ML20199L284
ML20199L284 | |
Person / Time | |
---|---|
Site: | Millstone ![]() |
Issue date: | 01/20/1999 |
From: | NORTHEAST NUCLEAR ENERGY CO. |
To: | |
Shared Package | |
ML20199L265 | List: |
References | |
NUDOCS 9901270190 | |
Download: ML20199L284 (20) | |
Text
-
.. ~. -
l MNPS 2 FSAR Assumotion (1)
Activity in Containment available for release 100% Noble Gases 25% lodines (2) initial lodine Chemical Form:
(w*- )
91% elemental 4% organic 5% particulate (3)
Purging Occurs 5 days after initiation of LOCA i
(4)
Breathing Rate = 2.32 x 104 m*/sec (5)
Power Level = 2700 MWt (6)
Dose Conversion Factors Reg. Guide 1.109 (7)
Purge Rate = 50 ft*/ min (8)
Containment Building Volume = 1.899 x 108 ft" (9)
Release Point Unit 1 Stock (10) Filter Efficiencies:
90% elementaliodine 70% organic iodine 90% particulate iodine (11) Duration of Purge = 30 days (12) X/O (sec/m*)
^
LPZ (0-30) days = 6.97 x 10
i 14.8.4 Radiological Consequences of the Design Basis Accident
.7A/fl6 0 4-14.8.4.1 General The DBA involves a oss release of activity from the fuel the containment building.
This section disc ses the consequences of such a rel e.
14.8.4.2 ethod of Analyses l
The diological consequences of a Desig asis LOCA at Millstone 2 were a yzed for w wind speed condition and a high nd speed condition. These are re esented by a
ases A and B, respectively.
l 9901270190 990120 PDR ADOCK 05000336-P P D R.__.
i MP214-8.MP2 14.8-11 June 1996l
l MNPS-2 FSAR
)
Case A.- Low Wind Speed Condition This case assumes meteorological conditions exist which will give 95 percent hi est X/O values (e.g., low wind speeds). For this scenario the activity which leak rom the containment building enters the enclosure building where it is treated by th enclosure building filtration system (EBFS) before being released through the Unit 1 ack. A small percentage (1.69 percent) of the containment leakage bypasses e EBFS and is released at ground level for the entire accident (30 days). All contai ent leakage for the first 110 seconds is assumed to be a ground level release. Thi is due to the fact that it takes 110 seconds for the enclosure building to achieve n gative pressure and thus assure that leakage will be into the enclosure building rat r than out. All assumptions used in this analysis are given in Table 14.8.4.
The radiological evaluation used thyroid dose conversio actors consistent with those stated in Regulatory Guide 1.109 Rev.1. In addition the Staff's acceptance of these dose "onversion factors in the published Safety Ev ation Report for Erie Nuclear Pcwer Plant (NUREG-0423),NNECO offers the fo wing justification:
(1)
Dose Conversion Factors The NRC has published a revi ed version of Regulatory Guide 1.109 (Octo-ber,1977) for use in Appe ix I calculations. The inhalation dose conver-sion factors (DCFs) cont ed in this guide are lower than those previously used in radiological ev stions to meet reactor siting criteria of 10CFR100.
The source of the i ine DCF previously used in radiological off site dose calculations is Tl -14844(March,1962). They were derived using the acute intake m el, a dose commitment of infinity and input parameters from ICRP-il. arameters from ICRP-Il such as effective half life, fraction of the isotope eaching the organ, and the effective energy released per disin-tegration ere based on the best available data at that time (1959). Due to
~
a lack o information, various conservatisms were employed in determining these arameters.
T DCF's in Regulatory Guide 1.109 were based on ICRP VI and X. They e a chronic intake model with a dose commitment extending for 50 years after intake. Credit is given for hold up of the nuclide in the lung before it reaches the thyroid. This has the effect of reducing the fraction of the nuclide reaching the organ of interest. The fraction was calculated based on information in ICRP X. The source of biological half lives was also based on ICRP X.
