0.0 0.0 0.0 21'odine concentration ratios 0-5.12 hr 5.12-720 hr 0
0 Q.O4 0-5.51 hr 0.0
- 5.51-720 hr
- 0. 04 0.0 0 0 Pipe inside diameter (actual/design basis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient-pipe inside surface d.
Pour post accident sampling lines 5.761 in/5.761 in 84.3 ft/84.0 127 ft2/127 ft2 0-720 hr 120oP 5.761 in/5.761 in 84.3 ft/84.0 ft 127 ft2/127 ft2 0-720 hr 1200P N/A N/A N/A Bypass leakage rates (fraction of dryvell volume per day)
(radvaste tunnel release)
Amendment 21 0-2 hr 0-8 hr 8-24 hr 2 of 11 3 ~ 31-5( 2 ) Q-2 hr 3.07"5 0-8 hr 2.98-5 8-24 hr 3-17-5 (2 )
- 2. 86-5
- 2. 77-5
- 0. 0 0.0 0.0 September 1995
Nine Mile Point Unit 2 FSAH TABLE 15. 6-13 (Cont)
Des~in Basis AssumJ tions Isothermal Case Isentro1Jic Case Realistic Basis Assumntions e.
Iodine concentration ratio Pipe 'nside diameter (a"tual/design basis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient pipe inside surfaces One feedvater line Bypass leakage rates (fraction of dryvell volume per day) <~>
(main steam tunnel release)
Iodine concentration ratio 24-96 hr 96-720 hr All times 0.04 0.18 in/0.19 in 935 ft/0 ft ft>/1.0 fte 0-720 hr 120oF 0-2 hr 0-4.89 hr 4.89-8 hr 8-24 hr 24-96 hr 96-720 hr 0-4.89 hr 4.89-8 hr 8-24 hr 24-96 hr 96-720 hr 2.82-5 2.21-5 P 0(2>
0 0
2.93-4 2.96-4 2.75-4 2.06-4 0
0 0 '24 0 228 0
112 0 04 24-96 hr 96-720 hr All times 0.04 0.18 in/0.18 in 935 ft/0 44 ft>/1.0 ft~
0-720 hr 120oF 0-2 hr 0-6.20 hr
- 6. 20-8 8-24 hr 24-96 hr 96-720 hr 0-6.20 hr 6.20-8 hr 8-24 hr 24-96 hr 96-720 hr 2 48-5
- 1. 72-5 P
QC2>
0.0
- 2. 35-4
- 2. 39-4
- 2. 16-4 1 53-4
- 0. 0
- 0. 419
- 0. 221
- 0. 087
- 0. 04 0 0
- 0. 0 0
0 N/A N/A N/A
- 0. 0 0.0 0.0 0.0 0
0 0.0 0.0
- 0. 0 0 0 0.0 0.0 21 Pipe inside diameter (actual/design basis)
Pipe length (actual/design basis)
Deposition surface
{actual/design basis)
Temperature transient pipe inside surface f.
Outboard main steam drain line 19.876 in/19.876 in 50 ft/50 260 fthm/260 fthm 0-24 hr 425-325oF 24-48 hr 325-225oF 48 72 hr 225 120oP 72-720 hr 120oF 19.876 in/19.876 in 50 ft/50 260 fthm/260 fthm 0-24 hr 425-325oF 24 48 hr 325 225oF 48-72 hr 225-120oF 72-720 hr 120oF N/A N/A N/A N/A N/A Bypass leakage rates (fraction of dryvell volume per day)
Amer daent 21 0-2 hr 0-8 hr 3 of 11 2.21-5< >>0-2 hr 2 05-5 0-8'r 2 11-5c2')
- 1. 91-5
- 0. 0
- 0. 0 September 1985
l
Nine Nile Point Unit 2 PSA>
TABLE 15. 6-13 (Cont)
Desi n Basis Assumntions
((
Isentropic Case Isothermal Case Realisti" Basis Assumotion s (main steam tunnel release)
Iodine concentration ratio 8-24 hr 24-96 hr 96-720 hr 0-2 hr 0-8 hr 8-24 hr 24-96 hr 96-720 hr 1 ~ oo 5
1 88-5 1 47-5 0.05 0.05 0.041 0.04 0
04 8-24 hr 24-96 hr 96-720 hr 0-2 hr 0-8 hr 8-24 hr 24-96 hr 96-720 hr
- 1. 85-5
- 1. 66-5
- 1. 15-5 P. 05
- 0. 05
- 0. 04
- 0. 04
- 0. 04 0.0 0.0 0.0 0.0 Pipe inside d'meter (a"tu a l/de sign ha sis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Te-perature transient pipe inside surface g.
