ML20148P408
| ML20148P408 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 07/30/1996 |
| From: | Mccullough R VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20148P392 | List: |
| References | |
| PROC-960730, NUDOCS 9707020222 | |
| Download: ML20148P408 (27) | |
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VERMONT YANKEE NUCLEAR-POWER STATION 0FF-SITE DOSE. CALCULATION MANUAL i
l REVISION 20 1
. Reviewed:
b CC,1/ b.
!y, MMH 4
9f2G[94, Plant Operations lJeviep Comittee date Approved:
(
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7 2.Th P4n(g)P ' '
Date Approved:
As OA ha 7'P'W
.. 7 Vice president, Dperations Date 970702O222 970627 -
l PDR ADOCK 05000271:
R PDR l
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Revision 20 Date 07/30/96
-iii-I
n 3.7 Method to Calculate the Gamma Air Oose from Noble Gases..
3-30 3.8 Method to' Calcuiate the Beta Air Oose from Noble Gases 3-34 3.9 Hethod to Calculate the Critical Organ Dose from Iodines.
Tritium and Particulates 3-38
~
3.10 Receptor Points and Annual Average Atmospheric Dispersion Factors for Important Exposure Pathways..........
3-45 3.11 Method to Calculate Direct Dose From Plant Operation 3-51 3.12 Cumulative Doses 3-59 4.0 ENVIRONMENTAL HONITORING PROGRAM 4-1 5.0 SETPOINT DETERMINATIONS...................... 5-1 5.1 Liquid Effluent Instrumentation Setpoints..........
5-2.
5.2 Gaseous Effluent Instrumentation Setpoints 5-11 l
6.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION HONITORS.'AND RADWASTE TREATHENT SYSTEMS 6r1 6.1 In-Plant Radioactive Liquid Effluent Pathways........ 6-1 6.2 In-Plant Radioactive Gas,eous Effluent Pathways 6-4 REFERENCES R-1 APPENDIX A: Hethod I Example Calculations A-1 1
APPENDIX B: Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee B-1 APPENDIX C: Response to NRC/EG&G Evaluation of ODCH Update Through Revision 4 C-1 APPENDIX 0: Assessment of Surveillance Criteria for Gas Releases from Waste Oil Incineration 0-1 APPENDIX E: NRC Safety s'alluation for Disposal of Slightly Contaminated Soll On-Site at VY (Below the Chem Lab Floor) - TAC No. H82152 E-1 Revision 20 Date 7/30/96
-vil-
TABLE 1.161
~ ~
(Continued)
Summary of Radioloof cal Effluent ' Technical Specifications-and imolementino Ecuations Technical Speciffcation Category Methodm
. Limi t 3.9.A.1 Liquid Effluent Monitor
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Setpoint Liquid Radwaste Alarm Setpoint Eq. 5-1 T.S. 3.8.A.1 1
Discharge Monitor s
3.9.B.1 Gaseous Effluent Monitor Setpoint Plant Stack and A0G Alarm / Trip Setpoint for Eq. 5-9 T.S. 3.8.E.la Offgas System Noble Gas Total Body Dose Rate (Total Body)-
Activity Monitors Alarm / Trip Setpoint for Eq. 5-10 T.S. 3.8.E.la Skin Dose Rate (Skin)
SJAE Noble Gas Activity Alarm Setpoint Eq. 5-21 T.S. 3.8.K.1 Monitors (1)
More accurate methods may be available (see subsequent chapters).
(2)
Technical Specification 3.8.M.2 requires'this evaluation only if twice the limit of Equations 3-1, 3 3, 3-21, 3-23, or 3-25 is reached.
If this occurs a Method 11 calculation shall be made considering available information for pathways of exposure to real individuals from liquid, gaseous, and direct radiation sources.
L b
Revision 20 _ Oate 7/30/96.