There is an appreciable difference between several factors upon which DCF's from TID-14844 and Regulatory Guide 1.109 were based. The first point of difference is the fraction (fa) of the nuclide which is deposited in the organ. As stated above, Reg. Guide 1.109 (Rev.1) takes credit for retention of the iodine isotopes in the lung. This is a more realistic ap-proach since it is expected that fewer of the short lived isotopes would be i
i able to reach the thyroid than those that are long lived. Another factor where there are differences is the effective energy deposited in the organ.
MP214 8.MP2 14.8-12 June 1996 l l.
l MNPS-2 FSAR The differences are only slight and are the result of different decay sc g
mes I
used to compute this parameter. The Reg. Guide values were based n
more recent data and hence are expected to be more accurate. It ould also be pointed out that the Reg. Guide values are slightly highe 'than ICRP 11 values which tend to increase the DCF's. The third fa or where differences occur is in the half lives. ICRP 11 reported a biot gical half life of 138 days while Reg. Guide 1.109 used a value of 100 d
- s. Once again the Reg. Guide value was based on ICRP X. It is not c at what the reason is for the difference in this factor. The differences i iological half lives is not critical since the effective half life is directly p portional to the DCF's.
The effective half lives do not significantly chan. The change in biologi-cal half lives, therefore, has little effect on DC s.
In summary, the major difference betwee the DCF's presented in TID-14844 and Reg. Guide 1.109 is the cre t taken by the Reg. Guide for hold up of iodine in the lung. TID-14844 sed its DCF's on the best informa-tion that was obtainable in 1959 w ereas those presented in the Reg.
Guide reflect the best informatio obtainable today. It is the conclusion of NNECO that the dose conversi factors in Reg. Guide 1.109 (Revision 1) are more applicable to offsite ose calculations and are therefore assumed in this dose analysis for Mil tons Unit No. 2.
(2)
Calculational Methods
)
In order to calculate fisite doses, the computer code TACT lit was em-ployed. This code valuates the activities, and integrated doses at a site following the ins ar'taneous or continuous release of halogens and noble gases from a c ntrol volume.
The input t the program consists of the time-dependent variables described below, th volume of the primary system, filter efficiencies, etc.
Eighte n isotopes are included in the model, including Kryptons, Xenons, and I dines. The isotope inventory may be input, or the program will cal late it based on TID source terms; decay is evaluated, as well as fi ration. The primary containment leak rate, atmospheric dispersion actors, and breathing rates may vary with time, at the option of the user.
Site dose calculations use the semi-infinite cloud dose models suggested by Regulatory Guide 1.4.
Case - High Wind Speed Conditions A
analysis was performed to determine the effect on the enclosure building of high and speeds. The wind speed at which the enclosure building will begin to exfiltrate is ne in which the corresponding wind velocity pressure is greater than the enclosure building negative pressure. The effect of wind on the enclosure building is discussed in Section 6.7.1.2.
The enclosure building filtration region (EBFR) design negative pressure is 0.25 in w.g.
The wind velocity corresponding to a velocity pressure equivalent to this EBFR design MP214-8.MP2 14.8-13 June 1996 {
_. - -. = _ _
1 i
MNPS-2 FSAR
/
pressur,e is 25 mph. Above this velocity displacement of the enclosure building f
atmosphere with outside air would begin. However, for conservatism it is assumed i
that this displacement would begin with a 23 mph wind.
The model formulated for site boundary and control room doses is as follows:
(1)
The wind is from the (plant) North direction.
(2)
The high wind condit;on exists for the first 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> follow' g the incident (5 percent of thirty days).
i (3)
Only those areas of the EBFR above grade are expose to the wind effects.
Therefore, only the enclosure building structure is s ject to air displace-l ment due to wind effects.
i (4)
The amount of postaccident containment leak e is assumed as 0.5 volume percent per day per technical specification
.1.
(5)
The amount of exfiltration based conser atively on a 30 mph wind is less than 10 percent of the EBFS exhaust pabilities assuming only one fan I
operating. Therefore,10 percent of he EBFR atmosphere is conservatively assumed to be an unfiltered groun release.
j All other assumptions and methodologies re the same for Case A.
14.8.4.3 Dose Calculations 14.8.4.3.1 Thyroid Doses and W te Body Exposures l
l The results of the calculated - ses for Cases A and B are shown in Table 14.8.4 2 and are within the limits of 10C 100.