Reactor va ter cleanup line 1.687 in/1.687 in 17.46 ft/0 ft 7.71 ftm/1.0 ftR 0-720 hr 120oF
- 1. 687 in/1. 687 in 17.46 ft/0 ft 7 71 fthm/1.0 fthm 0-720 hr 120oF N/A N/A N/A N/A Bypass leakage rates (fra"tion of dryvell volume per day) (>>
(main steam tunnel release)
Iodine concentration ratios 0-2 hr 0"8 8-10.03 h
10.03-24 hr 24-96 hr 96-720 hr 0-8 hr 8-10.03 hr 10.03-24 hr 24-96 hr 96-720 hr p p(2) 0.0 0.0 7.97-5
- 7. 51-5 5.89-5 0
0 0-0
- 0. 218 0.064 0.04 0-2 hr 0-8 hr 8-10. 80 hr 10.80-24 hr 24-96 hr 96-720 hr 0-8 hr 8-10. 80 hr 10.80-24 hr 24-96 hr 96-720 hr 0 p(2) 0 0
0.0
- 7. 41-5
- 6. 63-5
- 4. 59-5 0.,0 0.0 0.213
- 0. 059
- 0. 04 0.0 0.0
- 0. 0 0.0 0.0
- 0. 0 0.0
- 0. 0 n.p 0.0
- 0. 0 2};
Pipe inside (actual/design basis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient pipe inside surface Amendment 21 7.187 in/6.813 in 599 ft/250 ft 1,127 ft~/446 ftm 0-24 hr 551-450oF 24-48 hr 450-350oF 48-72 hr 350-250oF 72 96 hr 250 120oF 4of 11 7.187 in/6.813 in 599 ft/250 168.7 ft>/466 ft~
0 24 hr 551 450oP 24-48 hr 450-350oF 48-72 hr 350-250oF 72-96 }iz 250-120 P
N/A N/A N/A N/A N/A N/A September 1935
Nine Nile Point Unit 2 FSAII TABI,E 15.6-13 (Cont)
Desian Basis Assunntions Isentro}lie Case Isothermal Case oealistic Basis Assumntions h.
Drywell eguipment drain (DER) line 96 720 hr 120oP 96-720 hr 120oF Bypass leakage rates (fraction of drywell volume per day)
(radvaste reactor building vent release) 0-1. 24
}1?'.24-2hr 0-1 24 1.24-8 hr 8-24 hr 24-96 hl 96-720
}1?'