1-5 i
TABLE 1.1-12 Dose and Dose Rate Factors Specific for Vermont Yankee for iodines. Tritium. and Particulate Releases Stack Release
$round Level Release
- Critical Organ Critical Organ Critical Organ Critical Organ Dose Factor Dose Rate factor Oose Factor Dose Rate factor mrem' mremnec '
mrem' Radio-0FG,gco 0FG,,c, 0FG,,co 0FG emqc[
Ci yr-C1 Ci gic,, yr-pC1 nuclide 5
j j
H-3 3.13E-04 9.87E-03 1.06E-02 3.34E-01
)
C-14 1.90E-01 5.99E+00 6.43E+00 2.03E+02 Cr-51 6.11E-03 2.11E-01 4.16E-02 1.43E+00 Hn 7.01E-01 2.77E+01 4.71E+00 1.84E+02 Fe-55 3.17E-01 1.00E+01 2.05E+00 6.47E+01 Fe-59 6.99E-01 2.32E+01 4.60E+00 1.52E+02 Co-57 2.18E-01 8.23E+00 1.41E+00 5.33E.L1 l
Co-58 3.62E-01 1.30E+01 2.39E+00 8.52E+01 Co-60 7.63E+00 3.41E+02 4.99E+01 2.16E+03 Zn-65 3.71E+00 1.20E+02 2.36E+01 7.63E+02 Se-75 2.41E+00 7.76E+01 1.53E+01 4.92E+02 1
Sn-113 1.03E+00 3.25E+01 6.58E+00 2.08E+02
'l Sr-89 1.14E+01
.3.60E+02 7.27E+01 2.29E+03 Sr-90 4.31E+02 1.36E+04 2.82E403 8.89E+04 Zr-95 6.91E-01 2.28E+01 4.51E+00 1.49E+02 Sb-124 1.26E+00 4.23E+01 8.35 E+00 2.79E+02 l
Sb-125 1.25E+00 4.89E+01 8.01E+00 3.13E+02 I-131 7.71E+01-2.43E+03 5.02E+02 1.58E+04 l
I-133 8.22E-01 2.59E+01 8.30E+00 2.62E+02 l
Cs-134 1.58E+01 5.27E+02-1.02E+02 3.37E+03 Cs-137 1.63E+01 5.55E402 1.04E+02 3.53E+03 Ba-140 1.13E-01 3.66E+00 2.18E+00 6.94E+01 Ce-141 1.70E-01 5.42E+00 1.19E+00 3.78E+01 Ce-144 3.85E+00 1.22E+02 2.52E401 7.98E+02 The release point reference is the North Warehouse. These dose and dose rate factors are conservative for potential release applications associated with ground level effluents from other major facilities (i.e.. Turbine Building Reactor Building. A0G and CAB).
Revision 20 Date 7/30/96 1-22
I sT Atbs -
0.61
][
Qg
- OF8, (3-5)
L 1
~
l.
(
r 1
Parem' pCi-sec
'pCi '
mrem-m 3 Mr >
pC1-m
- sec, pCi-yr,
s u
where:
dji In the case of noble gases, the release rate from the plant stack (pCi/.sec) for each radionuclide. "i", identified. The release rate at the plant stack is based on measured radionuclide concentrations and distributions in periodic grab-samples taken at the stack.
As an alternative method the radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) can be used during plant operations, along with the Stack Gas Monitor effluent count rate, to estimate stack radionuclide r'eleases. The release rate at the stack when using SJAE samples can be stated as follows:
gsJAE D i H
~
sJAE (3-28) i uCi gi/cc'(cc)
(cpm) seC Cpm sec j
H Plant Stack Gas Monitor I or II count rate (cpm).
=
S, Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpm /(pCi/cc)).
F Stack flow rate (cc/sec).
6fJAE The last measured release rate at the steam jet air
=
ejector of noble gas i ( Ci/sec).
I i
- 0FB, Total body gamma dose factor (see Table 1.1-10).
j i
l Revision 20 Date 7/30/96 3-14
skins "
b
~
O F'5 i
i (3-7) fmrem' f Ci '
' mrem-s ec '
yr,
- sec, pCi-yr where:
jT In the case of noble gases, the noble gas release rate from the plant stack (pCi/sec) for each radionuclide. "i", identified.
The release rate at the plant stack is based on measured radionuclide concentrations and distributions in periodic grab samples taken at the stack.
As an alternative method, the radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) can be used during plant operations, along with the Stack Gas Monitor effluent count rate, to estimate stack radionuclide releases.
The release rate at the stack when using SJAE samples can be stated as follows:
gsJAE i
N gsdAE (3-28) i pCi
( Ci/cc)
(cc)
(cpm) sec cpm sec H
Plant stack gas monitor I or II count rate (cpm).
S, Appropriate or conservative plant stack monitor detector counting cfficiency for the given'nuclide mix (cpm /( Ci/cc)).
F Stack flow rate (cc/sec).
i 6fJAE The las.t measured release rate at the steam jet air ejector of noble gas i (pCi/seci.
0F,
i combined skin dose f actor (see Table 1.1-10) for stack release.
Revision 20 Date 7/30/96 3-20
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3.11 Method to Calculate Direct Oose From Plant Operation j
Technical Specification 3.8.H.1 restricts the dose to the whole body or any organ to any member of the public from all station sources (including direct radiation from fixed sources on-site) to 25 mrem in a calendar year (except the thyroid. which is limited to 7S mrem).
3.11.1 Turbine Buildine
~
The maximum contribution of direct dose to the whole body or to any organ due to N-16 decay from the turbine is:
E (3-27)
O "KN16(L) d I"#8"I (mrem)
(NW,h)
MW,h where:
Od The dose contribution from N-16 decay at either the site boundary of maximum impact _(west site boundary) or closest off-site residence - (mrem).