14.8.4.3.2 Control Roo Habitability
^
As a result of Three ile Island (TMI) Action Plan item Ill.D.3.4, the potential radiologi-cal doses to the
-2 control room operators have been reevaluated. The analysis is based on the co rol room assumptions and meteorological parameters given in Tables 14.8.4-and 14.8.4-4.
The contro com is designed to be occupied for the duration of the accident (30 days).
Two (2) sic sources of radiation have been evaluated. They were: (1) direct dose from so rces outside the control room, and (2) the dose received from airbome activity whic nters the control room. The analyses ensure that the operators will be ade-qua ly protected from all sources of radiation.
e radiation design objective for the control room walls is to limit the whole body dose to personnelinside the control room to less than 5 rem during any DBA. The external sources considered in the shielding evaluation are: (1) containment, (2) enclosure /
j reactor building, (3) filtration systems, and (4) piping sources. The affect of external
/
sources from DBA's at Millstone Units 1 and 2 were evaluated in the shielding analysis.
MP214-8.MP2 14.8-14 June 1996 \\
MNPS-2 FSAR Because the containment as well as other sources at Millstone Unit No. 3 are separa i
d from the Unit 2 control room by a relatively large distance as well as other structu s, a shielding analysis from a Unit 3 accident was determined to be unnecessary.
The assumptions used in the shielding evaluation are listed in Table 14.8.4-i
. The resulting doses from the shielding analysis are given in Table 14.8.4-6.
An EBFS signal from Millstone 2 initiates control room isolation and af >r a 42-second delay the control room emergency ventilation system will be operati g at 2,500 cfm.
Isolation of the control room will be complete within 5 seconds af r reception of an isolation signal. During this 5 second interval the normal outsid air flowrate through the damper was assumed to vary linearly from 2,000 cfm to cfm. The MP2 control
)
room emergency ventilation system recirculates air from ins' e the control room and filters the air through high efficiency particulate air (HEPA and charcoal filters before returning it to the control room.
Two separate cases were analyzed for the design b is LOCAs at MP2. These cases are representative of high wind speed and low wir speed conditions. As described in Section 14.8.4.2 (Case B)it has been assumed at the high wind speed condition exists for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> af ter the LOCA snd 10 per ent of the activity in the enclosure building bypasses the EBFS resulting in a gro nd level release to the environment.
Displacement of the enclosure building at sphere would begin at wind speeds above 25 mph. However, for conservatism, it i assumed that this displacement would begin with a wind speed of 23 mph. The lo wind speed conditions used assumptions
)
consistent with those for Case 1 and iven in Table 14.8.4-1, except for the 1.69 percent bypass leakage. For te control room analysis, the bypass leakage was reevaluated without assuming a ismic event and was determined to be negligible.
The calculated whole body, b a and thyroid doses are presented in Table 14.8.4-8 and are below the General Desi Criterion 19 limits. For accidents at either Unit 1 or Unit 3 (and for several Un' 2 accidents not involving a signal to automatically activate the EBPS) the Unit 2 co rol room will not automatically isolat,e and must rely on a high radiation signal to perf rm the isolation.
Under normal cond' ions, air is provided to the control room operators by the air intake duct. The duct i equipped with redundant radiation monitors which will automatically isolate the cont ol room upon a high radiation signal. Approximately 23.1 seconds of 1
continuous u iltered air intake is assumed to enter the control room subsequent to isolation by signal from either radiation monitor. Af ter 42 seconds the control room emergenc ventilation system will be operating. Control room air will be recirculated through EPA and charcoal filters.
Sinc other operating reactors are located on the site, an assessment was made of the l
I ha ability of the Millstone 2 control room subsequent to an assumed design basis L CA at either Millstone 1 or 3. Assumptions used for each of these plants are given
' i their respective FSARs. Because of the close proximity of the Millstone 1 turbine building with respect to the Unit 2 control room intake duct, an assessment was also made of a steam line break (SLB) accident at Unit 1 on the Unit 2 operators. The assumptions used in this accident are given in Table 14.8.4-7.
MP214 8.MP2 14.8-15 June 1996 l
._._._.__..___m_.
l t
MNPS-2 FSAR I
N The calculated Millstone 2 control room dose from Millstone 1 and Millstone t
released
)
are presented in Table 14.8.4-8.