P<2) 4.25-5<
0.0 4.01-5 3 98-5 3.76-5 2 94-5 0
1 29 hr
- 1. 29-2 hr 0-1.29 hr 1.29-8 hr 8-24 hr 24-96 hr 96-720 hr P P<2)
- 4. 00-5<>>
- 0. 0
- 3. 70-5
- 3. 70-5
- 3. 31-5
- 2. 29-5 0.0 0.0 0.0
- 0. 0 0.0 0.0 0.0 Iodine concentration ratio 0-1.24 hr 1.24-720 hr 0
0 0
04 0-1.29 hr 0.0 1.29-720 hr 0.04 0.0 0.0 Pipe inside diameter (actual/des'n basis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient-pipe inside surface Dryvell eguipment drain (DER) vent line 4.026 in/4.026 in 75 ft/35 ft 79 ft2/37 ft2 0-720 hr 120oP 4.026 'n/4.026 in 75 ft/35 ft 79 ft2/37 ft2 0-720 hr 120oP N/A H/A H/A 21 Bypass leakage rates (fraction of dryvell volume per day (radvaste reactor build'ng vent release)
Iodine concentration 0-0. 96 hr 0.96-2 hr 0-0.96 hr 0.96-8 hr 8-24 hr 24-96 }Ir 96-720 hr 0-0.96 br 0.96-720 hr P 0(2)
- 2. 12-5(
0.0 2.01-5 1.99-5 1.88-5 1 47-5 0.0
- 0. 04 0-1 hr 1-2 hr 0-1 hr 1-8 hr 8-24 hr 24-96 hr 96-720 hr 0-1.0 hr 1.0-720 hr P P(2) 2 00-5<2)
- 0. 0
- 1. 86-5
- 1. 85-5
- 1. 66-5
- 1. 15-5 0.0
- 0. 04 0.0 0.0
- 0. 0 0.0 0.0 0.0
- 0. 0 0.0 0.0 Pipe inside diameter (actual/design basis)
Pipe length (actual/design basis)
Depos'tion surface (actual/design basis)
Temperature transient Amendment 21 2.067 in/2.067 in 350 ft/54 189 ft2/29 ft2 0-720 hr 120oP 5 of 11 2.067 in/2.067 in 350 ft/54 ft 189 ft2/29 0-720 hr 120oF H/A N/A N/A September 1935
0 0
Hin e Nile Point Unit 2 FS AF.
pipe inside surface Drymell floor drain (DFR) line TABLB 15. 6-13 (Cont)
Desian Basis Assummtions Tsentropic Case Isothermal Case
>ealistic Basis hssumPtions Bypass leakage rates (fraction of dryvell volume per day)
(radwaste reactor building ven t release)
Iodine concentration ratio 0-2 hr 0-2.23 h?
2.23-8 hr 8"24 hr 24-96 hr 96-720 hr 0-2.23 2.23-720 hr P
PC2>
0.0 5.98-5 5.97-5 5.64-5 4.42-5 0.0 0
04 0-2 hr 0-2.36 hr 2 36-8 hr 8-24 hr 24-96 hr 96-720 hr 0-2.36 hr 2.36-720 hr Q
QC 21 0.0
- 5. 51-5
- 5. 55-5
- 4. 97-5 3 44-5 0.0
- 0. 04 0.0
- 0. 0
- 0. 0
- 0. 0 0 0
- 0. 0
- 0. 0 0.0 Pipe inside diameter (actual/design basis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient-pipe inside surface Dryvell floor drain (DFP) vent line 6.06 in/6.065 in
- 46. 8 ft/37 74.25 ft2/59 ft2 0-720 hr 120oF
- 6. 06 in/6. 065 in 46.8 ft/37 74.25 ft2/59 ft2 0-720 hr 120oF N/A N/A N/A H/h 21 Bypass leakage rates (fraction of dryvell volume per day)
(radwaste reactor building vent release)
Iodine concentration ratio 0-1.94 hr 1.94-2 hr 0-1.94 hr 1.94-8 hr 8-24 hr 24-96 hr 96-720 hr 0-1.94 hr
- 1.94-720 hr 0-QC2>
3 19-5C 21 0.0 2.99-5 2.98-5 2.82-5
- 2. 21-5 0.0 0 04 0-2 hr 0-2.04 hr 2.04-8 hr 8-24 hr 24-96 hr 96-720 hr 0-2.04 hr 2.04-720 hr Q
PC 2>
0.0
- 2. 76-5
- 2. 77-5
- 2. 4o-5
- 1. 72-5
- 0. 0
- 0. 04 0.0
- 0. 0 0.0
- 0. 0
- 0. 0 0
0 0.0 0.0 0
0 Pipe inside diamete" (actual/design basis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient-pipe inside surface h endment 21 3.068 in/3.068 in 114 ft/65 ft 91 5 ft2/52 ft2 0-720 hr 1200F 6of11 3.068 in/3.068 in 114 ft/65 ft 91.5 ft2/52 ft2 0-720 hr 1200F N/A N/A H/A September 1985
Hine Bile Point Unit 2 FSAR TABLE 15. 6-13 (Con t) 1.