I E
Gross electric output over the period of interest (MW,h),
KN16(L)
The N-16 dose conversion factor for (L) equal to either:
=
(1) 3.17E-06 for the maximum west site boundary; or (2) 1.26E-06 for the closest residence (mrem /MW,h).
3.11.2 North Warehouse Radioactive materials and low level waste can be stored in the north warehouse.
The maximum annual dose contributions to off-site receptors (west site boundary line) from sources in the shielded (east) end and the unshielded (west') end of the north warehouse are:
tevision 20 Date 7/30/96 3-51
fgy The fraction of a year that the intermodular gap is no,t shielded.
2.44E-2 The activity to site boundary dose conversion factor mrem for a one-inch wide intermodular gap y r-i n -C,i j
The site boundary dose from waste materials placed into storage on the low level Waste Storage Pad Facility is determined by combining the dose contribution due to direct radiation (line of sight) from Part (a) above with the skyshine scatter dose from Part (b) resin liner transfer dose from Part (c), and any intermodular gap Cose from Part (d).
j 3.11.4 -Total Direct Dose Summary The dose contributions from the N-16 source in the Turbine Building, fixed sources in the north warehouse, and fixed sources on the Low Level Waste Storage Pad Facility, shall be combined to obtain the estimate of total off, site dose to any member of the public from all fixed sources of radiation located on-site.
3.11.5 Other Fixed Sources In addition to the fixed sources noted above (Turbine Building. North Warehouse, and LLW Storage Pad), other identified temporary or fixed sources that are created due to plant operations will be included in the total direct summary of 3.11.4 ff the projected annual dose contribution would add any notable addition to the reported total (i.e.,10.1 mrem /yr).
In 1995 turbine rotors and casings were replaced in the Turbine Hall with the old rotors and casings placed in storage sheds located on site west of the switchyard along the railroad spur.
Radiation surveys (December 1995) of low level contamination (principally Co-60) on the components led to a projected maximum west site boundary dose of 0.2 mrem /yr. This contribution will be added to the maximum site boundary total dose until the contribution is less than 0.1 mrem /yr, or the components are removed from storage location.
Revision 20 Date 7/30/96 3-58
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4.0 ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring stations are listed in Table 4.1.
The locations of the stations with respect to the Vermont Yankee
- plant.are shown on the maps in Figures 4-1 to 4-6.
- 4.1 =Intercomparison Program All routine radiological analyses' foi environmental samples are performed at the Yankee Atomic Environmental Laboratory (YAEL). The YAEL f
participates in several government' and commercial Intercomparison quality assurance programs (0APs) that are traceable to the National Institute of 1
. Standards and Technology [NIST). These include:
Nuclear' Energy Institute (NEI)/NIST Measurement Aswrance Program for the Nuclear Power Industry: U.S.
Environmental Protection Agency's Intercomparison Program for Drinking Water:
j and Analytics, Inc. (commercial). YAEL also participates in the OAP for
[
environmental media. conducted by the Environmental Measurements Laboratory of 3
the U.S. Department.of Energy.
4.2 Airborne Pathway Monitoring The environmental sampling program is designed to achieve several major objectives, including sampling air in predominant up-valley and 'down-valley wind directions, and sampling air in nearby communities and at a proper.
control location, while ma,intaining continuity with two years of preoperational data and 18 years of operational data (as of 1990). The chosen air sampling locations are discussed below.
To assure that'an unnecessarily frequent ielocation of samplers will not be required due to short-term or annual fluctuations in meteorology, thus incurring. needless expense. and destroying the continuity of the program, long term, site specific ground level 0/Os (five-year averages - 1978 through 1982) j were evaluated in comparison to,the existing air monitoring locations to determine their adequacy in meeting the aboy'e'-stated objectives of the program and the intent of the NRC general guidance. The long-term average meteorological data base precludes the need for an annual re-evaluation of air sampling locations based on a single year's meteorological history.
The Cc"necticut River Valley in the vicinity of the Vermont Yankee plant has a pronounced up-and down-valley wind flow. Based on five years of meteorological data, wind blows into the 3 "up valley" sectors (N. NNW, and NW) 27 percent of the time, and the 4~"down-valley" sectors (S. SSE. SE and ESE) 40 percent of the time, for a total "in-valley" time of 67 percent.
Revision 20 Date 7/30/96 4-1
Station AP/CF-12 (NNW. 3.6 km) in North Hinsdale, New Hampshire. monitors the up-valley sector's.
It,is located in the sector that ranks fourth overall in terms of wind frequency (i.e...in terms of how often the wind blows into that I
sector), and is approximately 0.5 miles from the location of the calculated caximum ground level D/0 (i.~e..
for any location in any sector, for _the entire
)
Vermont Yankee environs). This station provides a second function by its location in that it also monitors North Hinsdale, New Hampshire, the com'munity with the second highest ground level D/0 for surrounding communities, and it has been in operation since the preoperational period.