14.8.4.4 Conclusions
{
lt is concluded that the exclusion boundary and low population z e (LPZ) ' guideline dose values of 10 CFR 100 would not be exceeded even for th BA.
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MP214 8.MP2 14.8-16 June 1996 l
m INSERT D - FSAR Pane 14.8-11 14.8.4.1 General A LOCA would increase the pressure in the containment resulting in a containment isolation and initiation of the ECCS and containment spray systems. A SIAS signal automatically stans the Enclosure Building Filtration System (EBFS) which maintains a negative pressure within the enclosure building during accident conditions. The nuclide inventory assumed to be initially available for release from within containment consists of 100 percent of the core noble gasses and 25% of the iodines, as described in Regulatory Guide 1.4. A SIAS also isolates the control room by closing the fresh air dampers within 5 seconds. Within 10 minutes after control room isolation, the control room emergency ventilation (CREV) is properly aligned. CREV recirculates air within the control room through a 90 percent charcoal filter at 2,500 cfm ( 10%) to remove iodines from the control room envelope.
The radiological consequences of a Design Basis LOCA at Millstone 2 were analyzed for a low and high wind speed condition. The low wind speed case was found to bound the high wind speed case. Therefore only the low wind speed case will be presented here.
14.8.4.2 Release Pathways The release pathways to the environment subsequent to a LOCA are leakages from containment and the enclosure building, which are collected and processed by EBFS and leakages from containment 'and the RWST which bypass EBFS.
Containment Leakane The containment is assumed to leak at the design leak rate for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident.
After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, since the pressure has been decreased significantly, Regulatory Guide 1.4 allows for the leak rate to be reduced to one-half the design leakage rate.
All containment leakage for the first 110 seconds is assumed to bypass EBFS and is released directly out the MP-2 containment. This is due to the fact that it takes 110 seconds for EBFS to achieve the required negative pressure in the enclosure building, thereby ensuring that leakage will be into the enclosure building rather than out.
EBFS collects most of the containment leakage and processes it through IIEPA and charcoal filters and re: eases it up the Unit I stack. All containment leakage is collected and l
filtered by EBFS except for the small amount that is assumed to bypass EBFS and is j
released directly out the MP-2 containment.
l Credit is taken for iodine removal due to containment sprays. The sprays are efTective after 101 seconds post-LOCA. The effectiveness of the sprays in removing elemental i
iodine ends at 0.715 hours0.00828 days <br />0.199 hours <br />0.00118 weeks <br />2.720575e-4 months <br /> and in removing particulate iodine at 1.58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br />.
i l
i i
l
ESF System Leakage Pathway Post-accident radioactive releases from the ESF system are derived from fluid leakages assumed during recirculation of the containment sump water through systems located outside containment. The nuclide inventory assumed to be available for release from this pathway consists of 50% of the core iodines. The quantity ofleakage is based on the assumption that the ESF equipment leaks at twice the maximum expected operational leak rate and that 10 percent of the iodine nuclides contained in the leakage fluid become airborne in the enclosure building. The nuclides which become airborne are collected and released to the environment through EBFS to the Unit I stack.
1 RWST Backleakane Pathway Post-accident radioactive releases from the ECCS system are a result of ECCS subsystems containing recirculated sump fluid backleaking to the RWST. The backflow rate to the RWST, as a result ofisolation valve leakage, is pre-defined and time dependent. Due to this time dependency, the contaminated sump fluid from backleakage does not enter into the RWST until 25.45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> post-LOCA. Since the RWST is vented to atmosphere, the release is a result of the breathing rate of the RWST due to solar heating. The EAB dose is a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose therefore it is not affected by backleakage.
14.8.4.3 Control Room IIabitability The radiation design objective of the control room is to limit the dose to personnel inside the control room to 5 rem whole body, or its equivalent, during a DBA. The potential radiation dose to a control room operator is evaluated for the LOCA. The analysis is based on the assumptions and meteorological parameters (X/Q values) given in Tables 14.8.4-3 and 14.8.4-4.