Dryvell purge inlet line Design Basis Assumotions-Isentronic Case Isothermal Case Realistic Basis Assumption s Bypass leakage rates (fraction of dryvell volume per day)
(SGTS building release) 0-2 hr 0-6.01 hr 6.01-8 hr 8-24 hr 24-96 hr 96-720 hr P.PCz) 0.0 1.07-4 1 08-4
- 1. 00-4
- 7. 51-5 0-2 hr 0-7.55 hr 7.55-8 hr 8-24 24-96 hr 96-720 hr 0.
0.
8.
8.
5.
PC 2) 0 59-5 73-5 89-5 57-5 0
0
- 0. 0
- 0. 0 0.0 0.0
- 0. 0 Iodine concentration ratio 0-6.01 hr 0.0
'6.01-720 hr 0.04 0-7.55 hr 7.55-720 hr 0.0
- 0. 04 0.0 H/A Pipe inside diameter.
(actual/design basis)
Pipe lengths (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient pipe inside surface Metvell purge inlet line 13.25 in/13.25 in 45 ft/32 156 fthm/111 fthm 0-720 hr 1040F 13.25 in/13.25 in 45 ft/32 ft 156 fthm/111 ft2 0-720 hr 104oF H/A H/A H/A H/A Bypass leakage rate (fraction of vetvell volume per day)
(SGTS building release) 0-2 hr 0-5.73 hr 5.73-8 hr 8-24 hr 24-96 hr 96-720 hr P
PCZ) 0.0
- 5. 50-5 5.56-5 5 17-5
- 3. 87-5 0-2 hr 0-7.19 hr 7.19-8 jir 8-24 hr 24-96 hr 96-720 hr p 0C2) 0.0
- 4. 42-5
- 4. 50-5
- 4. 06-5
- 2. 87-5 0.0 0.0 0.0 0
0 0.0 0.0 Iodine concentration rat.'o Piping inside diameter (actual/design basis)
Pipe length (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient pipe inside surface 0-5.73 hr 0.0 5.73-720 hr 0.04 12.0 in/12.0 in 129 ft/32 405 ft>/100 ft>
0-720 hr 104oF 0-7.2 hr 7.2-720 hr 12.0 in/12.0 in 129 ft/32 ft.
405 ft~/100 fti 0-720 hr 104oF 0.0 0 04-H/A H/A H/A Amendment 21 7 of 11 Septemher 1935
0
Sine Bile Point Unit 2 PSAF n.
Drywell purge makeup line TABI.E 15. 6-13 (Cont)
Desian Basis Assumptions Isentr~o ic Case isothermal Case Pealisti" Basis hssumotions Bypass leakage rate.