The down-valley direction is monitored by two stations - at River
~ Station Number 3.3 (AP/CF-11. SSE. 1.9 km) and at Northfield, Massachusetts (AP/CF-14 SSE. 11.3 km)." They both reside in the sector with the maximum wind frequency and they bound the down-valley point of. calculated maximum ground level D/0 (the second highest overall ground level 0/0 for any location in acy sector). Station AP/CF-11 is approximately one mile from this point, between it and the plant.
Station AP/CF-14 also serves as a community monitor for Northfield, Massachusetts.
Both stations have been in operation since the preoperational period.
i 1
In addition to the up-and down-valley locations, two communities have
'l been chosen for community sampling' locations. The four nearest population j
groups with the highest long-term average 0/0 values, in decreasing order, are J
Northfield, ^tssachusetts, North Hinsdale, New Hampshire. Brattleboro, Vermont, and slinsdale, New Hampshire. The community sampler for Northfield is at Station AP/CF'14 (mentioned above). ' North Hinsdale is already monitored by the up-valley station (AP/CF-12. NNW, 3.6 km), which also indirectly monitors the city of Brattleboro, located further out in the same sector. The second sampler specifically designated for a community is at Hinsdale Substation (AP/CF-13, E, 3.1 km) in Hinsdale.
The control air sampler was located at Spofford take (AP/CF-21 NNE, '16.1 km) due to its distance from the plant and the low frequency for j
wind blowing in that direction based on the long-term (five-year) meteorological history.
Sectors in the general west to southwest direction, which would otherwise have been preferable due to lower wind frequencies, were not chosen since they approached the region surrounding the Yankee Atomic
]
plant in Rowe, Massachusetts.
An additional air sampler is maintained at the Tyler Hill site (AP/CF-15. WNW, 3.4 km), which is along the western side of the valley in general proximity of historical dairy operations.
(The sixth location is not a specific Technical Specification requirement.)
Revision 20 Date 7/30/96 4-la.
I Table 4.1 Radiological Environmental Monitorina Stations")
' Exposure Pathway Sample Location Dista ce and/or Sample and Designated Code (2)
(km)(5)
Direction (5) 1.
AIRBORNE (Radioiodine and Particulate)
AP/CF-11 River Station 1.9 SSE No. 3.3 AP/CF-12 N. Hinsdale, NH 3.6 NNW AP/CF-13 Hinsdale Substation 3.1 E
AP/CF-14 Northfield. HA 11.3 SSE AP/CF-15 Tyler Hill Road (4) 3.2 WNW A'P/CF-21 Spofford take 16.1 NNE 2.
WATERBORNE a.
Surface WR-11 River Station 1.9 Downriver.
No. 3.3 WR-21 Rt. 9 Bridge 12.8 Upriver b.
Ground WG-11 Plant Well On-Site WG-12 Vernon Nursing Well 2.0 SSE WG-22 Skibniowsky Well 14.3 N
c.
Sediment SE-11 Shoreline Downriver 0.8 SSE From SE-12 North Storm 0.15 E
Shoreline Drain Outfa11(3) 3.
INGESTION a.
Hil k(8)
TM-11 Hiller Farm 0.8 WNW l
TM-14 Brown Farm 2.1 S
TM-16 Headow Crest Farm 4.4 WNW/NW TM-18 Blodgett Farm")
3.4 SE TM-24 County Farm 22.5 N
b.
Nixed TG-11 River Station 1.9 SSE Grasses No. 3.3 TG-12 N. Hinsdale NH 3.6 NNW TG-13 Hinsdale Substation 3.1 E
TG-14 Northfield. HA 11.3 SSE TG-15 Tyler Hill Rd.0) 3.2-WNW TG-21 Spofford take 16.1 NNE Revision 20 Date 7/30/96 4-2
m.__
Table 4.1 (Continued)
Radiological Environmental Monitorino Stationsu)
Exposure Pathway Sample Location Distance and/or Sample and Designated Code (2)
(km)(5)
Direction (5) c.
Silage TC-11 Miller Farm 0.8 WNW l
TC-14 Brown Farm 2.1 S
TC-16 Meadow Crest Farm 4.4 WNW/NW TC-18 Blodgett Farm")
County Farm 22.5 N
d.
Fish FH-11 Vernon Pond (6)
(6)
+
FH-21 Rt. 9 Bridge 12.8 Upriver 4.