The control room is designed to be continuously occupied for the duration of the accident, 30 days. Two basic sources of radiation have been evaluated: leakage of airborne activity into the control room from sources described in 14.8.4.2 and direct dose from sources outside the control room. The control room shielding serves to protect the operators from direct radiation due to the passing cloud of radioactive efiluent assumed to have leaked from containment, enclosure building and the RWST. The control room walls also provide shielding protection for radiation emanating from the CREV filters and containment shine.
A SIAS from Millstone 2 initiates control room isolation within 5 seconds by securmg the fresh air intake dampers. Within 10 minutes CREV is in operation recirculating air in the control room envelope through 90% efficient charcoal filters to remove radioactive iodines from the atmosphere. The calculated thyroid, whole body and skin doses from a Millstone 2 LOCA are presented in Table 14.8.4-8 and are below the General Design i
Criteria 19 limits.
i Normally outside air is provided to the control room via an air intake duct, which is equipped with redundant radiation monitors. These radiation monitors isolate the control
room within 10 seconds after a high radiation signal. This method ofisolation will occur aflei a Millstone 3 LOCA. The calculated thyroid, whole body and skin doses from a Millstone 3 LOCA are presented in Table 14.8.4-8 and are below the General Design Criteria 19 limits.
14.8.4.4 Dose Computation The radiological off-site dose consequences resulting from a postulated Millstone 2 LOCA are reported in Table 14.8.4-2. The off-site dose analysis show that the consequences to the EAB (0 - 2 hr) and LPZ (0 - 30 day) are less than the limits of 300 rem thyroid and 25 rem whole body as specified in 10CFR100. The assumptions used to perform the radiological analysis are summarized in Table 14.8.4-1.
14.8.4.5 Conciasion Analysis shows that the off-site radiological consequences are within 10CFR100 guidelines and the control room radiological consequences are within GDC19 criteria.
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l MNPS-2 FSAR i
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TABLE 14.8.4.-1 g/9e LOSS OF COOLANT ACCIDENT (oEF 5.77(
/tSSum PTIy-)
l Assumotion 2959 (1)
Core power level = -NOG MWt
' 2:
C;m... ;... &. - ;y..
3
($)
Core released fractions: Noble gases = 100%,lodines = 25%
3 (4)
- ; ce;..; composition:
91% elemental Tod,<1 e 4% organic 5% particulate 4
(ilI)
Reactor building le' k rate:
.5%/ day.<_24 hrs.
a
.25%/ day > 24 hrs.
S (p)
Enclosure Building Filtration System, charcoal filter efficiencies:
90% for elemental 70% for organic
)
90% for particulate
/,006-0V 3,c,6 E o y L
0.9p'),
L W E~o5
(/)
Bypass leakage fraction = ".S9 C
% 80 g.oS
- 3. oV E-o&
- 2. 3 / E-og 9
pidan 7;me 9.1Q E-04
/, be f-oS (g)
EBFS r.0;;ct";c p:cccurc irMohr, = 110 seconds 8
/,OY E d6 7.25'E-06 R 4.3 E.cy 9..D E-oG
($)
X/Os. -
Location Time Period Elevated Ground Release
-6B EAg (0-2) hrs.
03 x 10 39 x 10 LPZ (0-4) hrs.
3.1x1
~5 2.2x1 4 (4-8) hrs.
1.7 0*
2.1 x 04 (8 24) hrs.
2.6 10 4.76 104 4
(24-96) hrs.
1.
x 3.
I x 104 4
(96-720) hrs.