(fraction of vetvell volume per day)
(SSTS building release) 0-0. 83 hr 0.83-2 hr 0-0.83 hr 0.83-8 hr 8-24 hr 24 o6 hr 96-720 hr 0.0(>>
0.0-1.03 hr 1 64-5(2>
1.03-2 0 hr 0 00 0-1.03 hr 1.55-$
1.03-8 hr
- 1. 54-5 8-24 hr 1.43-5 24-96 hr 1 07-5 96-720
}IL P 0(2) 1 '4-5(2) 0.0
- 1. 25-5
- 1. 25-5
- 1. 13-5
- 7. 95-6 0
0 0
0 0.0 0.0 0.0
- 0. 0 0.0 Zodine concentrat on ratio 0-0.83 hr 0.83-720 hr 0.0 0.04 0-1. 03 hr P. 0 1 03-720 hr
- 0. 04 0.0 0.0 0 ~
Pipe inside diameter (actual/design basis)
Pipe lengths (actual/design basis)
Deposition surface (actual/design basis)
Temperature transient pipe inside surface Retvell purge makeup line 1.939 in/1.939 in 129 ft/41 65.5 ft2/20.8 ft2 0-720 hr 104oP 1.939 in/1.939 in 129 ft/41 65 5 ft2/20 8 ft2 0-720 hr 104oF 8/h
}t/A 8/A 8/A 21 Bypass leakage rates (fraction of vetvell volume per day)
(S"TS building release) 0-0. 83 hr 0.83-2.0 hr 0-0 83 hr 0.83-8 hr 8-24 hr 24-96 hr 96-720 hr P P(2) 9 88-6(2>
0 0
9.35-6 9 26-6 8 61-6
- 6. 44-6 0-1.03 hr 1.03-2.0 hr 0-1. 03 hr
- 1. 03-8 hr 8-24 hr 24-96 hr 96-720 hr 0 ~ 0( 2>
- 8. 04-6(>>
- 0. 0
- 7. 52-6
- 7. 49-6 6.77-6 4.78-6 0.0 0
0 0.0 0.0 0.0 0-0 0
0 Xodine concentration rat' 0-0. 83 hr 0
0.83-720 hr 0.04 0-1. 03 1.03-720 hr
- 0. 0
- 0. 04 0.0 0.0 Pipe inside diameter (a"tual/design basis)
Pipe length (actual/design basis Deposition surface (actual/design basis)
Temperature transient pipe inside surface 1.939 in/1.939 in 104 It./4 1 52.8 ft2/20. 8 ft2 0-720 hr 104oP 1.939 in/1.939 in 104 ft/41 52.8 ft2/20.8 ft2 0-720 hr 104oP Amendment 21 8 oi 11 September 1995
Nine Bile Point Jnit 2 PSA>
TABLL 15. 6-13 (Cont) p.
Con tain ment leakage rate (main stack release)
Design Basis hssumntions<<>
1.1%'er day of primary containment volume for duration of accident Bealistic Basis Assumotions 1.1% per day of primary containment volume for duration of accident g.
Transfer incore probe leakage rate (main stack release) r.
Reactor building leak rate (main stack release) s.
Percentage mixing in reacto" building air 0.21% per day of primary containment volume for duration of accident 3,500 cfm through standby gas treatment (SGTS) 50%
N/A 3, 500 cfn through SGTS 50+
t.
Beactor building pressurization time (ra dias te/r eac tor bui'ing vent release) u.
SOS halogen filtration efficiency v.
ESP leakage to reactor building (main stack release)
(1)
Leak rate (2)
Iodine partition factor (air/vater) 90 sec 99'%
gpm
- 0. 1 90 sec 99%
0.0 0.0 3.
All other pertinent data a.
(3)
Drywell free air volume Primary containment free air volume Suppression pool volume 2.85+5 fta
- 4. 73+5 ft>
1.45+5 ftm N/A 4.73+5 ft~
b.
Reactor building (1)
Pree air volume Amendment 21 3 88+6 ft>
9 of 11 3.88+6 ftm September 1995
t
B'ne Mile Point Unit 2 FSA~
TABLE 15. 6-13 (Cont)
C ~
Control room Desian Basis Assumntions<<)
Realistic Basis Assanotions (1)
(2)
(3)
(4)
Pree air volume intake rate Recirculation rate Intake/recirculation halogen filtration efficiency 2.096+5 fta 1.00+3 cfm
- 7. 50+2 cfn 99%
I 3.81+05 ft3 1.50+3 cfm 7.50+2 cfm 99%
Disnersion data (s/ma) a.
b.