DIRECT RADIATION OR-1 River Station 1.6 SSE
.No. 3.3 DR-2
- 11. Hinsdale, NH 3.9 NNW OR-3 Hinsdale Substation 3.0 E
OR-4 Northfield, MA 11.0 SSE OR-5 Spofford Lake 16.3 NNE OR-6 Vernon School 0.46 W53 OR-7 Site Boundary 0.27 W
DR-8 Site Boundary (7) 0.25 SW OR-9 Inner Ring-2.1 N
OR-10 Outer Ring 4.6 N
OR-11 Inner Ring 2.0 NNE OR-12 Outer Ring 3.6 NNE OR-13 Inner Ring 1.4 NE OR-14 Outer Ring 4.3 NE OR-15 Inner Ring 1.4 EllE l
OR-16 Outer Ring 2.9 ENE OR-17 Inner Ring 1.2 E
OR-18 Outer Ring 3.0 E
OR-19 Inner Ring 3.5 ESE OR-20 Outer Ring 5.3 ESE OR-21 Inner Ring 1.8 SE OR-22 Outer Ring 3.2 SE OR-23 Inner Ring 1.8 SSE OR-24 Outer Ring 3.9 SSE OR-25 Inner Ring 2.0 S
I J
Revision 20 Date 7/30/96 4-2a
i.
1 i
l l
N 7
i i
., j n
1 b
4 :b 4
TG-12 '
AP/CT !2 e
O' -
HINSDAl.E. N.H.
7 g
' A ' TG-1/.
IH-14 te
-.._)
r=-----
4 )$
t Q
I 9
I b
TCal$
i 8
7C-13 A u/CF-is 8
g l
A M/CF-13 8
PLAT (T E E
~
I M
S23" UCARCE* EAT IM TECU7.? 4-1' YEPJt0N CAM.
e t
8--
I TC= K d #/CF-11
'"* " 4 A
VERNON, V.T.
TC.n
- 12 TM-18 A ta.11 TCa16 0
1 2
3 LILY Palm KILOTTEP.5 i
l
'Tigure 4-2 Environmen:al Sszpling Locations nichi. 5
}= og,213ng R r vicinn 70
[h e r 7 / 7n /cg;
,z
.... ~, -,
..- - ~ _ - _
^
r g 0F'g 3 IO I-(5-12) f 0; DF'i s
=: Combined skin dose factor (see Table 1.1-10)
(mrem-sec/pCi-yr)
- 5. 2.1. 2 ' Plant Stack Noble Gas Activity Honitor Setpoint Example The following setpoint example for the plant stack noble gas activity
.i
- monitors demonstrates the.use of Equations 5-9 and 5-10 for determining
]
setpoints.
The plant stack noble gas activity monitors, referred'to as " Stack Gas I" (RM-17-156) and " Stack Gas II" (RM-17-157), consist of beta sensitive i
scintillation detectors electronics, a ratemeter readout, and a digital scaler which counts the detector output pulses.
A strip chart recorder provides a permanent record of the ratemeter output. The monitors.have=
typical calibration factors. S,. of about 3E+07 cpm per pCi/cc of noble gas.
The nominal plant stack flow is 7.32E+07 cc/sec ((155,000 cfm x 28.300-3 cc/ft )/60 sec/ min).
When monitor responses indicate that activity levels are below the LLDs at the = stack (or A0G) monitors, the relative contribution of each noble gas
- radionuclide can conservativelyLbe approximated by analysis of a sample of off-gas obtained during plant operations at the steam jet air ejector (SJAE).
-This setpoint example is based on the following data (see Table 1.1-10 for DFB, and DF'g):
sJAE Qg DF8:
OF'i s 3
i pCl )
( mrem-m )
( mrem-sec )
sec pC1 -yr pCi -yr Xe-138 1.03E+04 8.83E-03 1.06E-02 Kr-87 4.73E+02 5.92E-03 1.43E-02 Kr-88 2.57E+02 1.47E-02 1.28E-02 Kr-85m 1.20E+02 1.17E-03 2.35E-03 Xe-135 3.70E+02 1.81E-03 3.24E-03 Xe-133 1.97E+01 2.94E-04 5.58E-04
- Revision 20 Date 7/30/96 5-13
L E f AE ~ ' i DFB i
1 0F8c (5-11)
- E6?'#E 1
l EofJAEDF8i - (1.03E+04)(8.83E-03) + (4.73E-02)(5.92E-03) 1
+ (2.57E+02)(1.47E-02) + (1.20E+02)(1.17E-03)
+ (3.70E+02)(1.81E-03) + (1.97E+01)(2.94E-04) 3
- 9.83'E+01 ( Ci-mrem-m /sec-pCi-yr)
EOfJAE
- 1.03E+04 + 4.73E+02 + 2.57E+02 1
+ 1.20E+02 + 3.70E+02 + 1.97E441
= 1.15E+04 pCi/sec l
9.83E41 OFB*
1.15E 44 3
- 8.52E-03 (mrem-m /pci-yr) tb 1
1 R
= 818 S spt 8 7 0F8e I
I
- (818) (3E47) (7.32E47) (8.52E-03)
= 39.348 cpm Next:
EOfAED F't.