.97 x 1 *
.3x1
- 9
(%) Thyroid Inhalation DCFs from P.cg. CuMc ' 10e.re g/)30
/0 (H) Containment Free Air Volume = 1.899 x 10 ft 8
8 l
II (42) Breathing Rates (0-8) hr. = 3.47 x 104 m*/sec t
l (8-24) hr. = 1.75 x 104 m /sec 3
I i
(24-720) hr. = 2.32 x 104 m*/sec
~
4 JnSERT /)
14ss4.i.w2 1 of 1 October 1994 l
l 1
INSERT A l
ssumption
- 12) Release Points:
Filtered - MP-1 Stack I
Bypass - MP-2 Containment
- 13) Containment Sprayed Volume:
75.08 %
l l
- 14) Containment Spray Removal Coefficients:
elemental = 20 per hour particulate = 3.03 per hour i
- 15) Containment Spray Effectiveness Time: elemental: 101 seconds - 0.715 hours0.00828 days <br />0.199 hours <br />0.00118 weeks <br />2.720575e-4 months <br /> particulate: 101 seconds - 1.58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> 16)ESF Leakage: 24 gallons per hour 17)ESF leakage begins at 25 minutes post LOCA i~
- 18) Sump Volume: 2.86E+5 gallons i
- 19) RWST Backleakage begins at 25.45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> i
20)RWST Backleakage amount: 0.01 - 0.19 gpm l
l
- 21) lodine DF: 100 l
f t
i
MNPS-2 FSAR TABLE 14.8.4-2
SUMMARY
OF DOSES FOR LOSS OF COOLANT ACCIDENT T/4c DOSE (rems)
~
/
/
[
CASE A
[
CASE B ORGAN SITE N
SITE BOUNDARY HYDROGEN PURGE J DROGEN PURGE BOUNDA
/
NOT INCLUDED
/ NOT INCLUDED
/
Thyroi[
151
- 56.3
[
[2 6.1 W
e Body 3.8 1.4 [
[-
1.4
/
/
/
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lwk.le M l
Thy rolc) y f
^
gg 3,VC f +0/
D.3(oftoo I
Zpg l,31 E +o t 9.08 f*0!
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icso4-2 m e:
1 of 1
_ _ _ _ October 1994l
~
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MNPS 2 FSAR i
TABLE 14.8.4 -3
$055 of c:stA"r
[ CONTROL ROOM ASSUMPTIONS) 519e A ccroper
- 1., Control Room Volume = 7.7 x 10* f:' 3,5 (,5 f i'o y f f
/30 2.
Control Room Unfiltered inleakage in Recirculation Mode = +06 cfm
\\
100 j
- 3. Control Room Normal Makeup Air Flowrate = -iih000 cfm AP-3 tocs9 4.
Time from 9M c' AccMert to Time when Dampers Close = 5 sec.
i Timr 4' rom hat < I hw hsg A reJs.% A Tome wAia. Ayen case = /0 s ec.
- 5. Time when Control Room Emeroency Ventilatio g,p,9 System Operating at Full Speed = 42 cce.
(iFsitrdice) 1
.;Wfo
/ 0 '* "
- 6.. Control Room Emergency Ventilation 4 system Flowrate = -ib500 cfm
- 7. Charcoallodine Filtdr Efficiency = 90 percent i
f
- NOTES: For the lysis of assumed LOCA at MP2. or accidents at ehher Unit 1 or 3 damper closure e is 23.1 sec. to account for m tor response and damper closure time. Other nit 1 and Unit 3 assumptions are ven in the following references.
REFE CES:
. Unit 1 - W. G. Counsil,1981 (
SCO) Letter to D. M. Crutchfield (NR, transmitting '
Millstone Nuclear Power Sta 'on Unit 1 Systematic Evaluation Pro m,Section XV, l
Topics: Design Basis Eve s.
- 2. Unit 3 - Millstone U No. 3 FSAR.
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October 1994l
MNPS 2 FSAR TABLE 14.8.4 -4 ATMOSPHERIC DISPERSION DATA FOR MILLSTONE UNIT 2 CONTROL ROOM N
l m
MP-2 CONTRO OOM X/Qs (sec./m*)
Release Point MP1 MP-2 MP-3 a.
Groun evel (0-8) 4.43 x 10'8 2.69 x 10 *
.78 x 104 (8-4) hr.
3.05 x
-2 1.90 x 10~8 3.29 x 10d j
1-4) day 1.
x 10'8 7.56 x 10d 1.20 x 104
==m.
(4-30) day
.68 x 10 8 2.30 x 104 2.26 x 10 5 (0-24) hr.
N/A 9.14 x
~'
N/A j
(24-36) hr.
N/A 9.
x 104 N/A
- b. Elevated Rete e - Unit 1 Stack (0-8) hr.
2.0 x 10-'
2.0 x 10*
N/A (8-24 r.