c Stack 0-2 hr EAB 0-8 hr LPZ 8-24 hr LPZ 24-96 hr LPZ.96-720 hr LPZ 0-8 hr control room 8-24 hr control roo 24-96 hr cont ol roon 96-720 hr control room Radvaste/reactor building vent< ~~
0-2 hr EAB 0-8 hr LPZ 8-24 hr LPZ 24-96 hr LP Z 96-720 hr LPZ 0-8 hr control roon 8-24 hr control room 24-96 hr control roon 96-720 hr control roon Main steam tunnel
- 2. 97-5
- 1. 03-5
- 8. 85-7
- 3. 66-7
- 1. 03-7 8.10-5 2.44-8
- 2. 10-8
- 1. 69-8 1 90-4
- 1. 78-5
- 1. 19-5
- 4. 93-6
- 1. 40-6
- 2. 13-4
- 1. 66-4 9 ~ 88-5 4.70-5 1 16-7
- 4. 32-7
- 3. 21-7
- 1. 69-7
- 6. 73-8
- 8. 10-5
- 2. 44-8
- 2. 10-8
- 1. 69-8
- 2. 19-5
- 6. 48" 6 N/A N/A N/A
- 2. 13-4 N/A N/A N/A 0-2 hr EAB 0-8 hr LPZ 8-24 hr LPZ 24-96 hr LPZ 96-720 hr L PZ 0-8 hr control room 8-24 hr control roon
- 1. 90-4 1 78-5
- 1. 19-5 4.93-6
- 1. 40-6 1.29-3 9.90"4 N/A N/A N/A N/A N/A N/A N/A Amendnent 21 10 of 11 September 1985
Nine Nile Point Brit 2 FSAE 24-96 hr control roon 96-720 hr control room d.
Padvaste tunnel (PASS a ea) 0-2 hr EAB 0-8 hr LPZ 8-24 hr LPZ 24-96 hr LPZ 96-720 hr LPZ 0-8 hr control room 8-24 hr control room 24-96 hr control roon 9 6-720 hr control room TABLE 15. 6-13 (Cont)
Design Basis Assumptions<<>
- 3. 37-4 9.92-5
- 1. 90-4
- 1. 82-5
- 1. 21-5
- 5. 02-6 1.42-6
- 1. 83-4
- 1. 4 1-4
- 4. 81-5
- 1. 42-5 Bealisti" Bas s
Assamntions N/A N/A H/A N/A H/A H/A N/A H/A H/A H/A N/A e;
SGTS building 0-2 hr LAB 0-8 hr LPZ 8-24 hr LP2 24-96 hr LPZ 96-720 hr LPZ 0-8 hr control room 8-24 hr control room 24-96 hr control room 96-720 hr control roon
- 1. 90-4
- 1. 78-5
- 1. 19-5 4.93-6
- 1. 40-6
- 1. 75-3
- 1. 34-3
- 4. 57-4 1 35-4 N/A N/A N/A N/A H/A N/A N/A N/A N/A 21 NOTE:
5.92-4
= 5.92 x 10-i Dragon 4 Code, Dose and Radioactivity from Nuclear Pacility Gaseous
- Oatflovs, HU-115, Version 4, Level 1, April 1982.
EAB dose analysis only.
See Section 6.2.3.
<<> These values are used for the 90-sec release directly to the environment daring the period vhen reactor building pressure is above -0.25 in Q.G.
Common to isentropic and isothermal cases.
Amendment 21 11 of 11 September 1985
Nine Mile Point Unit 2 FSAR TABLE 15.6-14 LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS)
ACTIVITY AVAILABLE FOR RELEASE FROM PRIMARY CONTAINMENT AT t = 0
~ieoto e
'I-129 I-131 I-132 I-133 I-134 I-135 I-136 Br-83 Br-84 Br-85 Br-87 Kr-83m Kr-8Sm Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m
'e-133 Xe-135m Xe-135 Xe-137 Xe-138 (Ci)
~Activit
- 7. 60-1 2.30+7 3.35+7 4.80+7 5.28+7 4.53+7 2.19+7 2.73+6 4.85+6 5.80+6 9.78+6 1.10+7 2.35+7 1.05+6 4.SO+7 6.38+7 7.95+7 5.51+5 8.06+6 1.93+8 3.63+7 2.49+7 1.69+8 1.61+8 Amendment 13 1 of 1
August 1984
I
Nine Mile Point Unit 2 FSAR TABLE 15.6-15a LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS)
ACTIVITY RElEASE TO ENVIRONMENT* (ISOTHERMAL APPROACH)
(Ci)
~Isoto e
I-129 I-131 I-132 I-133 1-134 I-135 I-136 Br-83 Br-84 Br-85 Br-87 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138
~ACt1vI't (LATER)
Total
- Total release for 30 days.