OF',
1 (5-11)
EDf
- 1 l
l Revision 20 Date 7/30/96 5 14
j.
E0sm0F's, - (1.03E+04)(1.06E-02) + (4.73E-02)(1.43E-02) g
+ (2.57E+02)(1.28E-02) + (1.20E+02)(2.35E 03)
+ (3.70E+02)(3.24E-03) + (1.97E+01)(5.58E-04)
- 1.14E+02 (pCi-mrem-sec/sec-Ci-yr) i 0F'* = 1.14 E+02 1.1S E +04
- 9.91E-03 (mrem-sec/ Ci-yr) spt y
t R
- 3.000 S skin 9 7 0F'c
~
T7.32E+07) T9.91E-03)
- 124.067 com The setpoint. R,pt. is the lesser of R,*,D a nd R,*p*ti". For the noble e
gas mixture in this example R,*,b i s l e s s th a n R,*pg ". indicating that the kI c
total body dose rate is more restrictive.
Therefore, in this example the
" Stack Gas 'I" and " Stack Gas II" noble gas activity monitors should each be set at 39'348 cpm aboy'e background or at some con'servative value below this l
(such as that wMch might be based on controlling release rates from tlie plant in order to maip.ciin off lite air concentrations below 20 x ECL when averaged Cver an hour), or to account for other minor releases from the waste oil burnere For example, if an administrative limit of 70 percent of the Technical Specification whole body cose limit 500 rem /yr (39.348 cpm) is l
chosen. then the noble gas monitor alarms should be set at no more than 27.543 cpm above background (0.7 x 39.348 - 27.543).
l 5.2.1.3 Basis for the Plant Stack and A0G System Noble Gas Activity Monitor Setpoints The setpoints of. the plant stack and A0G system noble gas activity monitors must ensure that Technical Specification 3.8.E.1.a is not exceeded.
Sections 3.4 and 3.5 show that Equations 3-5 and 3-7 are acceptable methods for determining compliance with that Technical Specification. Which equation (i.e.. dose to total body or skin) is more limiting depends on the n)ble gas mixture.
Therefore, each equation must be considered separately.
The Revision 20 Date 7/30/96 5-15
l l
monitor, may be expressed in terms of Og by dividing by F, the appropriate 1
~
flow rate.
In the case of the plant stack noble gas activity monitors the appropriate flow rate is the' plant stack flow rate and for the A0G noble gas activity monitors the appropriate' flow rate is the A0G system flow rate.
i C,i -
Q
.(5-14) i 1
(pCI) (gCi-) (sec) l 3
3 cm sec cm where:
0, The release rate of noble gas "i' in the mixture.for each noble gas identified ( Ci/sec).
3 F
Appropriate flow rate (cm /sec)
=
Substituting the right half of Equat'lon 5-14 into Equation 5-13 for C,,
yields:
S O
(5-15)
R
=
gt f
t s
e 3
cpm-cm
'gCi '
see (cpm) pC1 sec 3
(cm,
t s
The detector' calibration procedure establishes a-counting efficiency for a reference radionuclide, Xe-133 (half life 5.24 days)..For routine conditions where offgas is processed through the A0G, all short lived gases are decayed away before discharge leaving only long lived radionuclides as the significant contributors to the monitor response.
In this case. Xe-133 as the
, reference radionuclide for the detector counting efficiency is representative of the expected release conditions.
For off normal conditions'that might lead to inclusion of short lived radioac'tivity in the gas stream being released.
Xe-133 as the reference radionuclide is expected to' lead to a conservative response factor for the detectors since the short lived noble gases tend to' have higher energies that can caus'e them to over respond. Therefore in Equation 5-15. one may substitute 5 for 5 where.S, represents the detector 5
9 9
counting efficiency determined from the Xe-133 calibration.
If necessary, the actual concentration and discharge rate of individual gases being released Rev i s i on,_2_,0_
Oate 7/30/96 5-17
i i
from the stack (or A0G) can be' determined by direct grab sample and laboratory analysis during specific periods of interest.
I (5-16)
R-S, 6,
(cpm) ( cpmM) ( sec )
g pCi) pCi 3
sec cm I
The total body dose rate due to noble gases is determined with Equation 3-5:
L Aes=
0.61 E
Og 0F81 (3-5) t t
i 3
(mrem) ( pCi-sec )
( Ci )
( mrem-ci pCi-yr )
yr Ci-m3 sec l
l' Where:
Atbs total body dose rate (mrem /yr) due to noble gases
]
from stack release 3
0.61 (1.0E+06) x (6.11E-07) (pCl-sec/pCi-m )
1E + 06 number of pCi per pCi-(pCi/pCf)
[X/0]T. maximum long term average gamma 6.11E - 07
=
3 atmospheric dispersion factor (sec/m )
Oi the release rate of noble gas "1" in the mixture i
for each noble gas identified (pC1/sec)
(EquivalenttoOfT for noble gases released at the plant stack.)