1.0 x 10-'
1.0 x 10*
N/A
)
(1- ) day 2.5 x 10
2.51 x 10
N/A (4-30) day 5.01 x O'"
5.01 x 10'"
N/A
~
x-r e l
14564 4 MP2 1 Of 1 October 1994\\
___._.--____.__.____.~._..__.~.._~__.y L
j r
i INSERT B i
Release Point l
l MP-2 RWST:
1 0 - 8 HR:
1.87E-3 8 - 24 HR:
1.20E-3 1 - 4 DAYS: 3.83E 4 - 30 DAYS: 5.81E-5 MP-2 Containment:
0 - 8 HR:
3.07E-3 8 - 24 HR:
2.09E-3 1 - 4 DAYS: 7.42E-4 1
4 - 30 DAYS: 1.93E-4 l
l 0 - 4 hrs:
2.51 E-4 MP-1 Stack 4 - 8 hrs:
! 8 - 24 hrs:
5.46E-6 i
24 - 96 hrs:
2.06E-7 96 - 720 hrs: 2.58E-9 i
i I
MP-3 Containment
.i 0 - 8 hr:
9.19E-4
)
8 - 24 hr:
5.29E-4 24 - 96 hr:
1.65E-4 96 - 720 hr:
2.75E-5 MP-3 Ventilation Vent 0 : 8 hr-1.25E-3 l
J - 24 hr:
7.49E-4 1
24 - 96 hr:
2.46E-4 96 - 720 hr:
4.08E-5 MP-3 MSVB 0 - 8 hr:
2.47E-3 l
8 - 24 hr:
1.48E-3 24 - 96 hr:
4.87E-4 j
96 - 720 hr:
8.1BE-5 i
2.08E-3 8 - 24 hr:
1.18E-3 24 - 96 hr:
3.88E-4 96 - 720 hr:
6.12E-5 I'
o MNPS-2 FSAR TABLE 14.8.4 5 ASSUMPTIONS USED TO CALCULATE DOSES FROM EXTERNAL SO Elio CES 1.
Assumptions Used to Calculate Dose from Containment Source:
a.
Source Term:
100 percent core noble gas inventory eleased to containment 50 percent core iodine inventory eleased to containment.
b.
Source assumed to be uniformly distributed to ontainment free air volume.
c.
Containment Volumes:
Unit 1 = 2.568 x 10 ft*
5 Unit 2 = 1.899 x 108 ft*
d.
Containrn' ent Concrete Wall Thic ess:
Unit 1 = 5' Unit 2 = 3.75' (walls). 3' ( ne) e.
Control Room Concrete all Thickness:
)
For wall facing Unit containment = 3'-6" 1
For wall facing Uni 2 containment = 2'-0*
2.
Assumptions Used to alculate Dose from Enclosure / Reactor Building Source:
l a.
Containmen Leak Rate:
1 Unit 1 = 1.2 percent / day Unit 2 0.5 percent / day b.
Vol e of Enclosure / Reactor Building nit 1 Reactor Building = 1.728 x 10' ft*
Unit 2 Enclosure Building = 1,44 x 10' ft*
c Ventilation Rate Unit 1 Reactor Building = 100 percent / day Unit 2 Enclosure Building = 6.000 cfm d.
Control Room Wall Thickness:
Control Room Wall Facing Enclosure Building = 2'-0" Control Room Wall Facing Reactor Building = 3'-6" uss44.w:
1 of 2 October 1994 k
o MNPS 2 FSAR TABLE 14.8.4-5 ASSUMPTIONS USED TO CALCULATE DOSES FROM EXTERNAL SOURC 3.
Assumptions Used to Calculate Dose from Filtration Systems:
Filtration Systems Considered:
a.
Millstone Unit 1 Standby Gas Trr.atment System GTS)
Millstone Unit 2 Enclosure Building Filtration Sy em (EBFS)
Millstone Unit 2 Control Room Filters b.
Thickness of Concrete Between Control Roomd:
Millstone Unit 1 SGTS = 9'-0" Millstone Unit 2 EBFS = 18'-0" Millstone Unit 2 Control Room Filt
= 2 '-0 "
4.
Assumptions 0 sed to Calculate Dose from verhead Plume:
Millstone Unit 2 Control Room C ~ ing Concrete Thickness = 2'-0" a.
b.