Amendment 21 1 of 1
September 1985
Nine Mile Point Unit 2 FSAR TABLE 15.6-15b LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS)
ACTIVITY RELEASE TO ENVIRONMENT* (ISENTROPIC APPROACH)
(Ci)
~Ieoto e
I-129 I-131 I-132 I-133 I-134 I-135 I-136 Br-83 Br-84 Br-85 Br-87 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Total
~Act1vit (LATER)
- Total release for 30 days.
Amendment 21 1 of 1
September 1985
I
Nine Mile Point Unit 2 FSAR TABLE 15.6-16a LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS)
RADIOLOGICAL EFFECTS (ISOTHERMAL APPROACH)
Exclusion area (2 hr)
Low-population zone (30 day)
Control room (30 day)
Whole-Body Thyroid Dose Dose
~Rem (LATER)
Beta Dose
~Rem Amendment 21 1 of 1
September 1985
t
Nine Mile Point Unit, 2 FSAR TABLE 15. 6-16b LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS ANALYSIS)
RADIOLOGICAL EFFECTS (ISENTROPIC APPROACH)
Whole-Body Thyroid Beta Dose Dose Dose Exclusion area (2 hr)
Low-population zone (30-day)
Control room (30-day)
( LATER)
Amendment 21 1 of 1
September 1985
P
0
~
Nine Mile Point Unit 2 FSAR TABLE 15.6-17 LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS)
REACTOR COOLANT IODINE CONCENTRATIONS
~Isoto e
I-131 I-132 I-133 I-134 I-135 Design Reactor Coolant
~uCi~m
- 1. 3-2 2.2-1 1.6-1 3.9-1 1.7-1 Normalized Reactor Coolant
- 6. 1-1 1.0+1 7.5+1 1.8+1 8.0+0 NOTE:
1.3-2
= 1.3 x
10
- Design reactor coolant iodines normalized to maximum technical specification limit of 4 uCi/gm.
Amendment 13 1 of 1
August 1984
4s
)
~
Nine Mile Point Unit 2 FSAR TABLE 15.6-18 LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS)
ACTIVITY AIRBORNE IN CONTAINMENT DUE TO IODINE SPIKING
~iscto c
I-131 I-132 I-133 I-134 I-135 (Ci)
Initial Airborne
~hctivit 8.33+4 1.41+6 1.02+6 2.50+6 1.90+6 Airborne Activity Available for Release
- 1. 67+4 2.82+5 2.04+5 5.00+5 2.18+5 NOTE:
8.33+4
=
8.33 x 10" Amendmeng 13 1 of 1
August 1984
i
Nine Mile Point Unit 2 FSAR TABLE 15.6-19 LOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS)
ACTIVITY RELEASE TO ENVIRONMENT
~Iooto o
I-131 I-132 I-133 I-134 I-135 Xe-131m*
Xe-133m+
Xe-133+
Xe-135m*
Xe-,135*
(Ci)
Activity Released 1.79+1 4.21+0 2.70+1 2.87+0 9.59+0 6.85+1 4.25+2 1.28+4 8.72+3 5.27+3 NOTE:
1.79+1
= 1.79 x
10'The xenon i sotopes vere produced by the decay of the iodines released.
Amendment 13 1 of 1
August 1984
c
Nine Mile Point Unit 2 FSAR TABLE 15.6-20 IOSS-OF-COOLANT ACCIDENT (REALISTIC ANALYSIS)
RADIOLOGICAL EFFECTS Whole-Body Dose Rem Thyroid Dose
~Rem Beta Dose
~Rem Exclusion area (2 hr)
Low-population zone (30 d)
Control room (30 d) e 1.53-4 4.65-4 3.75-4 1.41-2 3.54-5 5.83-3 2.14-4 1.48-3
. 4.42-3 NOTE:
1.53-4
= 1.53 x
10 Amendment 13 1 of 1
August 1984
gf V
/
t