0FBq total body dose factor (see Table 1.1-10) 3 (mrem-m /pCi-yr)
Revision 20 Date 7/30/96 5.
.. ~. - ~....
...- ~_
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I p* *tc y
UNITED STATES i
j NUCLEAR REGULATORY COMMISSION wAssencron. o.c. 2osss-oooi g
%....p SA'FETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION j
l RELATED TO ONSITE DISPOSAL OF CONTAMINATED SOIL VERMONT YANKEE NUCLEAR POWER CORPORATION-VERHONT YANKEE NUCLEAR POWER STATION
.j r
DOCKET NO. 50-271 INTRODUCTION j
1 By letters dated November 18,199', and July 10, 1992, the Vermont Yankee Nuclear Power Corporation (the licensee) requested approval pursuant to Section 20.2002 of Title 10 of the Code of Federal Regulation; (CFR) for the disposal of licensed material not previously considered in the Vermont Yankee Final Environmental Statement (FES), dated 1972.
The3ermont Yankee request contains:
(a) a detailed description of the V
licensed material to be disposed of, including the physical and chemical prcperties important to risk evaluation, and the proposed manner and ccnditions of waste disposal; (b) an analysis and evaluation of pertinent information on the nature of the environment; (c) the nature and location of other potentially affected licensed and unlicensed facilities; and (d) analyses and procedures to ensure that doses are maintained as-low-as is raasonably achievable and within the dose limits in 10 CFR Part 20.
DESCRIPTION OF WASTE-In 1991, a leak was discovered in a chemistry laboratory drain inside the rtdiation control area (RCA) that allowed discharge from the chemistry itboratory sink to seep directly into the structural fill soil beneath the ficor slab. The fill soil is a 15 foot layer of fine-grained sand with some silt and minor gravel. The area is confined on three sides by existing
. fcundations and on the bottom by bedrock. All of the soil volume under the 150-foot length of buried pipe is contaminated; the total volume is about i
58,500 cubic feet. The end of the pipe has been capped and the area of excavation has been backfilled with concrete to the original floor line so
-that the line is inaccessible.
l New piping for the sink has been run above the floor to the collection tank.
This new piping is accessible over its full length for periodic inspection to preclude a repeat of this event.
The licensee has no way of determining how long the drain has been leaking; in order to bound the potential impacts associated with the leakage, the licensee assumed that the drain line had been leaking fo'r 10 years.. Samples of soil from grade to bedrock were obtained l
.- =_
~
r i.
l from a split-spoon boring through the floor of the chemistry laboratory.
Samples were analyzed for chemical and radionuclide distribution and concentration. Estimated amount of the principal radionuclides bound in the contaminated soil are listed in Table 1.
The activity remaining after a 20-year decay period are also presented in the table.
Table 1 Radionuclide Activity and Concentration Nuclide (hal f-life) -
Activity Activity (in years) yCi pCi H-3 (12.2) 8.0E+04 2.6E+04 Hn-54 (0.85) 5.4E+01 4.9E-06 Fe-55.(2.7) 4.4E+02 2.6E+04 Co-60 (5.27) 4.1E+02 3.0E+01 Cs-134 (2.06) 3.9E+01 4.8E-02 Cs-127 (30.17) 1.4E4 2 8.7E+01 Sr-90 (28.6) 3.2E-01 2.0E-01
- Activity after 10 years of weekly " batch" releases
- That activity after a 20 year decay period.
The chemistry laboratory is located in the lower level of the office building at the north end of the turbine building complex. During plant construction, this' area was excavated to bedrock,15 feet below the chemistry laboratory (El. 233 feet). The area under the laboratory was then filled to its current grade and the concrete laboratory floor was poured.
It is impractical to remove this contaminated material because it is located underneath building
. structures.
PROPOSED DISPOSAL HETH00
~
The licensee proposed to leave the contaminated soil in place. By terminating the release of liquids into the failed drain line, there is no significant driving force to cause any further movement of the activity now in the soll below the chemistry laboratory floor any deeper toward the groundwater level.
.The natural groundwater surface appears to be below the bedrock surface beneath the chemistry laboratory.
The total quantity now present is sufficiently small that it does not present a direct radiation exposur e hazard l
in the chemistry laboratory.
To remove the material would, however, require i
major excavation under the laboratory floor in proximity to the reactor building foundation and other critical structures, and would directly expose workers performing the excavation to the hazard.