Filtration System Filter Effici cies:
SGTS = 90 percen (all forms of iodine)
)
EBFS = 90 perc t (elemental and particulate iodine) 70 per ent (organic iodine)
=
c.
Plume Centerline /Os:
(0-8) hr. = 4.
x 10~* sec/m*
(8 24) br.
4.19 x 104 sec/m*
(1-4) d
= 1.65 x 104 sec/m*
(4-
) day = 9.92 x 10'8 5.
Assu ptions Used to Calculate Dose from Piping Sources:
Sources in the vicinity of the Unit 2 Contro! Room = Unit 1 core spray line.
a.
Source Term:
50 percent core iodine inventory 1 percent solid fission products Concrete thickness of Control Room wall = 3'-6" c.
i 34ss44 m n 2 of 2 October 1994 l
l.,
i MNPS-2 FSAR I
l TABLE 14.8.4 6
)
SUMMARY
O'F DOSES FROM EXTERN OURCES (1) 3/p l
6 30-DAY DOSE (MREMS)
MILLSTONE UNIT 2 4
CONTROL R M OPERATORS UNIT 1 ACCIDENT OVERHEAD ECT FROM SECONDARY CORE SPRAY TYPE PLUME IMARY CONT.
CONTAINMENT LINE TOTAL l
4
. 88 x 10~'
3.493 x 10" 4.846 x 10' 5.972 x 10' j
it 1 LOCA 4
- 4. 44 x 10' 5.214 x 102 it 2 CA 9.292 x 108
/
)
i c
\\
}
l 14 sed o.MP2
)g October 1994 l
2 c
MNPS-2 FSAR TABLE 14.8.4-7 ASSUMPTIONS USED IN A MILLSTONE UNIT 1 MAIN STREAMLI BREAK SM 1.
Mass of Coolant / Steam Released = 1.753 x 10'gm 2.
Coolant Concentration:
DEQ (I-131) = 0.2 micro Ci/gm i
Noble Gas = 100/E micro Cilgm 3.
Duration of Release = 5.5 sec.
4.
MP 2 Control Room Normal Ventilat' System Flowrate = 2,000 ft / min. (Note:
8 This flowrate assumed for entire closure = 23.1 sec.)
sec since monitor response and damper 5.
Time for MP-2 tontrol Roo Ventilation System to Operate at Full Speed = 42 sec.
6.
MP-2 Recirculation stem Flowrate = 2,500 cfm 7.
Time When 0 ators are Assumed to Don Scott Air Paks = 20 minutes 8.
Effective
}
s of Scott Air Paks = 10,000 9.
Time en Operators are Assumed to Purge the Control Room = 30 min.
- 10. P ge Time Span = 30 min. to 4 hrs.
- 11. Purge Flowrate = 16,500 2.
Control Room X/O (0-8 hrs.) = 4.43 x 10~'sec/m' 14ss4 7.un 1 of 1 October 1994 l
- _ ~ -.-..
o MNPS 2 FSAR f
TABLE 14.8.4g ska DOSE TO MILLSTONE UNIT 2 CONTROL ROOM OPERATORS 1
t l
i Whole Bodyll)
Beta Thyroid Dose Gamma Dose Skin Dose Release (Rems)
(Rems)
_ (Rems)
IMillst e 1 (LOCA)(2) 2.27 x 1 7.58 x 10-3 2.45 x 10
L illstone 1 (MSLB) 2.
x 10' 4.23 10
5.
x 100 Millstone 2 (LOCA 0.25 m 10"
-0.45 s 10" 2 G5 x iG '
i Le.^..u Sp J Ce..d;;;;..;;
g,y)xf,'
9,55 y/o #
y, gryfo*
Wind R eed ions)
Millstone 3 (LOCA)
-2.48 a 10' 2.00 x 10" 2.07. 10^
2.IS X/o'
/.4Vt/*/o
/. Yo x'to '
NOTES:
(1)
Dose through wall and ceiling from external sources included.
(2) time of 8 hou is assumed for urbine Building E ust to b j
initiated.
I N
l I
l'
(
(
)
I 14&B4 8.MP2 1 of 1 October 1994 l