The direct exposure, as well
l
' l cs potential airborne exposures to workers performing remediation, outweigh the risk of leaving the contamination in place, and exceed by far the potential risk to a future population from leaving the contaminated soil where it is. There is no practical way for.this material to be removed from the plant at this time.
RADIOLOGICAL IMPACTS The licensee evaluated the following potential exposure pathways to members of l
the general public from the. radionuclides in'the contaminated soil:
(1) external exposure caused by farming on the contaminated grounds, (2) internal exposure caused by inhaling of resuspended radionuclides, and (3) internal exposure from ingesting groundwater, and water from onsite potable wells.
Table 2 presents the doses calculated by the licensee for the max'imum exposed member of the public from the contaminated soil under the floor of the chemistry laboratory.
These doses are based on the radionuclide activities in Table 1.
The doses were calculated for an inadvertent intruder for the following pathways:
ingestion of food from crops raised on contaminated land, ingestion of milk from cows grazing on the contaminated land, and inhalation of suspended material.
It is also assumed that the family and animals raised on the land also drink water from the contaminated land and breathe only air affected by the contaminated area.
Table 2 Intruder Exposures Pathway Whole Body Organ (mrem)
(mrem)
Orinking, water ingestion 2.5E-05 6.3E-05 l
Irrigation exposure pathway 1.2E-04 4.0E-04 l
Well water ingestion 3.8E-01 1.9E-01 Direct ground plane 2.7E-01 0.0E-00 Inhalation (resuspension) 1.1E-01 6.5E-01 Leafy vsgetable 2.5E-02 2.4E-01 Cow milk 1.6E-01 1.5E-01 The itcensee conservatively calculated these values with the assumption that the total exhumation of the 58,500 cubic feet of radioactive material and spreading in a layer equivalent to the plow depth, results in a continuous i
annual exposure of less than 1 mrem.
This is a small fraction of the 300 mrem received annually by members of the. general public in the United States and Canada from sources of natural background radiation.
The guidelines used by the fiRC staff for onsite disposal of licensed material and the staff evaluation of how each guideline has been satisfied are given in l
Table 3.
prvkinn 70 nmc 7/10/96 E-6
l l
l The staff has reviewed the licensee's calculational methods and assemptions-rnd find that they are consistent with NUREG-Il01, "Onsite Disposal of Radioactive Waste," Volunies 1 and 2 November 1986 and February 1987, and Regulatory Guide 1.109, " Calculation of Annual Doses to Han From Routine i
R21 eases of Reactor Effluent for the Purpose of Evaluating Compliance With 10 CFR Part 50, Appendix I," Revision 1 (October 1977). The staff finds the assessment methodology acceptable.
On this basis, the staff finds the licensee's procedures and amendments acceptable as documented in this safety evaluation.
This safety evaluation will be added to the Itcensee's Offsite Dose Calculation Hanual (00CH). No future modifications are necessary prior to decommissioning of the plant.
The licensee's proposal to dispose of the contaminated soil under the chemistry laboratory (onsite) in a manner described in the Vermont Yankee submittal dated July 10, 1992, is acceptable.
I a
I l
i llevision 70 Date 7/30/96 E-7
_s_
l Table 3 Guidelines for Onsite Disposal of Licensed Haterial The radioactive material should The nature of the disposed l
be disposed of in such a manner material-makes it unlikely that it that it is unlikely that the would be recycled to the general material would be recycled.
public.
Ooses to the total body and any This guideline is addressed in body organ of a minimally Table 2.
exposed individual (a member of the general public or a non-l occupationally exposed worker)
)
l from the probable pathways of exposure to the disposed I
material should be less than 1 mrem / year.
Doses to the total body and any 18ecause the material is insitu, body organ of an inadvertent the staff considers the ma'ximally intruder from the probable exposed individual scenario to pathways of exposure should be also address the intruder less than 5 mrem / year.
scenario.
Doses to the total body and any Even if recycling were to occur body organ of an individual from after release from regulatory assumed recycling of the control, the dose to a maximally disposed material at the time exposed member of the public is the disposal site is released not expected to exceed 1 from regulatory control from all mrem / year, based on exposure likely pathways of exposure scenarios considered in this should be less than 1 mrem.
analysi s.
Principal Contributor:
J. Hinns Date:
l l
l' i
l l
i Revi9on 20 Date 7/30/96 EJ l
APPENDIX 1 RADI0 ACTIVE LIQUID. GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS Reauirement Technical Specification 6.14.A requires that licensee initiated major changes to the radioactive waste systems (liquid. gaseous, and solid) be reported to the Commission in the. Semiannual i
Radioactive Effluent Release Report for the peri'ad in which the
~
evaluation was reviewed by the Plant Operation Rcview Committee.
l-
, ~
ResDonse:
There were no licensee-initiated major changes to the radioactive waste systems during this reporting period.
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